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Sample records for processing plutonium solutions

  1. PROCESS OF ELIMINATING HYDROGEN PEROXIDE IN SOLUTIONS CONTAINING PLUTONIUM VALUES

    Science.gov (United States)

    Barrick, J.G.; Fries, B.A.

    1960-09-27

    A procedure is given for peroxide precipitation processes for separating and recovering plutonium values contained in an aqueous solution. When plutonium peroxide is precipitated from an aqueous solution, the supernatant contains appreciable quantities of plutonium and peroxide. It is desirable to process this solution further to recover plutonium contained therein, but the presence of the peroxide introduces difficulties; residual hydrogen peroxide contained in the supernatant solution is eliminated by adding a nitrite or a sulfite to this solution.

  2. 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.)

  3. 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.

  4. A portable concentrator for processing plutonium containing solutions

    International Nuclear Information System (INIS)

    Chamberlain, D.B.; Conner, C.; Chen, L.

    1995-01-01

    This report describes a horizontal, compact agitated-film concentrator called a Rototherm, manufactured by Artisan Industries, Inc. which can be used to process aqueous solutions of radioactive wastes containing plutonium. The unit is designed to concentrate liquid streams to a high-solid content slurry

  5. Production of Plutonium Metal from Aqueous Solutions

    Energy Technology Data Exchange (ETDEWEB)

    Orth, D.A.

    2003-01-16

    The primary separation of plutonium from irradiated uranium by the Purex solvent extraction process at the Savannah River Plant produces a dilute plutonium solution containing residual fission products and uranium. A cation exchange process is used for concentration and further decontamination of the plutonium, as the first step in the final preparation of metal. This paper discusses the production of plutonium metal from the aqueous solutions.

  6. Plutonium recovery from carbonate wash solutions

    International Nuclear Information System (INIS)

    Gray, J.H.; Reif, D.J.; Chostner, D.F.; Holcomb, H.P.

    1991-01-01

    540Periodically higher than expected levels of plutonium are found in carbonate solutions used to wash second plutonium cycle solvent. The recent accumulation of plutonium in carbonate wash solutions has led to studies to determine the cause of that plutonium accumulation, to evaluate the quality of all canyon solvents, and to develop additional criteria needed to establish when solvent quality is acceptable. Solvent from three canyon solvent extraction cycles was used to evaluate technology required to measure tributyl phosphate (TBP) degradation products and was used to evaluate solvent quality criteria during the development of plutonium recovery processes. 1 fig

  7. Cation exchange process for recovery of plutonium from laboratory solutions containing chloride

    International Nuclear Information System (INIS)

    Gray, L.W.

    1978-10-01

    A cation exchange technique was developed for the separation of plutonium from laboratory solutions containing either Pu(III) or Pu(III)--Pu(IV) mixtures in acidic solutions containing chloride ions. The procedure consists of adjusting the acid concentration to less than one molar and adjusting the valence of the plutonium ion to the (III) state, if necessary. The adjusted solution is fed to a cation exchange column and washed with distilled water to remove residual chlorides from the column. Plutonium is then eluted from the column with 5M nitric acid containing 0.34M sulfamic acid. This procedure was used to separate plutonium from 1.2M chloride solution on a production-scale column. Typical plutonium recovery was 99.97%, while greater than 96% of the original chloride was rejected

  8. 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

  9. Selecting a plutonium vitrification process

    Energy Technology Data Exchange (ETDEWEB)

    Jouan, A. [Centre d`Etudes de la Vallee du Rhone, Bagnols sur Ceze (France)

    1996-05-01

    Vitrification of plutonium is one means of mitigating its potential danger. This option is technically feasible, even if it is not the solution advocated in France. Two situations are possible, depending on whether or not the glass matrix also contains fission products; concentrations of up to 15% should be achievable for plutonium alone, whereas the upper limit is 3% in the presence of fission products. The French continuous vitrification process appears to be particularly suitable for plutonium vitrification: its capacity is compatible with the required throughout, and the compact dimensions of the process equipment prevent a criticality hazard. Preprocessing of plutonium metal, to convert it to PuO{sub 2} or to a nitric acid solution, may prove advantageous or even necessary depending on whether a dry or wet process is adopted. The process may involve a single step (vitrification of Pu or PuO{sub 2} mixed with glass frit) or may include a prior calcination step - notably if the plutonium is to be incorporated into a fission product glass. It is important to weigh the advantages and drawbacks of all the possible options in terms of feasibility, safety and cost-effectiveness.

  10. In-line measurement of plutonium and americium in mixed solutions

    International Nuclear Information System (INIS)

    Li, T.K.

    1981-01-01

    A solution assay instrument (SAI) has been developed at the Los Alamos National Laboratory and installed in the plutonium purification and americium recovery process area in the Los Alamos Plutonium Processing Facility. The instrument is designed for accurate, timely, and simultaneous nondestructive analysis of plutonium and americium in process solutions that have a wide range of concentrations and Am/Pu ratios. For a 25-mL sample, the assay precision is 5 g/L within a 2000-s count time

  11. IMPROVED PROCESS OF PLUTONIUM CARRIER PRECIPITATION

    Science.gov (United States)

    Faris, B.F.

    1959-06-30

    This patent relates to an improvement in the bismuth phosphate process for separating and recovering plutonium from neutron irradiated uranium, resulting in improved decontamination even without the use of scavenging precipitates in the by-product precipitation step and subsequently more complete recovery of the plutonium in the product precipitation step. This improvement is achieved by addition of fluomolybdic acid, or a water soluble fluomolybdate, such as the ammonium, sodium, or potassium salt thereof, to the aqueous nitric acid solution containing tetravalent plutonium ions and contaminating fission products, so as to establish a fluomolybdate ion concentration of about 0.05 M. The solution is then treated to form the bismuth phosphate plutonium carrying precipitate.

  12. Aqueous Solution Chemistry of Plutonium

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-28

    Things I have learned working with plutonium: Chemistry of plutonium is complex; Redox equilibria make Pu solution chemistry particularly challenging in the absence of complexing ligands; Understanding this behavior is key to successful Pu chemistry experiments; There is no suitable chemical analog for plutonium.

  13. Density of nitric acid solutions of plutonium; Densite des solutions nitriques de plutonium

    Energy Technology Data Exchange (ETDEWEB)

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

    1960-07-01

    The report is intended to furnish an expression making it possible to determine the density of a nitric acid solution of plutonium. Under certain defined experimental conditions, the equation found makes it possible to deduce, for a solution whose concentration, free acidity and temperature are known, the corresponding value of the density of that solution. (author) [French] L'expose a pour but de donner une formule permettant la determination de la densite d'une solution nitrique de plutonium. Suivant certaines conditions experimentales precisees, l'equation trouvee permet, pour une solution dont la concentration, l'acidite libre nitrique et la temperature sont donnees, de deduire la valeur correspondant de la densite de cette solution. (auteur)

  14. Removal of sulfamic acid from plutonium sulfamate--sulfamic acid solution

    International Nuclear Information System (INIS)

    Gray, L.W.

    1978-10-01

    Plutonium metal can be readily dissolved in aqueous solutions of sulfamic acid. When the plutonium sulfamate--sulfamic acid solutions are added to normal purex process streams, the sulfamate ion is oxidized by addition of sodium nitrite. This generates sodium sulfate which must be stored as radioactive waste. When recovery of ingrown 241 Am or storage of the dissolved plutonium must be considered, the sulfamate ion poses major and undesirable precipitation problems in the process streams. The present studies show that 40 to 80% of the sulfamate present in the dissolver solutions can be removed by precipitation as sulfamic acid by the addition of concentrated nitric acid. Addition of 64% nitric acid allows precipitation of 40 to 50% of the sulfamate; addition of 72% nitric acid allows precipitation of 50 to 60% of the sulfamate. If the solutions are chilled, additional sulfamic acid will precipitate. If the solutions are chilled to -10 0 C, about 70 to 80% of the orginal sulfamic acid in the dissolver will precipitate. A single, low-volume wash of the sulfamic acid crystals with concentrated nitric acid will decontaminate the crystals to a plutonium content of 5 dis/(min-gram)

  15. 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

  16. PLUTONIUM PURIFICATION PROCESS EMPLOYING THORIUM PYROPHOSPHATE CARRIER

    Science.gov (United States)

    King, E.L.

    1959-04-28

    The separation and purification of plutonium from the radioactive elements of lower atomic weight is described. The process of this invention comprises forming a 0.5 to 2 M aqueous acidffc solution containing plutonium fons in the tetravalent state and elements with which it is normally contaminated in neutron irradiated uranium, treating the solution with a double thorium compound and a soluble pyrophosphate compound (Na/sub 4/P/sub 2/O/sub 7/) whereby a carrier precipitate of thorium A method is presented of reducing neptunium and - trite is advantageous since it destroys any hydrazine f so that they can be removed from solutions in which they are contained is described. In the carrier precipitation process for the separation of plutonium from uranium and fission products including zirconium and columbium, the precipitated blsmuth phosphate carries some zirconium, columbium, and uranium impurities. According to the invention such impurities can be complexed and removed by dissolving the contaminated carrier precipitate in 10M nitric acid, followed by addition of fluosilicic acid to about 1M, diluting the solution to about 1M in nitric acid, and then adding phosphoric acid to re-precipitate bismuth phosphate carrying plutonium.

  17. Plans and equipment for criticality measurements on plutonium-uranium nitrate solutions

    International Nuclear Information System (INIS)

    Lloyd, R.C.; Clayton, E.D.; Durst, B.M.

    1982-01-01

    Data from critical experiments are required on the criticality of plutonium-uranium nitrate solutions to accurately establish criticality control limits for use in processing and handling of breeder type fuels. Since the fuel must be processed both safely and economically, it is necessary that criticality considerations be based on accurate experimental data. Previous experiments have been reported on plutonium-uranium solutions with Pu weight ratios extending up to some 38 wt %. No data have been presented, however, for plutonium-uranium nitrate solutions beyond this Pu weight ratio. The current research emphasis is on the procurement of criticality data for plutonium-uranium mixtures up to 60 wt % Pu that will serve as the basis for handling criticality problems subsequently encountered in the development of technology for the breeder community. Such data also will provide necessary benchmarks for data testing and analysis on integral criticality experiments for verification of the analytical techniques used in support of criticality control. Experiments are currently being performed with plutonium-uranium nitrate solutions in stainless steel cylindrical vessels and an expandable slab tank system. A schematic of the experimental systems is presented

  18. Reclamation of plutonium from pyrochemical processing residues

    International Nuclear Information System (INIS)

    Gray, L.W.; Gray, J.H.; Holcomb, H.P.; Chostner, D.F.

    1987-04-01

    Savannah River Laboratory (SRL), Savannah River Plant (SRP), and Rocky Flats Plant (RFP) have jointly developed a process to recover plutonium from molten salt extraction residues. These NaCl, KCL, and MgCl 2 residues, which are generated in the pyrochemical extraction of 241 Am from aged plutonium metal, contain up to 25 wt % dissolved plutonium and up to 2 wt % americium. The overall objective was to develop a process to convert these residues to a pure plutonium metal product and discardable waste. To meet this objective a combination of pyrochemical and aqueous unit operations was used. The first step was to scrub the salt residue with a molten metal (aluminum and magnesium) to form a heterogeneous ''scrub alloy'' containing nominally 25 wt % plutonium. This unit operation, performed at RFP, effectively separated the actinides from the bulk of the chloride salts. After packaging in aluminum cans, the ''scrub alloy'' was then dissolved in a nitric acid - hydrofluoric acid - mercuric nitrate solution at SRP. Residual chloride was separated from the dissolver solution by precipitation with Hg 2 (NO 3 ) 2 followed by centrifuging. Plutonium was then separated from the aluminum, americium and magnesium using the Purex solvent extraction system. The 241 Am was diverted to the waste tank farm, but could be recovered if desired

  19. 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.

  20. A method for the gravimetric determination of plutonium in pure plutonium nitrate concentrate solution

    International Nuclear Information System (INIS)

    Mair, M.A.; Savage, D.J.

    1986-12-01

    Plutonium nitrate solution is treated with sulphuric acid before being heated and finally ignited. The stoichiometric plutonium dioxide so formed is weighed and hence the plutonium content is calculated. (author)

  1. Liquid waste processing from plutonium (III) oxalate precipitation

    International Nuclear Information System (INIS)

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

    1990-01-01

    Plutonium (III) oxalate filtrates contain about 0.2M oxalic acid, 0.09M ascorbic acid, 0.05M hydrazine, 1M nitric acid and 20-100 mg/l of plutonium. The developed treatment of liquid wastes consist in two main steps: a) Distillation to reduce up to 10% of the initial volume and refluxing to destroy organic material. Then, the treated solution is suitable to adjust the plutonium at the tetravalent state by addition of hydrogen peroxide and the nitric molarity up to 8.6M. b) Recovery and purification of plutonium by anion exchange using two columns in series containing Dowex 1-X4 resin. With the proposed process, it is possible to transform 38 litres of filtrates with 40mg/l of Pu into 0.1 l of purified solution with 15-20g/l of Pu. This solution is suitable to be recycled in the Pu (III) oxalate precipitation process. This process has several potential advantages over similar liquid waste treatments. These include: 1) It does not increase the liquid volume. 2) It consumes only few reagents. 3) The operations involved are simple, requiring limited handling and they are feasible to automatization. 4) The Pu recovery factor is about 99%. (Author) [es

  2. Review of criticality safety benchmark data of plutonium solution in ICSBEP handbook

    International Nuclear Information System (INIS)

    Yamamoto, Toshihiro; Miyoshi, Yoshinori; Okubo, Kiyoshi

    2003-01-01

    The criticality data of plutonium solutions published in the ICSBEP Handbook were reviewed. Criticality data for lower plutonium concentration and higher 240 Pu content, which correspond to a reprocessing process condition, are very scarce and hence the criticality data in this area are desired. While the calculated k eff 's with ENDF/B-V show the dependence of the plutonium concentration, the dependence has been corrected in JENDL-3.3 because of energy distribution of the capture cross section of 239 Pu. Based on the generalized perturbation theory, the sensitivity coefficient of k eff with respect to fission and capture cross section in plutonium solutions were obtained. In a plutonium solution with a lower concentration, cross sections in the thermal energy less than 0.1 eV have significant effects on the criticality. On the other hand, the criticality of higher concentration plutonium solutions is mostly dominated by cross sections in the energy range larger than 0.1 eV. Regarding the effect of 240 Pu on criticality, the capture cross section 240 Pu around the resonance peak near 1 eV is dominant regardless of the concentration. (author)

  3. Method of stripping plutonium from tributyl phosphate solution which contains dibutyl phosphate-plutonium stable complexes

    International Nuclear Information System (INIS)

    Ochsenfeld, W.; Schmieder, H.

    1976-01-01

    Fast breeder fuel elements which have been highly burnt-up are reprocessed by extracting uranium and plutonium into an organic solution containing tributyl phosphate. The tributyl phosphate degenerates at least partially into dibutyl phosphate and monobutyl phosphate, which form stable complexes with tetravalent plutonium in the organic solution. This tetravalent plutonium is released from its complexed state and stripped into aqueous phase by contacting the organic solution with an aqueous phase containing tetravalent uranium. 6 claims, 1 drawing figure

  4. The influence of plutonium concentration and solution flow rate on the effective capacity of macroporous anion exchange resin

    International Nuclear Information System (INIS)

    Marsh, S.F.; Gallegos, T.D.

    1987-07-01

    The principal aqueous process used to recover and purify plutonium at the Los Alamos Plutonium Facility is anion exchange in nitric acid. Previous studies with gel-type anion exchange resin have shown an inverse relationship between plutonium concentration in the feed solution and the optimum flow rate for this process. Because gel-type resin has been replaced with macroporous resin at Los Alamos, the relationship between plutonium concentration and solution flow rate was reexamined with the selected Lewatit MP-500-FK resin using solutions of plutonium in nitric acid and in nitric acid with high levels of added nitrate salts. Our results with this resin differ significantly from previous data obtained with gel-type resin. Flow-rate variation from 10 to 80 liters per hour had essentially no effect on the measured quantities of plutonium sorbed by the macroporous resin. However, the effect of plutonium concentration in the feed solutions was pronounced, as feed solutions that contained the highest concentrations of plutonium also produced the highest resin loadings. The most notable effect of high concentrations of dissolved nitrate salts in these solutions was an increased resin capacity for plutonium at low flow rates. 16 refs., 7 figs., 2 tabs

  5. Corrosion performance of several metals in plutonium nitrate solution

    International Nuclear Information System (INIS)

    Takeda, Seiichiro; Nagai, Takayuki; Yasu, Shozo; Koizumi, Tsutomu

    1995-01-01

    Corrosion behavior of several metals exposed in plutonium nitrate solution was studied. Plutonium nitrate solution with the plutonium concentration ranging from 0.01 to 300 g/l was used as a corrosive medium. Specimens tested were type 304 ULC (304 ULC) stainless steel, type 310 Nb (310 Nb) stainless steel, titanium (Ti), titanium-5% tantalum alloy (Ti-5Ta), and zirconium (Zr). Corrosion behavior of these metals in plutonium nitrate solution was evaluated through examining electrochemical characteristics and corrosion rates obtained by weight loss measurement. From the results of the corrosion tests, it was found that the corrosion rate of stainless steels i.e. 304 ULC and 310 Nb, increases by the presence of plutonium in nitric acid solution. The corrosion potential of the stainless steels shifted linearly towards the noble direction as the concentration of plutonium increases. It is thought that the shifts in corrosion potential of the stainless steels to the noble direction results an increase in anodic current and, hence, corrosion rate. Valve metals, i.e. Ti, Ti-5Ta and Zr, showed good corrosion resistance over the whole range of plutonium concentration examined here. (author)

  6. 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)

  7. 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)

  8. Gamma ray NDA assay system for total plutonium and isotopics in plutonium product solutions

    International Nuclear Information System (INIS)

    Cowder, L.R.; Hsue, S.T.; Johnson, S.S.; Parker, J.L.; Russo, P.A.; Sprinkle, J.K.; Asakura, Y.; Fukuda, T.; Kondo, I.

    1979-01-01

    A LASL-designed gamma-ray NDA instrument for assay of total plutonium and isotopics of product solutions at Tokai-Mura is currently installed and operating. The instrument is, optimally, a densitometer that uses radioisotopic sources for total plutonium measurements at the K absorption edge. The measured transmissions of additional gamma-ray lines from the same radioisotopic sources are used to correct for self-attenuation of passive gamma rays from plutonium. The corrected passive data give the plutonium isotopic content of freshly separated to moderately aged solutions. This off-line instrument is fully automated under computer control, with the exception of sample positioning, and operates routinely in a mode designed for measurement control. A one-half percent precision in total plutonium concentration is achieved with a 15-minute measurement

  9. Plutonium solution storage in plastic bottles: Operational experience and safety issues

    International Nuclear Information System (INIS)

    Conner, W.V.

    1995-01-01

    Computer spread sheet models were developed to gain a better understanding of the factors that lead to pressurization and failure of plastic bottles containing plutonium solutions. These models were developed using data obtained from the literature on gas generation rates for plutonium solutions. Leak rates from sealed plastic bottles were obtained from bottle leak tests conducted at Rocky Flats. Results from these bottle leak tests showed that narrow mouth four liter bottles will seal much better than wide mouth four liter bottles. The gas generation rate and leak rate data were used to develop models for predicting the rate of pressurization and maximum pressures expected in sealed bottles of plutonium solution containing various plutonium and acid concentrations. The computer models were used to develop proposed time limits for storing or transporting plutonium solutions in sealed plastic bottles. For plutonium solutions containing 1.5 g/l plutonium, storage in sealed bottles should not be allowed. However, transportation of higher concentration plutonium solution in sealed bottles is required, and safe transportation times of 1 shift to 6 days are proposed

  10. 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

  11. 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

  12. Plutonium scrap processing at the Los Alamos Scientific Laboratory

    International Nuclear Information System (INIS)

    Nixon, A.E.; McKerley, B.J.; Christensen, E.L.

    1980-01-01

    The Los Alamos Scientific Laboratory currently has the newest plutonium handling facility in the nation. Los Alamos has been active in the processing of plutonium almost since the discovery of this man-made element in 1941. One of the functions of the new facility is the processing of plutonium scrap generated at LASL and other sites. The feed for the scrap processing program is extremely varied, and a wide variety of contaminants are often encountered. Depending upon the scrap matrix and contaminants present, the majority of material receives a nitric acid/hydrofluoric acid or nitric acid/calcium fluoride leach. The plutonium nitrate solutions are then loaded onto an anion exchange column charged with DOWEX 1 x 4, 50 to 100 mesh, nitrate form resin. The column is eluted with 0.48 M hydroxyl amine nitrate. The Pu(NO 3 ) 3 is then precipitated as plutonium III oxalate which is calcined at 450 to 500 0 C to yield a purified PuO 2 product

  13. 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

  14. Determination of free acid in plutonium (IV) solutions - thermometrically, potentiometrically

    International Nuclear Information System (INIS)

    Williams, T.L.; Tucker, G.M.; Huff, G.A.; Jordan, L.G.

    1981-09-01

    The thermometric titration technique was found to offer certain advantages over potentiometry in the determination of free acid in Pu(IV) solutions. The thermometric technique was applied to the determination of free acid in plutonium nitrate solutions using potassium fluoride to suppress the hydrolytic interference of plutonium(IV). The results indicate that 0.2 to 2.0 milliequivalents of free acid can be determined with acceptable bias and precision in solutions containing up to 30 milligrams of plutonium. In contrast, neither the thermometric nor the potentiometric technique was suitable for samples containing more than eight milligrams of plutonium complexed with potassium oxalate

  15. CSER 00-003: Criticality Safety Evaluation report for PFP Magnesium Hydroxide Precipitation Process for Plutonium Stabilization Glovebox 3

    International Nuclear Information System (INIS)

    LAN, J.S.

    2000-01-01

    This Criticality Safety Evaluation Report analyzes the stabilization of plutonium/uranium solutions in Glovebox 3 using the magnesium hydroxide precipitation process at PFP. The process covered are the receipt of diluted plutonium solutions into three precipitation tanks, the precipitation of plutonium from the solution, the filtering of the plutonium precipitate from the solution, the scraping of the precipitate from the filter into boats, and the initial drying of the precipitated slurry on a hot plate. A batch (up to 2.5 kg) is brought into the glovebox as plutonium nitrate, processed, and is then removed in boats for further processing. This CSER establishes limits for the magnesium hydroxide precipitation process in Glovebox 3 to maintain criticality safety while handling fissionable material

  16. 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.

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

    International Nuclear Information System (INIS)

    Germain, M.; Gillet, B.; Pasquiou, J.Y.

    1989-01-01

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

  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. Treatment of plutonium contaminated ashes by electrogenerated Ag(II): a new, simple and efficient process

    International Nuclear Information System (INIS)

    Madic, C.; Saulze, J.L.; Bourges, J.; Lecomte, M.; Koehly, G.

    1990-01-01

    Incineration is a very attractive technique for managing plutonium contaminated solid wastes, allowing for large volume and mass reduction factors. After waste incineration, the plutonium is concentrated in the ashes and an efficient method must be designed for its recovery. To achieve this goal, a process based on the dissolution of plutonium in nitric solution under the agressive action of electrogenerated Ag(II) was developed. This process is very simple, requiring very few steps. Plutonium recovery yields up to 98% can be obtained and, in addition, the plutonium bearing solutions generated by the treatment can be processed by the PUREX technique for plutonium recovery. This process constitutes the basis for the development of industrial facilities: 1) a pilot facility is being built in MARCOULE (COGEMA, UP1 plant), to treat active ash in 1990; 2) an industrial facility will be built in the MELOX plant under construction at MARCOULE (COGEMA plant)

  20. Radiolysis of aqueous solution of plutonium; Radiolyse des solutions aqueuses de plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Pages nee Flon, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1960-07-01

    This study is concerned first with the effects of gamma rays on plutonium aqueous solutions at various valency states, in presence of sulfuric, perchloric, nitric and hydrochloric acids. The main feature is the reduction of Pu VI into Pu V, followed by dismutation from V to IV and VI. For sulfuric and perchloric acid solutions (0,2 N) the following process is given: radiolysis of water produces OH, H{sub 2}O{sub 2}, H and H{sub 2}. H and H{sub 2}O{sub 2} reduce Pu VI while Pu V is oxidised by OH radicals. However the reaction of hydrogen peroxide is slow and leads to an after effect. A parallel study of the action of H{sub 2}O{sub 2} has given a confirmation. Spectrophotometric measurements were carried out on disappearance of Pu VI and formation of Pu IV and it was possible to make determination of G{sub H{sub 2}}{sub O{sub 2}} = 0,8 and G{sub H} - G{sub OH} = 0,8. G{sub H{sub 2}} = 0,41 was measured by gas analysis. The calculation of G{sub {sup -}}{sub H{sub 2}}{sub O} gave 4,35. The re-oxidation of Pu V is dependant on the concentration of sulfate ions. In perchloric acid solution reduction goes on to Pu Ill. Cl{sup -} and NO{sub 3}{sup -} ions inhibit the reduction and even suppress it. The effect of alpha particles both from plutonium and from polonium is very similar to the effect of hydrogen peroxyde. Induction time were observed mainly in presence of HSO{sub 4}{sup -}, depending on the accumulation of H{sub 2}O{sub 2} and (or) of plutonium peroxide. At low acidities, Pu IV peroxide seems to lead to polymer forms. In the more simple cases (H{sub 2}SO{sub 4} and HClO{sub 4} 0,2N), the following yields were found: G (equivalent reduced) = 3.2, G{sub H{sub 2}}{sub O{sub 2}} = 1.35, G{sub H{sub 2}} = 1.6 and G{sub H} - G{sub OH} = 0.1, assuming G{sub HO{sub 2}} = 0.2; and the usual hypothesis on radiolysis of water by alpha particles. Radiation induced oxidation of Pu III into Pu IV was also observed in H{sub 2}SO{sub 4} and HNO{sub 3} aqueous solutions, and

  1. ''FIXBOX'' - a new technique for the reliable conditioning of plutonium waste solutions

    International Nuclear Information System (INIS)

    Bruchertseifer, H.; Sommer, E.; Steinemann, M.; Bart, G.

    1994-01-01

    ''FIXBOX'' - A new technique and facility for the conditioning of plutonium waste solutions has been developed and brought into operation in the Hot-laboratory at PSI, for the solidification of the waste from the research programmes. The facility is situated in glove-boxes for handling alpha activity and gamma-shielded for conditioning of fission product-containing waste. This report gives a brief description of the FIXBOX facility, the procedure and the first results of the cementation of plutonium waste solutions. As a result of this solidification, the actinide waste is homogeneous and strongly bound in the cement. The presence of gluconic acid and other complexing agents in the waste solution will not disturb this process. (author) figs., tabs., refs

  2. 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.

  3. Concentration and purification of plutonium solutions by means of ion-exchange columns

    Energy Technology Data Exchange (ETDEWEB)

    Durham, R W; Aikin, A M

    1953-02-15

    Equilibrium experiments using Dowex 50 ion-exchange resin and nitric acid solutions of Pu{sup 3+}, UO{sub 2}{sup 2+}, Fe{sup 2+} cations have yielded values for the absorption affinities for these ions. Trivalent plutonium was found to be far more strongly absorbed than UO{sub 2}{sup 2+} and Fe{sup 2+}. Column studies have shown that uranium can be completely separated from plutonium even when the initial concentration of uranium is very much greater than that of the plutonium. A plutonium concentration increase of about fifty-fold can be obtained from solutions about 10{sup -3} M in plutonium and 1.0M in nitric acid. The equation K{sub Pu}{sup 3+} = X{sub R} (1-X{sub S}){sup 3} C{sub S}{sup 2}/X{sub S} (1-X{sub R}){sup 3} C{sub R}{sup 2} for estimating the maximum amount of plutonium taken up by a column of resin of unit volume from a solution of total equivalent concentration, C{sub S} , has been shown to hold for values of C{sub S} up to 3 equivalents per litre. X{sub R}, the equivalent fraction of plutonium on the resin, is the number of equivalents of plutonium absorbed by the resin divided by the total capacity of the column. X{sub S}, the equivalent fraction of plutonium in solution, is the equivalent concentration of plutonium divided by the total equivalent concentration of cations in solution. C{sub R} is the total capacity of the resin in milli-equivalents per gram of dry resin. Recommendations have been made for the application and operation of ion-exchange columns in the Plutonium-Extraction Plant. (author)

  4. 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

  5. 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

  6. The chemistry of plutonium in sol-gel processes

    International Nuclear Information System (INIS)

    Lloyd, M.H.; Haire, R.G.

    1978-01-01

    Studies of plutonium chemical behavior conducted in conjunction with plutonia sol-gel process development at ORNL are described. The colloidal solutions produced consist of 'Pu(IV) polymer,' and this is therefore the study of polymeric plutonium behavior. Spectrophotometric, electron diffraction, and electron microscopy studies, in addition to specific studies that were concerned with the colloidal behavior of Pu(IV) polymer, indicate several characteristics of polymer that are not generally recognized. The particle nature of Pu(IV) polymer indicated by electron microscopy, the amorphous-crystalline characteristics of primary polymer particles demonstrated by electron diffraction, and the reversible and irreversible aggregation of the primary particles shown by spectrophotometric techniques present a useful view of the nature of Pu(IV) polymer that has been helpful in solving or understanding various types of processing problems involving plutonium hydrolytic behavior. The colloidal characteristics of Pu(IV) polymer and crystallite growth of primary polymer particles by thermal denitration are also described. (orig.) [de

  7. 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

  8. Addressing mixed waste in plutonium processing

    International Nuclear Information System (INIS)

    Christensen, D.C.; Sohn, C.L.; Reid, R.A.

    1991-01-01

    The overall goal is the minimization of all waste generated in actinide processing facilities. Current emphasis is directed toward reducing and managing mixed waste in plutonium processing facilities. More specifically, the focus is on prioritizing plutonium processing technologies for development that will address major problems in mixed waste management. A five step methodological approach to identify, analyze, solve, and initiate corrective action for mixed waste problems in plutonium processing facilities has been developed

  9. 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.

  10. Uranium/plutonium and uranium/neptunium separation by the Purex process using hydroxyurea

    International Nuclear Information System (INIS)

    Zhu Zhaowu; He Jianyu; Zhang Zefu; Zhang Yu; Zhu Jianmin; Zhen Weifang

    2004-01-01

    Hydroxyurea dissolved in nitric acid can strip plutonium and neptunium from tri-butyl phosphate efficiently and has little influence on the uranium distribution between the two phases. Simulating the 1B contactor of the Purex process by hydroxyurea with nitric acid solution as a stripping agent, the separation factors of uranium/plutonium and uranium/neptunium can reach values as high as 4.7 x 10 4 and 260, respectively. This indicates that hydroxyurea is a promising salt free agent for uranium/plutonium and uranium/neptunium separations. (author)

  11. 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.

  12. Nondestructive assay instrument for measurement of plutonium in solutions

    International Nuclear Information System (INIS)

    Shirk, D.G.; Hsue, F.; Li, T.K.; Canada, T.R.

    1979-01-01

    A nondestructive assay (NDA) instrument that measures the 239 Pu content in solutions, using a passive gamma-ray spectroscopy technique, has been developed and installed in the LASL Plutonium Processing Facility. A detailed evaluation of this instrument has been performed. The results show that the instrument can routinely determine 239 Pu concentrations of 1 to 500 g/l with accuracies of 1 to 5% and assay times of 1 to 2 x 10 3 s

  13. Criticality experiments with annular cylinders containing plutonium solutions; Experiences de criticite sur des cylindres annulaires contenant des solutions de plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Molbert, M; Sauve, A; Houelle, M; Deilgat, E [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The criticality station of Dijon involves three cells, shielded by concrete walls of 1.46 meter thickness. Those cells are designed to contain the criticality experiment apparatus. The engineering building is also involving: one chemical laboratory where plutonium solutions are prepared, one analysis laboratory, several activated solutions storages, several control rooms, One cell contains the B system, which is designed to study: annular cylindrical geometries, slab of 10 cm thickness, interaction between annular cylinders. This report includes the first results given by experiments on annular cylinders defined by their own geometry (outer and inner diameter of ring containing plutonium solutions). Those results have been plotted in curves, for several concentrations and for different reflection conditions (outer or inner light water reflector, cadmium screen), H{sub c} and M{sub c} = f (c) (where H{sub c} is the critical height of solution, M{sub c} is the critical mass, c is the plutonium concentration: 42,3 g/lplutonium and give H{sub c} and M{sub c} versus the distance between the two solutions. - an insulated annular cylinder 500 x 200: incomplete results are published the experiments on this cylinder being unfinished to the date of this present report publication. On this miscellaneous results, we have following informations know: - Screen effect of light water in central hole. Strengthened effect by cadmium foil on the inside wall. - Normalized interaction curves ( {alpha}*H{sub c}/H{sub c{infinity}} ) versus the distance between the two vessels, where H{sub c{infinity}} critical height of an insulated cylinder, shows that: 1) In light water, two cylinders set aside from 15 cm, can be considers like separated. 2) For some configurations, {alpha} vary

  14. Plutonium mobilization from sedimentary sources to solution in the marine environment

    International Nuclear Information System (INIS)

    Noshkin, V.E.; Wong, K.M.

    1979-01-01

    Inventories of plutonium radionuclides greatly in excess of global fallout levels persists in the benthic environments of Bikini and Eniwetok Atolls. It now appears that the atolls have reached a chemical steadystate condition with respect to the partitioning of 239+240 Pu between solution and solid phases of the environment. The mobilized 239+240 Pu has solute-like characteristics, passes rapidly and readily through dialysis membranes, has adsorption characteristics similar to those of fallout plutonium in the open ocean, and exists in solution primarily as some oxidized +5 or +6 chemical species. Water-column profiles of 239+240 Pu taken outside the atolls show a plutonium excess in the deep water mass. This remobilized 239+240 Pu possibly originates from the contaminated sediments previously deposited on the outer slopes of the atolls and surrounding basins

  15. Development of a plutonium solution-assay instrument with isotopic capability

    International Nuclear Information System (INIS)

    Hsue, S.T.; Marks, T.

    1992-01-01

    A new generation of solution-assay instrument has been developed to satisfy all the assay requirements of an aqueous plutonium-recovery operation. The assay is based on a transmission-corrected passive assay technique. We have demonstrated that the system can cover a concentration range of 0.5--300 g/ell with simultaneous isotopic determination. The system can be used to assay input and eluate streams of the recovery operation. The system can be modified to measure low-concentration effluent solutions from the recovery operation covering 0.01--40 g/ell. The same system has also been modified to assay plutonium solutions enriched in 242 Pu. 6 refs

  16. Potentiometric determination of free nitric-acid in trilaurylamine solutions containing plutonium nitrate

    International Nuclear Information System (INIS)

    Perez, J.J.; Saey, J.C.

    1965-01-01

    A potentiometric method of determination of the free nitric acid in trilaurylamine solutions containing plutonium or thorium nitrates is described. The potentiometric titration is carried out in a mixture of benzene and 1,2-dichloro ethane with a standard solution of trilaurylamine as the titrant. When thorium nitrate is present the metal complex is not dissociated then the titration has a single end-point. In the case of plutonium nitrate the partial dissociation of the plutonium complex corresponds to a second point. The experimental error in duplicate analyses of 50 samples is about 1 per cent for free acid concentrations in the range of 0,03 to 0,1 N and plutonium concentrations between 1 to 5 g/l. (authors) [fr

  17. Continuous precipitation process of plutonium salts

    International Nuclear Information System (INIS)

    Richard, P.

    1967-03-01

    This work concerns the continuous precipitation process of plutonium oxalate. Investigations about the solubility of different valence states in nitric-oxalic and in nitric-sulfuric-oxalic medium lead to select the precipitation process of tetravalent plutonium oxalate. Settling velocity and granulometry of tetravalent oxalate plutonium have been studied with variation of several precipitation parameters such as: temperature, acidity, excess of oxalic acid and aging time. Then are given test results of some laboratory continuous apparatus. Conditions of operation with adopted tubular apparatus are defined in conclusion. A flow-sheet is given for a process at industrial scale. (author) [fr

  18. 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

  19. Immobilization of plutonium from solutions on porous matrices by the method of high temperature sorption

    Energy Technology Data Exchange (ETDEWEB)

    Nardova, A.K.; Filippov, E.A. [All Research Institute of Chemical Technologies, Moscow (Russian Federation); Glagolenko, Y.B. [and others

    1996-05-01

    This report presents the results of investigations of plutonium immobilization from solutions on inorganic matrices with the purpose of producing a solid waste form. High-temperature sorption is described which entails the adsorption of radionuclides from solutions on porous, inorganic matrices, as for example silica gel. The solution is brought to a boil with additional thermal process (calcination) of the saturated granules.

  20. Preventing pollution from plutonium processing

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1993-01-01

    The plutonium processing facility at Los Alamos has adopted the strategic goal of becoming a facility that processes plutonium in a way that produces only environmentally benign waste streams. Pollution prevention through source reduction and environmentally sound recycling are being pursued. General approaches to waste reductions are administrative controls, modification of process technologies, and additional waste polishing. Recycling of waste materials, such as spent acids and salts, are technical possibilities and are being pursued to accomplish additional waste reduction. Liquid waste stream polishing to remove final traces of plutonium and hazardous chemical constituents is accomplished through (a) process modifications, (b) use of alternative chemicals and sorbents for residue removal, (c) acid recycling, and (d) judicious use of a variety of waste polishing technologies. Technologies that show promise in waste minimization and pollution prevention are identified. Working toward this goal of pollution prevention is a worthwhile endeavor, not only for Los Alamos, but for the Nuclear Complex of the future

  1. Preventing pollution from plutonium processing

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1995-01-01

    The plutonium processing facility at Los Alamos has adopted the strategic goal of becoming a facility that processes plutonium in a way that produces only environmentally benign waste streams. Pollution prevention through source reduction and environmentally sound recycling are being pursued. General approaches to waste reductions are administrative controls, modification of process technologies, and additional waste polishing. Recycling of waste materials, such as spent acids and salts, are technical possibilities and are being pursued to accomplish additional waste reduction. Liquid waste stream polishing to remove final traces of plutonium and hazardous chemical constituents is accomplished through process modifications, use of alternative chemicals and sorbents for residue removal, acid recycling, and judicious use of a variety of waste polishing technologies. Technologies that show promise in waste minimization and pollution prevention are identified. Working toward this goal of pollution prevention is a worthwhile endeavor , not only for Los Alamos, but for the Nuclear Complex of the future. (author) 12 refs.; 2 figs

  2. SEPARATION OF PLUTONIUM

    Science.gov (United States)

    Maddock, A.G.; Smith, F.

    1959-08-25

    A method is described for separating plutonium from uranium and fission products by treating a nitrate solution of fission products, uranium, and hexavalent plutonium with a relatively water-insoluble fluoride to adsorb fission products on the fluoride, treating the residual solution with a reducing agent for plutonium to reduce its valence to four and less, treating the reduced plutonium solution with a relatively insoluble fluoride to adsorb the plutonium on the fluoride, removing the solution, and subsequently treating the fluoride with its adsorbed plutonium with a concentrated aqueous solution of at least one of a group consisting of aluminum nitrate, ferric nitrate, and manganous nitrate to remove the plutonium from the fluoride.

  3. Plutonium determination in solution with excess hydrofluoric acid

    International Nuclear Information System (INIS)

    Krtil, J.; Kuvik, V.; Spevackova, V.

    1975-01-01

    The determination is described of plutonium in solutions in the presence of fluoride ions resulting from the hydrolysis of PuF 6 . The method is based on reduction of Pu(VI) by excess of Fe(II) and on re-titration of Fe(II) with ceric salt. The effect of fluoride ions on plutonium determination was studied. It was found that a 3 mole excess of HF with respect to Pu decreased the results of Pu determination. The interference of fluoride ions was eliminated by a two-fold evaporation of the solution to be titrated with HNO 3 to dryness or by complex formation with boric arid. The amount of 20.50 mg Pu in the presence of a 10 mole excess of fluoride ions (17 mg HF) was determined with an error of +- 0.09 mg ). (author)

  4. 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

  5. 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.

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

    International Nuclear Information System (INIS)

    Aparo, M.; Mattia, B.; Zeppa, P.; Pagliai, V.; Frazzoli, F.V.

    1989-03-01

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

  7. 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)

  8. Spectrophotometric determination of uranium and plutonium in nitric acid solutions at their co-presence

    International Nuclear Information System (INIS)

    Levakov, B.I.; Mishenev, V.B.; Nezgovorov, N.Yu.; Ryazanova, G.K.; Timofeev, G.A.

    1986-01-01

    The method of spectrophotometric determination of uranium (6) and plutonium (4) in nitric acid solutions is described. Uranium is determined by light absorption of the complex with arsenazo 3 in 0.05 mol/l nitric acid at λ=654 nm, plutonium - by light absorption of the complex with xylenol orange in 0.1 mol/l nitric acid at λ=540 nm. To disguise plutonium, tetravalent and certain trivalent elements DTPA is introduced into photometered solution for uranium determination. The relative root-mean square deviation of determination results does not exceed 0.03 in uranium concenration ranges 0.5-5 μg/ml, of plutonium -1-3 μg/ml

  9. Treatment of plutonium-bearing solutions: A brief survey of the DOE complex

    International Nuclear Information System (INIS)

    Conner, C.; Chamberlain, D.B.; Chen, L.; Vandegrift, G.F.

    1995-03-01

    With the abrupt shutdown of some DOE facilities, a significant volume of in-process material was left in place and still requires treatment for interim storage. Because the systems containing these process streams have deteriorated since shutdown, a portable system for treating the solutions may be useful. A brief survey was made of the DOE complex on the need for a portable treatment system to treat plutonium-bearing solutions. A survey was completed to determine (1) the compositions and volumes of solutions and heels present, (2) the methods that have been used to treat these solutions and heels in the past, and (3) the potential problems that exist in removing and treating these solutions. Based on the surveys and on the Defense Nuclear Facilities Safety Board Recommendation 94-1, design criteria for a portable treatment system were generated

  10. A vision for environmentally conscious plutonium processing

    International Nuclear Information System (INIS)

    Avens, L.R.; Eller, P.G.; Christensen, D.C.; Miller, W.L.

    1998-01-01

    Regardless of individual technical and political opinions about the uses of plutonium, it is virtually certain that plutonium processing will continue on a significant global scale for many decades for the purposes of national defense, nuclear power and remediation. An unavoidable aspect of plutonium processing is that radioactive contaminated gas, liquid, and solid streams are generated. These streams need to be handled in a manner that is not only in full compliance with today's laws,but also will be considered environmentally and economically responsible now and in the future. In this regard, it is indeed ironic that the multibillion dollar and multidecade radioactive cleanup mortgage that the US Department of Energy (and its Russian counterpart) now owns resulted from waste management practices that were at the time in full legal compliance. The theme of this paper is that recent dramatic advances in actinide science and technology now make it possible to drastically minimize or even eliminate the problematic waste streams of traditional plutonium processing operations. Advanced technology thereby provides the means to avoid passing on to our children and grandchildren significant environmental and economic legacies that traditional processing inevitably produces. This paper will describe such a vision for plutonium processing that could be implemented fully within five years at a facility such as the Los Alamos Plutonium Facility (TA55). As a significant bonus, even on this short time scale, the initial technology investment is handsomely returned in avoided waste management costs

  11. Continuous monitoring of plutonium solution in a conversion plant

    International Nuclear Information System (INIS)

    Hassan, B.; Piana, M.; Mousalli, G.; Saukkonen, H.; Hosima, T.; Kawa, T.

    2000-01-01

    This paper describes the implementation of a safeguards Tank Monitoring System (TAMS) in a Plutonium Conversion Plant (PCP). TAMS main objective is to provide the International Atomic Energy Agency (IAEA) (the Agency) with continuous data for safeguards evaluation and review of inventories and flows of plutonium solutions. It has been designed to monitor, in unattended mode, the inventory of each tank and transactions of solutions between tanks, as well as to confirm the absence of borrowing plutonium solutions from and to a neighboring reprocessing plant. The instrumentation consists of one electronic scanner that collects pressure data from electromanometers connected to the tank dip tubes, one uninterruptable power supply and one personal computer operating in a Windows-NT environment. The pressure data transmitted to the acquisition system is saved and converted to volume and density values, coupled with a graph capability to display events in each tank at intervals of 15 seconds. The system operation has not only strengthened the safeguards measures in PCP but also reduced inspection effort while minimizing intrusion to normal plant activities and radiation exposure to personnel. TAMS is a powerful, reliable tool that has significantly improved the effectiveness of safeguards implementation at PCP. The future combined use of TAMS with remote monitoring (RM) will further enhance efficiency of the safeguards measures at PCP. (author)

  12. 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

  13. Long-term behavior of refractory thorium-plutonium dioxide solid solutions

    Energy Technology Data Exchange (ETDEWEB)

    Claparede, Laurent, E-mail: laurent.claparede@umontpellier.fr [ICSM, UMR 5257 CNRS/CEA/Univ. Montpellier/ENSCM, Site de Marcoule, Bât. 426, BP 17171, 30207 Bagnols/Cèze (France); Guigue, Mireille [CEA, Nuclear Energy Division, RadioChemistry & Processes Department, BP 17171, 30207 Bagnols/Cèze (France); Jouan, Gauthier [CEA, Nuclear Energy Division, DTEC Department, BP 17171, 30207 Bagnols/Cèze (France); Nadah, Nassima [CEA, Nuclear Energy Division, RadioChemistry & Processes Department, BP 17171, 30207 Bagnols/Cèze (France); Dacheux, Nicolas [ICSM, UMR 5257 CNRS/CEA/Univ. Montpellier/ENSCM, Site de Marcoule, Bât. 426, BP 17171, 30207 Bagnols/Cèze (France); Moisy, Philippe [CEA, Nuclear Energy Division, RadioChemistry & Processes Department, BP 17171, 30207 Bagnols/Cèze (France)

    2017-01-15

    The long-term behavior of Th{sub 0.87}Pu{sub 0.13}O{sub 2} was examined in nitric acid concentrations. The normalized dissolution rates after 3380 days, range from (1.4 ± 0.2) × 10{sup −6} g m{sup −2} d{sup −1} in 5 M HNO{sub 3} down to (3.2 ± 0.4) × 10{sup −8} g m{sup −2} d{sup −1} in 10{sup −3} M HNO{sub 3}, which confirms the high chemical durability of this solid solution. The amounts of plutonium measured in solution lead to 0.9% and 2.1% of dissolved solid in 1 M and 5 M HNO{sub 3}, respectively. In such conditions, the time required to reach the full dissolution of the material varies from 430 years (5 M HNO{sub 3}) to 18,000 years (10{sup −3} M HNO{sub 3}). Moreover, the partial order related to the proton activity (n = 0.45 ± 0.03) suggests that the dissolution is mainly driven by surface reactions occurring at the solid/liquid interface. The characterization of the leached samples by SEM shows small microstructural modifications (i.e. detachment of crystallites) and the absence of neoformed phase while from PXRD, the unit cell parameter and crystallite size are not significantly affected. - Highlights: • Leaching tests of Th{sub 0.87}Pu{sub 0.13}O{sub 2} were performed for 9 years in several nitric acid solutions. • The high chemical durability of thorium-plutonium oxide solid solutions was confirmed. • The solubility of plutonium(IV) was not controlled by the precipitation of plutonium tetrahydroxide in these experiments.

  14. A vision for environmentally conscious plutonium processing

    International Nuclear Information System (INIS)

    Avens, L.R.; Eller, P.G.; Christensen, D.C.; Miller, W.L.

    1998-01-01

    Regardless of individual technical and political opinions about the uses of plutonium, it is virtually certain that plutonium processing will continue on a significant global scale for many decades for the purposes of national defense, nuclear power, and remediation. An unavoidable aspect of plutonium processing is that radioactively contaminated gas, liquid, and solid waste streams are generated. These streams need to be handled in a manner that not only is in full compliance with today's laws but also will be considered environmentally and economically responsible now and in the future. In this regard, it is indeed ironic that the multibillion dollar and multidecade radioactive cleanup mortgage that the US Department of Energy (and its Russian counterpart) now owns resulted from waste management practices that were at the time in full legal compliance. It is now abundantly evident that in the long run, these practices have proven to be neither environmentally nor economically sound. Recent dramatic advances in actinide science and technology now make it possible to drastically minimize or even eliminate the problematic waste streams of traditional plutonium processing operations. Advanced technology thereby provides the means to avoid passing on to children and grandchildren significant environmental and economic legacies that traditional processing inevitably produces. The authors describe such a vision for plutonium processing that could be implemented fully within 5 yr at a facility such as the Los Alamos National Laboratory Plutonium Facility (TA55). As a significant bonus, even on this short timescale, the initial technology investment is handsomely returned in avoided waste management costs

  15. Precipitation of plutonium oxalate from homogeneous solutions

    International Nuclear Information System (INIS)

    Rao, V.K.; Pius, I.C.; Subbarao, M.; Chinnusamy, A.; Natarajan, P.R.

    1986-01-01

    A method for the precipitation of plutonium(IV) oxalate from homogeneous solutions using diethyl oxalate is reported. The precipitate obtained is crystalline and easily filterable with yields in the range of 92-98% for precipitations involving a few mg to g quantities of plutonium. Decontamination factors for common impurities such as U(VI), Am(III) and Fe(III) were determined. TGA and chemical analysis of the compound indicate its composition as Pu(Csub(2)Osub(4))sub(2).6Hsub(2)O. Data are obtained on the solubility of the oxalate in nitric acid and in mixtures of nitric acid and oxalic acid of varying concentrations. Green PuOsub(2) obtained by calcination of the oxalate has specifications within the recommended values for trace foreign substances such as chlorine, fluorine, carbon and nitrogen. (author)

  16. Waste minimization at a plutonium processing facility

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1995-01-01

    As part of Los Alamos National Laboratory's (LANL) mission to reduce the nuclear danger throughout the world, the plutonium processing facility at LANL maintains expertise and skills in nuclear weapons technologies as well as leadership in all peaceful applications of plutonium technologies, including fuel fabrication for terrestrial and space reactors and heat sources and thermoelectric generators for space missions. Another near-term challenge resulted from two safety assessments performed by the Defense Nuclear Facilities Safety Board and the U.S. Department of Energy during the past two years. These assessments have necessitated the processing and stabilization of plutonium contained in tons of residues so that they can be stored safely for an indefinite period. This report describes waste streams and approaches to waste reduction of plutonium management

  17. Preliminary evaluation of the electrapette for possible use in the glovebox for pipetting plutonium solutions

    Energy Technology Data Exchange (ETDEWEB)

    Hansbury, E.; Ortiz, B.; Roybal, C.

    1990-12-01

    At the Los Alamos Laboratory Plutonium Facility, Solution Assay Instruments (SAIs) are used to provide real-time information on the plutonium (Pu) content of the process stream at various stages in the process. Much of the solution analysis must be carried and as a glovebox to protect the operator from radiation. In order to overcome some of the difficulties usually encountered when working in a glovebox, an electronic solution-volume measuring device called an Electrapette was ordered from Matrix Technologies Corporation. It is said to be highly accurate, simple to use, and can handle the 25 ml of solution required for SAI analyses. It is microprocessor-controlled and comes in two components connected by a detachable cable so that the electronic part can be installed outside the box, while the nosepiece is inside. The two pieces are connected through a plug-in on the glovebox wall. The Electrapette was tested in three sets of experiments: a cold'' lab set, a set run is a hood in a production building, and a third set run in a glovebox using a process solution whose density had been predetermined. The accuracy of the determination could not be determined because the samples had been mixed with other feed before being sent for analysis by the Electrapette. 2 refs., 5 tabs.

  18. 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

  19. Magnetic separation as a plutonium residue enrichment process

    International Nuclear Information System (INIS)

    Avens, L.R.; Gallegos, U.F.; McFarlan, J.T.

    1990-01-01

    Several plutonium contaminated residues have been subjected to Open Gradient Magnetic Separation (OGMS) on an experimental scale. OGMS experiments on graphite and bomb reduction residues resulted in a plutonium rich fraction and a plutonium lean fraction. Values for the bulk quantity rejected to the lean fraction varied between about 20% to 85% of the feed bulk. The plutonium content of the lean fraction can be reduced from about 2% in the feed to the 0.1% to 0.5% range dependent on the portion of the feed rejected to this lean fraction. These values are low enough in plutonium to meet economic discard limits and be considered for direct discard. Magnetic separation of pyrochemical salts gave less favorable results. While a fraction very rich in plutonium could be obtained, the lean fraction plutonium content was too high for direct discard. This may still have chemical processing applications. OGMS experiments at low magnetic field strength on incinerator ash did give two fractions but the plutonium content of each fraction was essentially identical. Thus, no chemical processing advantage was identified for magnetic separation of this residue. 6 refs., 1 fig., 9 tabs

  20. Nondestructive assay of plutonium residue in horizontal storage tanks

    International Nuclear Information System (INIS)

    Marsh, S.F.

    1985-01-01

    Aqueous plutonium recovery and purification processes often involve the temporary storage of plutonium solutions in holding tanks. Because plutonium is known to precipitate from aqueous solutions under certain conditions, there is a continuing need to assay emptied tanks for plutonium residue. A portable gamma spectrometer system, specifically designed for this purpose, provides rapid assay of such plutonium residues in horizontal storage tanks. A means is thus available for the nondestructive analysis of these tanks on a regular schedule to ensure that significant deposits of plutonium are not allowed to accumulate. 5 figs

  1. Properties of concentrated plutonium nitrate solutions

    International Nuclear Information System (INIS)

    Gray, J.H.; Swanson, J.L.

    1978-01-01

    Selected properties were measured for solutions containing about 500 and 700 g/l plutonium (IV) in 4--5M nitric acid: density, viscosity, vapor pressure, boiling point, radiolytic gas (H 2 ) evolution rates, and corrosion rate on Ti and 304L stainless steel. Pu solubility was determined to be 550 to 800 g/l in 2.5 to 7M HNO 3 at ambient temperature and 820 to 860 g/l in 3M HNO 3 at 50 0 C

  2. An intercomparison experiment on isotope dilution thermal ionisation mass spectrometry using plutonium-239 spike for the determination of plutonium concentration in dissolver solution of irradiated fuel

    International Nuclear Information System (INIS)

    Aggarwal, S.K.; Shah, P.M.; Saxena, M.K.; Jain, H.C.; Gurba, P.B.; Babbar, R.K.; Udagatti, S.V.; Moorthy, A.D.; Singh, R.K.; Bajpai, D.D.

    1996-01-01

    Determination of plutonium concentration in the dissolver solution of irradiated fuel is one of the key measurements in the nuclear fuel cycle. This report presents the results of an intercomparison experiment performed between Fuel Chemistry Division (FCD) at BARC and PREFRE, Tarapur for determining plutonium concentration in dissolver solution of irradiated fuel using 239 Pu spike in isotope dilution thermal ionisation mass spectrometry (ID-TIMS). The 239 Pu spike method was previously established at FCD as viable alternative to the imported enriched 242 Pu or 244 Pu; the spike used internationally for plutonium concentration determination by IDMS in dissolver solution of irradiated fuel. Precision and accuracy achievable for determining plutonium concentration are compared under the laboratory and the plant conditions using 239 Pu spike in IDMS. For this purpose, two different dissolver solutions with 240 Pu/ 239 Pu atom ratios of about 0.3 and 0.07 corresponding, respectively, to high and low burn-up fuels, were used. The results of the intercomparison experiment demonstrate that there is no difference in the precision values obtained under the laboratory and the plant conditions; with mean precision values of better than 0.2%. Further, the plutonium concentration values determined by the two laboratories agreed within 0.3%. This exercise, therefore, demonstrates that ID-TIMS method using 239 Pu spike can be used for determining plutonium concentration in dissolver solution of irradiated fuel, under the plant conditions. 7 refs., 8 tabs

  3. Continuous Material Balance Reconciliation for a Modern Plutonium Processing Facility

    International Nuclear Information System (INIS)

    CLARK, THOMASG.

    2004-01-01

    This paper describes a safeguards approach that can be deployed at any modern plutonium processing facility to increase the level of safeguards assurance and significantly reduce the impact of safeguards on process operations. One of the most perplexing problems facing the designers of plutonium processing facilities is the constraint placed upon the limit of error of the inventory difference (LEID). The current DOE manual constrains the LEID for Category I and II material balance areas to 2 per cent of active inventory up to a Category II quantity of the material being processed. For 239Pu a Category II quantity is two kilograms. Due to the large material throughput anticipated for some of the modern plutonium facilities, the required LEID cannot be achieved reliably during a nominal two month inventory period, even by using state-of-the-science non-destructive assay (NDA) methods. The most cost-effective and least disruptive solution appears to be increasing the frequency of material balance closure and thus reducing the throughput being measured during each inventory period. Current inventory accounting practices and systems can already provide the book inventory values at any point in time. However, closing the material balance with measured values has typically required the process to be cleaned out, and in-process materials packaged and measured. This process requires one to two weeks of facility down time every two months for each inventory, thus significantly reducing productivity. To provide a solution to this problem, a non-traditional approach is proposed that will include using in-line instruments to provide measurement of the process materials on a near real-time basis. A new software component will be developed that will operate with the standard LANMAS application to provide the running material balance reconciliation, including the calculation of the inventory difference and variance propagation. The combined measurement system and software

  4. History and stabilization of the Plutonium Finishing Plant (PFP) complex, Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Gerber, M.S., Fluor Daniel Hanford

    1997-02-18

    The 231-Z Isolation Building or Plutonium Metallurgy Building is located in the Hanford Site`s 200 West Area, approximately 300 yards north of the Plutonium Finishing Plant (PFP) (234-5 Building). When the Hanford Engineer Works (HEW) built it in 1944 to contain the final step for processing plutonium, it was called the Isolation Building. At that time, HEW used a bismuth phosphate radiochemical separations process to make `AT solution,` which was then dried and shipped to Los Alamos, New Mexico. (AT solution is a code name used during World War II for the final HEW product.) The process was carried out first in T Plant and the 224-T Bulk Reduction Building and B Plant and the 224-B Bulk Reduction Building. The 224-T and -B processes produced a concentrated plutonium nitrate stream, which then was sent in 8-gallon batches to the 231-Z Building for final purification. In the 231-Z Building, the plutonium nitrate solution underwent peroxide `strikes` (additions of hydrogen peroxide to further separate the plutonium from its carrier solutions), to form the AT solution. The AT solution was dried and shipped to the Los Alamos Site, where it was made into metallic plutonium and then into weapons hemispheres.` The 231-Z Building began `hot` operations (operations using radioactive materials) with regular runs of plutonium nitrate on January 16, 1945.

  5. Field test of plutonium and thorium contaminated clay soils from the Mound Site using the ACT*DE*CON Process

    International Nuclear Information System (INIS)

    Johnson, J.O.; Swift, N.A.; Church, R.H.; Neff, R.A.

    1998-01-01

    A treatability test was run during the summer and fall of 1997 to demonstrate the effectiveness of ACT*DE*CON for removing plutonium and thorium from the clay soils around Mound. ACT*DE*CON is a proprietary solution patented by Selentec. The process utilized a highly selective dissolution of the contaminants by the use of a chemical wash. The pilot scale process involved pretreatment of the soil in an attrition scrubber with ACT*DE*CON solution. This blended solution was then passed through a counter-current extraction chamber where additional contact with ACT*DE*CON solution occurred, followed by a rinse cycle. During this process sand was added to aid contact of the solution with the soil particles. The sand is removed during the rinse step and reused. The chelating agent is separated from the contaminant and recycled back into the process, along with the reverse osmosis permeate. The resulting solution can be further treated to concentrate the contaminant. Three different types of environmental soils were tested -- plutonium and thorium contaminated soils with the natural clay content, and plutonium contaminated soils with a high percentage of fine clay particles. The goal of these tests was to reduce the plutonium levels from several hundreds of pCi/g to between 25 and 75 pCi/g and the thorium from a couple hundred pCi/g to less than 5 pCi/g. The results of these four tests are presented along with a discussion of the operating parameters and the lessons learned relating to full scale implementation at Mound as well as other potential applications of this process

  6. Continuous precipitation process of plutonium salts; Procede continu de precipitation des sels de plutonium

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-03-01

    This work concerns the continuous precipitation process of plutonium oxalate. Investigations about the solubility of different valence states in nitric-oxalic and in nitric-sulfuric-oxalic medium lead to select the precipitation process of tetravalent plutonium oxalate. Settling velocity and granulometry of tetravalent oxalate plutonium have been studied with variation of several precipitation parameters such as: temperature, acidity, excess of oxalic acid and aging time. Then are given test results of some laboratory continuous apparatus. Conditions of operation with adopted tubular apparatus are defined in conclusion. A flow-sheet is given for a process at industrial scale. (author) [French] Cette etude porte sur la precipitation continue de l'oxalate de plutonium. L'etude de la solubilite des differentes valences du plutonium dans des milieux acides nitrique-oxalique, puis nitrique-sulfurique-oxalique conduit a choisir la precipitation de l'oxalate de plutonium tetravalent. L'etude porte ensuite sur la sedimentation et la granulometrie de l'oxalate de Pu{sup 4+} obtenue en faisant varier differents parametres de la precipitation : la temperature, l'acidite, l'exces oxalique et le temps de murissement. La derniere partie traite des resultats obtenus avec plusieurs types d'appareils continus essayes au laboratoire. En conclusion sont donnees les conditions de marche de l'appareil tubulaire adopte, ainsi qu'une extrapolation a l'echelle industrielle sous forme d'un flow-sheet. (auteur)

  7. 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.

  8. Influence of organic components on plutonium and americium speciation in soils and soil solutions

    International Nuclear Information System (INIS)

    Sokolik, G.A.; Ovsyannikova, S.V.; Kimlenko, I.M.

    2003-01-01

    Group composition of humic substances of organic and mineral soils sampled in the 30-km zone of the Chernobyl accident was analyzed for studying influence of organic components on migration properties of plutonium and americium in soils and soil solutions by the method of gel-chromatography and chemical fractionation. It was ascertained that humus of organic soils binds plutonium and americium stronger than humus of mineral soils. Elevated mobility of americium compared to plutonium one stems from lower ability of the latter to from hard to solve organic and organomineral complexes, as well as from its ability to form anionic complexes in soil solutions [ru

  9. 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

  10. 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.

  11. 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

  12. An experimental study of factors in the recovery of plutonium from combustible wastes treated by incineration, pyrolysis and other processes

    International Nuclear Information System (INIS)

    Bamber, D.C.; McDonald, L.A.; Roberts, W.G.; Sutcliffe, P.W.; Wilkins, J.D.

    1984-01-01

    The work described in this report is concerned with the incineration and pyrolysis of plutonium-contaminated combustible wastes, the leaching of the ashes and chars and the subsequent treatment of the leach solutions. A range of ashes and chars have been prepared from a range of plutonium-contaminated materials covering a variety of combustible materials (e.g. PVC, neoprene, Hypalon) and plutonium contaminants [e.g. PuO 2 , Pu(NO 3 ) 4 , (U, Pu)O 2 ]. Treatment temperatures in the range of 550-900 0 C have been investigated, the best results being obtained at or below 700 0 C with pyrolysis followed by char oxidation being the favoured process. A number of methods for treatment of the leach solutions have been considered and some have been investigated experimentally. Extraction of plutonium and americium with tributylphosphate (TBP) from a leach solution conditioned to 0.1 M H/+5 M NO 3 - has been studied. The key stage has been found to be the conditioning step where precautions must be taken to ensure that plutonium-containing precipitates and non-extractable plutonium are not formed. Consideration has also been given to treatment of the americium containing raffinates from a high acid TBP extraction and some methods have been investigated. A range of simple washing experiments have been carried out in order to compare the process with incineration/pyrolysis

  13. 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

  14. Design of plutonium processing facilities

    International Nuclear Information System (INIS)

    Derbyshire, W.; Sills, R.J.

    1982-01-01

    Five considerations for the design of plutonium processing facilities are identified. These are: Toxicity, Radiation, Criticality, Containment and Remote Operation. They are examined with reference to reprocessing spent nuclear fuel and application is detailed both for liquid and dry processes. (author)

  15. Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium in fast breeder reactors (FBRs) fuel reprocessing

    International Nuclear Information System (INIS)

    Govindan, P.; Palamalai, A.; Vijayan, K.S.; Subba Rao, R.V.; Venkataraman, M.; Natarajan, R.

    2013-01-01

    Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium is developed for the recovery of uranium and plutonium present in spent fuel of fast breeder reactors (FBRs). Effect of pH on the solubility of carbonates of uranium and plutonium in ammonium carbonate medium is studied. Effect of mole ratios of uranium and plutonium as a function of uranium and plutonium concentration at pH 8.0-8.5 for effective separation of uranium and plutonium to each other is studied. Feasibility of reconversion of plutonium in carbonate medium is also studied. The studies indicate that uranium is selectively precipitated as AUC at pH 8.0-8.5 by adding ammonium carbonate solution leaving plutonium in the filtrate. Plutonium in the filtrate after acidified with concentrated nitric acid could also be precipitated as carbonate at pH 6.5-7.0 by adding ammonium carbonate solution. A flow sheet is proposed and evaluated for partitioning and reconversion of uranium and plutonium simultaneously in the FBR fuel reprocessing. (author)

  16. Chemical Disposition of Plutonium in Hanford Site Tank Wastes

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-07

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

  17. Preparation and application of potassium and sodium titanate for removal of plutonium from basic solution

    International Nuclear Information System (INIS)

    Patil, Prashant; Pathak, Sachin S.; Pius, I.C.; Mukerjee, S.K.

    2014-01-01

    In PUREX process, after extraction and stripping of uranium and plutonium, the extractant, tributyl phosphate is usually washed with sodium carbonate solution before reuse for the removal of radiolytic/hydrolytic degradation products of TBP and small amounts of HNO 3 , uranium and plutonium goes into aqueous phase during carbonate washings. Partial neutralization of carbonate by the acid converts it to bicarbonate. Removal of plutonium from such sodium carbonate/bicarbonate streams facilitates their disposal. In the present work, studies were carried out to prepare inorganic ion-exchangers such as potassium and sodium titanates for their application as ion-exchange material. It is essential to prepare these materials in granular form to obtain good liquid flow property for ion exchange column operations, however, it is also important that the final product is having good surface area and porosity so that they may exhibit good ion exchange capacity

  18. Effect of compositional variation in plutonium on process shielding design

    Energy Technology Data Exchange (ETDEWEB)

    Brown, T.H.

    1997-11-01

    Radiation dose rate from plutonium with high {sup 239}Pu content varies with initial nuclidic content, radioactive decay time, and impurity elemental content. The two idealized states of old plutonium and clean plutonium, whose initial compositions are given, provide approximate upper and lower bounds on dose rate variation. Whole-body dose rates were calculated for the two composition states, using unshielded and shielded plutonium spheres of varying density. The dose rates from these variable density spheres are similar to those from expanded plutonium configurations encountered during processing. The dose location of 40 cm from the sphere center is representative of operator standoff for direct handling of plutonium inside a glove box. The results have shielding implications for glove boxes with only structurally inherent shielding, especially for processing of old plutonium in an expanded configuration. Further reduction in total dose rate by using lead to reduce photon dose rate is shown for two density cases representing compact and expanded plutonium configurations.

  19. Effect of compositional variation in plutonium on process shielding design

    International Nuclear Information System (INIS)

    Brown, T.H.

    1997-11-01

    Radiation dose rate from plutonium with high 239 Pu content varies with initial nuclidic content, radioactive decay time, and impurity elemental content. The two idealized states of old plutonium and clean plutonium, whose initial compositions are given, provide approximate upper and lower bounds on dose rate variation. Whole-body dose rates were calculated for the two composition states, using unshielded and shielded plutonium spheres of varying density. The dose rates from these variable density spheres are similar to those from expanded plutonium configurations encountered during processing. The dose location of 40 cm from the sphere center is representative of operator standoff for direct handling of plutonium inside a glove box. The results have shielding implications for glove boxes with only structurally inherent shielding, especially for processing of old plutonium in an expanded configuration. Further reduction in total dose rate by using lead to reduce photon dose rate is shown for two density cases representing compact and expanded plutonium configurations

  20. Studies on removal of plutonium from oxalic acid containing hydrochloric acid solutions

    Energy Technology Data Exchange (ETDEWEB)

    Ghadse, D R; Noronha, D M; Joshi, A R [Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    Solution containing hydrochloric acid, oxalic acid and considerable quantities of plutonium may be generated while recycling of scrap produced during the metallic fuel fabrication. Plutonium from such waste is normally recovered by anion exchange method after the destruction of oxalic acid using suitable oxidising agent. Solvent extraction and ion exchange methods are being explored in this laboratory for recovery of Pu from oxalic acid containing HCl solutions without prior destruction of oxalic acid. This paper describes the results on the determination of distribution ratios for extraction of Pu(IV) from hydrochloric acid using Aliquot-336 or HDEHP under varying experimental conditions. (author). 5 refs., 5 tabs.

  1. 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.

  2. 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.

  3. 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...

  4. Chromium in aqueous nitrate plutonium process streams: Corrosion of 316 stainless steel and chromium speciation

    International Nuclear Information System (INIS)

    Smith, W.H.; Purdy, G.M.

    1995-01-01

    This study was undertaken to determine if chromium(+6) could exist in plutonium process solutions under normal operating conditions. Four individual reactions were studied: the rate of dissolution of stainless steel, which is the principal source of chromium in process solutions; the rate of oxidation of chromium(+3) to chromium(+6) by nitric acid; and the reduction of chromium(+6) back to chromium(+3) by reaction with stainless steel and with oxalic acid. The stainless steel corrosion rate was found to increase with increasing nitric acid concentration, increasing hydrofluoric acid concentration, and increasing temperature. Oxidation of chromium(+3) to chromium(+6) was negligible at room temperature and only became significant in hot concentrated nitric acid. The rate of reduction of chromium(+6) back to chromium(+3) by reaction with stainless steel or oxalic acid was found to be much greater than the rate of the reverse oxidation reaction. Based on these findings and taking into account normal operating conditions, it was determined that although there would be considerable chromium in plutonium process streams it would rarely be found in the (+6) oxidation state and would not exist in the (+6) state in the final process waste solutions

  5. Chromium in aqueous nitrate plutonium process streams: Corrosion of 316 stainless steel and chromium speciation

    International Nuclear Information System (INIS)

    Smith, W.H.; Purdy, G.

    1995-01-01

    This study was undertaken to determine if chromium +6 could exist in plutonium process solutions under normal operating conditions. Four individual reactions were studied: the rate of dissolution of stainless steel, which is the principal source of chromium in process solutions; the rate of oxidation of chromium +3 to chromium +6 by nitric. acid; and the reduction of chromium +6 back to chromium +3 by reaction with stainless steel and with oxalic acid. The stainless steel corrosion rate was found to increase with increasing nitric acid concentration, increasing hydrofluoric acid concentration, and increasing temperature. Oxidation of chromium +3 to chromium +6 was negligible at room temperature and only became significant in hot concentrated nitric acid. The rate of reduction of chromium +6 back to chromium +3 by reaction with stainless steel or oxalic acid was found to be much greater than the rate of the reverse oxidation reaction. Based on these findings and taking into account normal operating conditions, it was determined that although there would be considerable chromium in plutonium process streams it would rarely be found in the +6 oxidation state and would not exist in the +6 state in the final process waste solutions

  6. Plutonium solution in concentration range from 8 to 17 G/liter

    Energy Technology Data Exchange (ETDEWEB)

    Rothe, R.E.

    1997-06-01

    This paper very briefly discusses the need for a fundamental criticality study of low concentrations of plutonium solutions. Examples of the occurrence of such solutions, which are characteristic of waste, are cited. Due to the prevalence of decontaminating and decommissioning activities, low concentration solutions are expected to become an important concern. Technical deficiencies in previous calculations are also discussed as a reason for performing low concentration criticality studies. 3 refs.

  7. Plutonium solution in concentration range from 8 to 17 G/liter

    International Nuclear Information System (INIS)

    Rothe, R.E.

    1997-01-01

    This paper very briefly discusses the need for a fundamental criticality study of low concentrations of plutonium solutions. Examples of the occurrence of such solutions, which are characteristic of waste, are cited. Due to the prevalence of decontaminating and decommissioning activities, low concentration solutions are expected to become an important concern. Technical deficiencies in previous calculations are also discussed as a reason for performing low concentration criticality studies. 3 refs

  8. Organic components and plutonium and americium state in soils and soil solutions

    International Nuclear Information System (INIS)

    Sokolik, G.A.; Ovsyannikova, S.V.; Kimlenko, I.M.

    2002-01-01

    The fraction composition of humus substances of different type soils and soil solutions have been studied. A distribution of Pu 239, 240 and Am 241 between humus substances fractions of different dispersity and mobility in soil-vegetation cover has been established. It was shown that humus of organic soils fixes plutonium and americium in soil medium in greater extent than humus of mineral soils. That leads to lower migration ability of radionuclides in organic soils. The lower ability of americium to form difficultly soluble organic and organic-mineral complexes and predomination of its anion complexes in soil solutions may be a reason of higher mobility and biological availability of americium in comparison to plutonium during soil-plant transfer (authors)

  9. Potentiometric determination of free nitric-acid in trilaurylamine solutions containing plutonium nitrate; Dosage potentiometrique de l'acidite nitrique libre dans les solutions organiques de trilaurylamine

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    A potentiometric method of determination of the free nitric acid in trilaurylamine solutions containing plutonium or thorium nitrates is described. The potentiometric titration is carried out in a mixture of benzene and 1,2-dichloro ethane with a standard solution of trilaurylamine as the titrant. When thorium nitrate is present the metal complex is not dissociated then the titration has a single end-point. In the case of plutonium nitrate the partial dissociation of the plutonium complex corresponds to a second point. The experimental error in duplicate analyses of 50 samples is about 1 per cent for free acid concentrations in the range of 0,03 to 0,1 N and plutonium concentrations between 1 to 5 g/l. (authors) [French] Une methode potentiometrique de dosage de l'acidite nitrique libre dans les solutions de trilaurylamine contenant un complexe de plutonium ou de thorium est decrite. La potentiometrie est effectuee en prenant comme base titrante la trilaurylamine et comme milieu de dilution un melange de benzene et de 1,2 dichloroethane. Dans le cas du thorium, le complexe organometallique n'est pas deplace et la courbe de titrage presente un seul point d'inflexion. Dans le cas du plutonium le complexe est partiellement dissocie ce qui correspond a un second saut de potentiel. La moyenne des erreurs experimentales sur 50 echantillons doses a ete d'environ {+-} 1 pour cent sur l'acide libre. Les solutions experimentees contenaient de 0,03 a 0,1 N en acide et de 1 a 5 g/l en plutonium. (auteurs)

  10. On-line monitoring of low-level plutonium concentrations

    International Nuclear Information System (INIS)

    Hofstetter, K.J.; Huff, G.A.; Rebagay, T.V.

    1979-10-01

    An on-line monitor has been developed to assay plutonium in nitric acid solutions. The performance of the monitor has been assessed by a laboratory experimentation program using solutions with plutonium concentrations from 0.1 to 10 g/l. These conditions are typical of the plutonium solutions in an input stream to a plutonium-purification cycle in a reprocessing plant following uranium/plutonium partitioning. The monitoring system can be fully automated and shows great promise for detecting and quantifying plutonium in situ, thus minimizing the reliance on traditional sampling and laboratory-analysis techniques. The total concentration and isotopic abundance of plutonium are determined by measuring the absolute intensities of the low-energy gamma rays characteristics of 238 Pu, 239 Pu, and 240 Pu nuclides by direct gamma-ray spectroscopy and computer analysis of the spectral data. The addition of a monitoring system of this type to the input stream of a plutonium-purification cycle along with other suitable monitors on the waste streams and on the product stream provides the basis for a near real-time materials control and inventory system. Results of the laboratory-evaluation program employing plutonium in solutions with isotopic compositions typical of those involved in processing light water reactor fuels are presented. The detailed design of a monitoring cell and detection system is given. The precision and accuracy of the results relative to those measured by mass spectrometry and controlled potential coulometry are also summarized

  11. Interpretation of criticality experiments on homogeneous solutions of plutonium and uranium; Interpretation des experiences de criticite sur des solutions homogenes de plutonium et d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Ithurralde, M F; Kremser, J; Leclerc, J; Lombard, Ch; Moreau, J; Robin, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Criticality experiments on solutions of fissionable materials have been carried out in tanks of various geometries (cylinder, isolated annular cylinder, interacting annular cylinders); the reflexion conditions have also been varied (without reflection, semi-reflection and total reflexion by water). The range of the studied concentrations is rather large (18,8 to 104 gms/liter). The interpretation of these experiments has been undertaken in order to resolve the problems of the industrial use of homogeneous plutonium and uranium solutions. Several methods the fields of application of which are different have been used: diffusion method, transport method and Monte-Carlo method. (authors) [French] Des experiences critiques sur des solutions de matieres fissiles ont ete faites dans des cuves de diverses geometries (cylindre, cylindre annulaire isole, cylindre annulaire en interaction), les conditions de reflexion ont ete egalement variees (sans reflexion, semi reflexion et reflexion totale par l'eau). La gamme des concentrations etudiees est assez etendue (18,8 a 104 g/l ). L'interpretation de ces experiences a ete entreprise dans le but de pouvoir resoudre les problemes poses par l'emploi industriel de solutions homogenes de plutonium et d'uranium, plusieurs methodes dont les domaines d'application sont differents ont ete employees: methode de diffusion, methode de transport, methode de Monte-Carlo. (auteurs)

  12. 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

  13. 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

  14. Decontamination and demolition of a former plutonium processing facility's process exhaust system, firescreen, and filter plenum buildings

    International Nuclear Information System (INIS)

    LaFrate, P.J. Jr.; Stout, D.S.; Elliott, J.W.

    1996-01-01

    The Los Alamos National Laboratory (LANL) Decommissioning Project has decontaminated, demolished, and decommissioned a process exhaust system, two filter plenum buildings, and a firescreen plenum structure at Technical Area 21 (TA-2 1). The project began in August 1995 and was completed in January 1996. These high-efficiency particulate air (HEPA) filter plenums and associated ventilation ductwork provided process exhaust to fume hoods and glove boxes in TA-21 Buildings 2 through 5 when these buildings were active plutonium and uranium processing and research facilities. This paper summarizes the history of TA-21 plutonium and uranium processing and research activities and provides a detailed discussion of integrated work process controls, characterize-as-you-go methodology, unique engineering controls, decontamination techniques, demolition methodology, waste minimization, and volume reduction. Also presented in detail are the challenges facing the LANL Decommissioning Project to safely and economically decontaminate and demolish surplus facilities and the unique solutions to tough problems. This paper also shows the effectiveness of the integrated work package concept to control work through all phases

  15. Decontamination and demolition of a former plutonium processing facility's process exhaust system, firescreen, and filter plenum buildings

    International Nuclear Information System (INIS)

    LaFrate, P.J. Jr.; Stout, D.S.; Elliott, J.W.

    1996-01-01

    The Los Alamos National Laboratory (LANL) Decommissioning Project has decontaminated, demolished, and decommissioned a process exhaust system, two filter plenum buildings, and a firescreen plenum structure at Technical Area 21 (TA-21). The project began in August 1995 and was completed in January 1996. These high-efficiency particulate air (HEPA) filter plenums and associated ventilation ductwork provided process exhaust to fume hoods and glove boxes in TA-21 Buildings 2 through 5 when these buildings were active plutonium and uranium processing and research facilities. This paper summarizes the history of TA-21 plutonium and uranium processing and research activities and provides a detailed discussion of integrated work process controls, characterize-as-you-go methodology, unique engineering controls, decontamination techniques, demolition methodology, waste minimization, and volume reduction. Also presented in detail are the challenges facing the LANL Decommissioning Project to safely and economically decontaminate and demolish surplus facilities and the unique solutions to tough problems. This paper also shows the effectiveness of the integrated work package concept to control work through all phases

  16. 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.

  17. Simultaneous determination of uranium and plutonium in dissolver solution of irradiated fuel, using ID-TIMS. IRP-11

    International Nuclear Information System (INIS)

    Shah, Raju; Sasi Bhushan, K.; Govindan, R.; Alamelu, D.; Khodade, P.S.; Aggarwal, S.K.

    2007-01-01

    A simple sample preparation and simultaneous analysis method to determine uranium and plutonium from dissolver solution, employing the technique of Isotope Dilution Mass spectrometry has been demonstrated. The method used, co-elusion of Uranium and Plutonium from anion exchanger column after initial elution of major part of uranium in 1:5 HNO 3 in order to reduce the initial U/Pu ratio from 1000 to about 100-200 in the co-eluted fraction. Due to the availability of variable multi-collector system, different Faraday cups were adjusted to collect the different ion intensities corresponding to the different masses, during the simultaneous analysis of Uranium and Plutonium, loaded on Re double filament assembly. 233 U and PR grade Plutonium were used as spikes to determine Uranium and Plutonium from dissolver solution of irradiated fuel from research reactor. The possibility of getting the isotopic composition of uranium from the simultaneous analysis of co-eluted purified fraction of U and Pu from spiked aliquots is also explained. (author)

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

    International Nuclear Information System (INIS)

    Rudisill, T.S.

    1999-01-01

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

  19. Plutonium production story at the Hanford site: processes and facilities history

    Energy Technology Data Exchange (ETDEWEB)

    Gerber, M.S., Westinghouse Hanford

    1996-06-20

    This document tells the history of the actual plutonium production process at the Hanford Site. It contains five major sections: Fuel Fabrication Processes, Irradiation of Nuclear Fuel, Spent Fuel Handling, Radiochemical Reprocessing of Irradiated Fuel, and Plutonium Finishing Operations. Within each section the story of the earliest operations is told, along with changes over time until the end of operations. Chemical and physical processes are described, along with the facilities where these processes were carried out. This document is a processes and facilities history. It does not deal with the waste products of plutonium production.

  20. 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

  1. 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

  2. Study on the process variables in the anion exchange plutonium separation process

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, D T

    1957-11-15

    This report discusses the study of the process variables in the Anion Exchange Process Pilot Plant for the separation of plutonium from irradiated uranium. Variables associated with the feed, wash and elution cycles were studied with the aim of improving the quality of the final plutonium product, reduce cycling time and reagent requirements, and also to obtain data for prediction of resin column behaviour under various feed conditions. A cation resin column and a silica gel column were installed in the system and these were studied for plutonium recovery and product quality. The product obtained from the plant was acceptable in all the impurities except the associated gamma activity which was too high for easy product handling. (author)

  3. Optimum Condition for Plutonium Electrodeposition Process in Radiochemistry and Environment Laboratory, Nuclear Malaysia

    International Nuclear Information System (INIS)

    Yii, Mei-Wo; Abdullah Siddiqi Ismail

    2014-01-01

    Determination of alpha emitting plutonium radionuclides such as Pu-238, Pu-239 and Pu-240 concentrations inside a sample require lots of radiochemistry purification process to separate them from other interfering alpha emitters. These pure isotopes are then been electrodeposited onto a stainless steel disc and quantified using alpha spectrometry counter. In Radiochemistry and Environment Laboratory (RAS), Nuclear Malaysia, the quantification is done by comparing these isotopes with the recovery of known amount plutonium tracer, Pu-242, that been added into the sample prior analysis. This study been conducted to find the optimum conditions for the electrolysis process used at RAS. Four variable parameters that may interfere the percentage recovery of tracer hence the current, cathode to anode distance, pH and electrolysis duration had been identify and studied. Study was carry out using Pu-242 standard solution and the deposition disc was counted using Zinc Sulphite (silver) counter. Studies outcome suggested that the optimum conditions to reduce plutonium ion happens at 1-1.1 ampere of current, 3-5 mm of electrodes distance, pH 2.2-2.5 and a minimal electrolysis duration of 2 hours. (author)

  4. The study of reductive reextraction of plutonium in the Purex process

    International Nuclear Information System (INIS)

    Poczynajlo, A.

    1985-01-01

    The methods of separation of U and Pu in the Purex process and the thermodynamic and kinetic properties of Pu(4) reductants are discussed. The kinetic equation of the process of reductive reextraction of plutonium for the first order reaction with respect to Pu(4) is derived. The kinetics of plutonium reextraction with the use of uranium (4), ascorbic acid and other reductants has been studied. The necessity of application of the stoichiometric excess of reductant has been explained by simultaneously occured reoxidation process of plutonium. The method of calculation of the steady- state plutonium concentration profiles has been elaborated for counter-current separation of U and Pu in multistage contactor. 90 refs., 20 tabs., 29 figs. (author)

  5. Study of the influence of radiolysis on the stability of plutonium III. Application to a heterogeneous medium formed by a nitric solution of ferrous ions and an organic solution of trilauryl-ammonium nitrate

    International Nuclear Information System (INIS)

    Fourmaux, J.M.

    1980-01-01

    The objective of this research thesis is to study the behaviour of plutonium 238 in media which are commonly used to isolate it from other elements such as neptunium and fission products created during the neutron irradiation of the neptunium 237 isotope. As plutonium 238 purification processes are all based on redox reaction, it is essential to know the influence of radiolysis on the redox behaviour, and on the distribution coefficients of this isotope in solutions used during its separation from the neptunium 237 isotope. Therefore, it is necessary to study the influence of radiolysis on the stability of plutonium with an oxidation III level. As this extraction is performed by an organic solvent (trilauryl-ammonium nitrate), this study addresses the behaviour of plutonium in an emulsion formed by this solvent and the nitric aqueous solution previously adjusted in terms of Fe 2+ ions. After a brief recall of bibliographical generalities related to radiolysis, the author presents and comments the Nernst law in the case of a two-phase system (emulsion), and reports the use of this law to obtain the plutonium potential-distribution coefficient relationship. The last part reports experimental data

  6. 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

  7. 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.

  8. 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

  9. Chromium in aqueous nitrate plutonium process streams: Corrosion of 316 stainless steel and chromium speciation

    International Nuclear Information System (INIS)

    Smith, W.H.; Purdy, G.

    1994-01-01

    According to the measurements made in this study, the only situation in which chromium (+6) could exist in a plutonium process solution is one in which a feed containing chromium is dissolved in a glass pot dissolver in high nitric acid concentration and at high temperature. But when the resulting feed is prepared for ion exchange, the chemical treatment reduces chromium to the +3 state. Any solution being processed through the evaporator will only contain chromium in the +3 state and any chromium salts remaining in the evaporator bottoms will be chromium +3 salts

  10. Aqueous recovery of plutonium from pyrochemical processing residues

    International Nuclear Information System (INIS)

    Gray, L.W.; Gray, J.H.

    1984-01-01

    Pyrochemical processes provide rapid methods to reclaim plutonium from scrap residues. Frequently, however, these processes yield an impure plutonium product and waste residues that are contaminated with actinides and are therefore nondiscardable. The Savannah River Laboratory and Plant and the Rocky Flats Plant are jointly developing new processes using both pyrochemistry and aqueous chemistry to generate pure product and discardable waste. An example of residue being treated is that from the molten salt extraction (MSE), a mixture of NaCl, KCl, MgCl 2 , PuCl 3 , AmCl 3 , PuO 2 , and Pu 0 . This mixture is scrubbed with molten aluminum containing a small amount of magnesium to produce a nonhomogeneous Al-Pu-Am-Mg alloy. This process, which rejects most of the NaCl-KCl-MgCl 2 salts, results in a product easily dissolved in 6M HNO 3 -0.1M HF. Any residual chloride in the product is removed by precipitation with Hg(I) followed by centrifuging. Plutonium and americium are then separated by the standard Purex process. The americium, initially diverted to the solvent extraction waste stream, can either be recovered or sent to waste

  11. Electrochemical studies of plutonium(IV) complexes in aqueous nitrate solutions

    International Nuclear Information System (INIS)

    Kim, Seong-Yun; Asakura, Toshihide; Morita, Yasuji

    2005-01-01

    Electrochemistry has been used to investigate the behavior of plutonium (IV) in 1-7 M HNO 3 solutions. These Pu(IV) complexes were found to be reduced quasi-reversibly to Pu(III) species. The formal redox potentials (E 0 ) for Pu(IV)/Pu(III) couples were determined to be +0.721, +0.712, +0.706, +0.705, +0.704, 0.694, and +0.696 V (vs. Ag/AgCl(SSE)) for Pu(IV) complexes in 1, 2, 3, 4, 5, 6, 7 M HNO 3 solutions, respectively. These results indicate that the reduction product of Pu(IV) is Pu(III), which is considerably stable in HNO 3 solution. (author)

  12. Critical and subcritical parameters of the system simulating plutonium metal dissolution

    International Nuclear Information System (INIS)

    Vasilev, Yury Yu.; Ryazanov, Boris G.; Sviridov, Victor I.; Mozhayeva, Lubov I.

    2003-01-01

    Dissolution of plutonium metal was simulated using the Monte Carlo computer code to calculate criticality safety limits for the process. Calculations were made for the constant masses of plutonium charged to the dissolving vessel considering distribution of plutonium in metal and solution phases. Critical parameters and limits were calculated as a function of dissolving vessel volume and plutonium metal mass. 240 Pu content was assumed to be from 0% to 10% (mass). Critical parameters were evaluated for the system with a water reflector. Results of this paper may be used in the designing process equipment for plutonium metal dissolution. (author)

  13. Plutonium solution analyzer

    International Nuclear Information System (INIS)

    Burns, D.A.

    1994-09-01

    A fully automated analyzer has been developed for plutonium solutions. It was assembled from several commercially available modules, is based upon segmented flow analysis, and exhibits precision about an order of magnitude better than commercial units (0.5%-O.05% RSD). The system was designed to accept unmeasured, untreated liquid samples in the concentration range 40-240 g/L and produce a report with sample identification, sample concentrations, and an abundance of statistics. Optional hydraulics can accommodate samples in the concentration range 0.4-4.0 g/L. Operating at a typical rate of 30 to 40 samples per hour, it consumes only 0.074 mL of each sample and standard, and generates waste at the rate of about 1.5 mL per minute. No radioactive material passes through its multichannel peristaltic pump (which remains outside the glovebox, uncontaminated) but rather is handled by a 6-port, 2-position chromatography-type loop valve. An accompanying computer is programmed in QuickBASIC 4.5 to provide both instrument control and data reduction. The program is truly user-friendly and communication between operator and instrument is via computer screen displays and keyboard. Two important issues which have been addressed are waste minimization and operator safety (the analyzer can run in the absence of an operator, once its autosampler has been loaded)

  14. Plutonium solution analyzer

    Energy Technology Data Exchange (ETDEWEB)

    Burns, D.A.

    1994-09-01

    A fully automated analyzer has been developed for plutonium solutions. It was assembled from several commercially available modules, is based upon segmented flow analysis, and exhibits precision about an order of magnitude better than commercial units (0.5%-O.05% RSD). The system was designed to accept unmeasured, untreated liquid samples in the concentration range 40-240 g/L and produce a report with sample identification, sample concentrations, and an abundance of statistics. Optional hydraulics can accommodate samples in the concentration range 0.4-4.0 g/L. Operating at a typical rate of 30 to 40 samples per hour, it consumes only 0.074 mL of each sample and standard, and generates waste at the rate of about 1.5 mL per minute. No radioactive material passes through its multichannel peristaltic pump (which remains outside the glovebox, uncontaminated) but rather is handled by a 6-port, 2-position chromatography-type loop valve. An accompanying computer is programmed in QuickBASIC 4.5 to provide both instrument control and data reduction. The program is truly user-friendly and communication between operator and instrument is via computer screen displays and keyboard. Two important issues which have been addressed are waste minimization and operator safety (the analyzer can run in the absence of an operator, once its autosampler has been loaded).

  15. Variations of uranium and plutonium coprocessing as proliferation-resistant alternatives to the classical purex process

    International Nuclear Information System (INIS)

    Buckham, J.A.; Sumner, W.B.

    1979-08-01

    Evaluation of these alternatives for processing LWR fuel has led to the following conclusions: (1) None of the alternaives provide a pure, technical solution which completely eliminates the potential for proliferation of nuclear weapons by utilizing plutonium from the light water reactors. (2) The heat spike alternative appears feasible and provides the most effective method of rendering the LWR plutonim unattractive for weapons use. (3) The low-DF process alternate would require demonstration to: (a) determine the reliability of the in-cell recycle streams which are used to prevent reversion of the process for purification of plutonium, and (b) verify the fission product decontamination factors. (4) The alternates evaluated have no significant impacts on the design of waste treatment facilities, although the required capacities of high-level solid waste processing and high-level liquid waste storage can be significantly altered. (5) The impact of these alternate processes on fuel fabrication and other aspects of the fuel cycle requires additional evaluation

  16. Co-precipitation of plutonium(IV) and americium(III) from nitric acid-oxalic acid solutions with bismuth oxalate

    International Nuclear Information System (INIS)

    Pius, I.C.; Noronha, D.M.; Chaudhury, Satyajeet

    2017-01-01

    Co-precipitation of plutonium and americium from nitric acid-oxalic acid solutions with bismuth oxalate has been investigated for the removal of these long lived α-active nuclides from waste solutions. Effect of concentration of bismuth and oxalic acid on the co-precipitation of Pu(IV) from 3 M HNO_3 has been investigated. Similar experiments were also carried out from 3.75 M HNO_3 on co-precipitation of Am(III) to optimize the conditions of precipitation. Strong co-precipitation of Pu(IV) and Am(III) with bismuth oxalate indicate feasibility of treatment of plutonium and americium bearing waste solutions. (author)

  17. Chemical behaviour of plutonium in aqueous chloride solutions

    International Nuclear Information System (INIS)

    Bueppelmann, K.; Kim, J.I.

    1988-06-01

    The chemical behaviour of Plutonium has been investigated in concentrated NaCl solutions in the neutral pH range. The α-radiation induced radiolysis reactions oxidize the Cl - -ion to Cl 2 , HClO, ClO - and other species, which produce a strongly oxidizing medium. Under these conditions the Pu ions of lower oxidation states are readily oxidized to Pu(VI), which then undergo depending on the pH of the solution, various chemical reactions to produce PuO 2 Cl n , PuO 2 (ClO) m or PuO 2 (OH) x species. In addition to primary radiolysis reactions taking place in NaCl solutions, the reactions leading to the PuO 2 (Cl) n and PuO 2 (ClO) m species have been characterized and quantified systematically by spectroscopic and thermodynamic evaluation. The redox and complexation reactions of Pu ions under varying NaCl concentration, specific α-activity and pH are discussed. (orig.) [de

  18. 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)

  19. 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.

  20. A portable concentrator for processing plutonium

    International Nuclear Information System (INIS)

    Chamberlain, D.B.; Conner, C.; Chen, L.

    1995-01-01

    A horizontal, agitated film concentrator designed to concentrate liquid streams to a high solid content slurry is briefly described. The Rototherm unit is being studied for use at US Department of Energy facilities to handle large quantities of aqueous plutonium solutions. Capabilities for evaporating more than 98% of the water present in a single pass have been demonstrated. Decontamination factors of 10 6 to 10 7 are expected. The unit may also be useful for recycling aqueous waste treatment reagents from the decontamination of gaseous diffusion plants

  1. Electrochemical preparation of uranium and plutonium measuring probes for alpha spectroscopy from organic solutions

    International Nuclear Information System (INIS)

    Gruner, W.; Beutmann, A.

    1980-01-01

    A method for preparation of uranium and plutonium measuring probes for α-spectrometry is described. The method is based on electrodeposition from isopropanol and especially from ethanol and methanol solution. It was shown that a definite additions of a little amount of water lead to an increase of the deposition rate. It is possible to reach a 100% deposition in ethanol after an electrolysis time of 3 minutes for uranium and 30 minutes for plutonium with voltages of 150-200 V. (author)

  2. 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

  3. Standard test methods for chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade plutonium nitrate solutions

    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, spectrochemical, nuclear, and radiochemical analysis of nuclear-grade plutonium nitrate solutions to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Plutonium by Controlled-Potential Coulometry Plutonium by Amperometric Titration with Iron(II) Plutonium by Diode Array Spectrophotometry Free Acid by Titration in an Oxalate Solution 8 to 15 Free Acid by Iodate Precipitation-Potentiometric Titration Test Method 16 to 22 Uranium by Arsenazo I Spectrophotometric Test Method 23 to 33 Thorium by Thorin Spectrophotometric Test Method 34 to 42 Iron by 1,10-Phenanthroline Spectrophotometric Test Method 43 to 50 Impurities by ICP-AES Chloride by Thiocyanate Spectrophotometric Test Method 51 to 58 Fluoride by Distillation-Spectrophotometric Test Method 59 to 66 Sulfate by Barium Sulfate Turbidimetric Test Method 67 to 74 Isotopic Composition by Mass Spectrom...

  4. In situ observation of plutonium transfer processes in the marine environment

    International Nuclear Information System (INIS)

    Guary, J.-C.; Fraizier, Andre

    1975-09-01

    A preliminary observation of plutonium transfer processes in the marine environment was carried out and showed that concentration of the radionuclide was lower when marine organisms stood at a higher trophic level. This observation supplemented by an investigation on contamination pathways showed that plutonium was not concentrated along the food chain and its uptake occured preferentially by direct contact of species with seawater, a process chiefly affecting producers and primary consumers. It appeared that the marine sediment was not a significant vector of plutonium transfer in burrowing species [fr

  5. 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)

  6. Development of a freeze-drying process of waste-solution, 2

    International Nuclear Information System (INIS)

    Kondo, Isao; Kawasaki, Takeshi

    1988-01-01

    The waste solution treatment process in Plutonium Conversion Development Facility (PCDF) consists of Evaporation-Condensation and Neutrazation-Agglometation-Precipitation process, which produces the distillate as recovered acid at first step and separates Pu-U element from condenced solution at second step. This process needs many stages to get high decontamination efficiency and then the Evaporator is in very corrosive state because the nitric acid solution is heated over 100 degrees C to be evaporated. So, in PCDF, it was started the development of Freeze-Drying process to waste solution treatment. This process is suitable for a little quantity of the solution including nitric acid as produced in the Microwave Heating method. Moreover the process has high decontamination efficiency and has good performance of equipment. The result of the cold test of Freeze-Drying process with nitric acid is discribed in this paper. (author)

  7. 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

  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. Extraction of plutonium and uranium from oxalate bearing solutions using phosphonic acid

    International Nuclear Information System (INIS)

    Godbole, A.G.; Mapara, P.M.; Swarup, Rajendra

    1995-01-01

    A feasibility study on the solvent extraction of plutonium and uranium from solutions containing oxalic and nitric acids using a phosphonic acid extractant (PC88A) was made to explore the possibility of recovering Pu from these solutions. Batch experiments on the extraction of Pu(IV) and U(VI) under different parameters were carried out using PC88A in dodecane. The results indicated that Pu could be extracted quantitatively by PC88A from these solutions. A good separation of Pu from U could be achieved at higher temperatures. (author). 6 refs., 3 tabs

  10. Recent studies of uranium and plutonium chemistry in alkaline radioactive waste solutions

    International Nuclear Information System (INIS)

    King, William D.; Wilmarth, William R.; Hobbs, David T.; Edwards, Thomas B.

    2008-01-01

    Solubility studies of uranium and plutonium in a caustic, radioactive Savannah River Site tank waste solution revealed the existence of uranium supersaturation in the as-received sample. Comparison of the results to predictions generated from previously published models for solubility in these waste types revealed that the U model poorly predicts solubility while Pu model predictions are quite consistent with experimental observations. Separate studies using simulated Savannah River Site evaporator feed solution revealed that the known formation of sodium aluminosilicate solids in waste evaporators can promote rapid precipitation of uranium from supersaturated solutions

  11. Recovery of plutonium from electrorefining anode heels at Savannah River

    International Nuclear Information System (INIS)

    Gray, J.H.; Gray, L.W.; Karraker, D.G.

    1987-03-01

    In a joint effort, the Savannah River Laboratory (SRL), Savannah River Plant (SRP), and the Rocky Flats Plant (RFP) have developed two processes to recover plutonium from electrorefining anode heel residues. Aqueous dissolution of anode heel metal was demonstrated at SRL on a laboratory scale and on a larger pilot scale using either sulfamic acid or nitric acid-hydrazine-fluoride solutions. This direct anode heel metal dissolution requires the use of a geometrically favorable dissolver. The second process developed involves first diluting the plutonium in the anode heel residues by alloying with aluminum. The alloyed anode heel plutonium can then be dissolved using a nitric acid-fluoride-mercury(II) solution in large non-geometrically favorable equipment where nuclear safety is ensured by concentration control

  12. Electronic spectra of plutonium ions in nitric acid and in lithium nitrate solutions

    International Nuclear Information System (INIS)

    Mekhail, F.M.

    1987-01-01

    The absorption spectra of plutonium ions in nitric acid have been described. There is a characteristic change in the absorption spectra of Pu v in lithium nitrate solutions. In 2 M-lithium nitrate a new peak at 969 nm and high absorption at 1200 nm are noticed. A decrease in the absorption by about 20% and the appearance of a new shoulder at 1120 nm in 6 M-lithium nitrate are found. There is no change in the spectrum in 4 M-lithium nitrate. The absorption spectra of plutonium ions in the spectral range 200 - 400 nm are interesting. All plutonium ions have an intense band in the region 250 - 260 nm as well as a less intense and rather diffuse band at 320 - 330 nm in lithium nitrate solutions the sharp band at 250 - 260 nm has disappeared. This suggests that this band is very sensitive to the environmental field. The band is probably produced by 5 F q → 5 f q-1 6 d transition as well as electron transfer. It is believed that the spectrum of Pu V at pH 6.5 represents the hydrolysis product Pu O 2 (O H). 9 fig., 4 tab

  13. 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

  14. Decontamination of plutonium-contaminated surfaces; Essais de decontamination des surfaces contaminees par du plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Bertrand, J; Clouet d' Orval, Ch; Tachon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The measure of the neutron distribution in the core of 'Proserpine', by means of activation detectors, requires no contact between the plutonium sulfate solution and the detectors. These detectors are put into PVC or polyethylene bags. This report describes the process used to decontaminate these bags. A washing by nitric acid followed by coating with plexiglass is kept, with this process we have no contamination on the detectors. (author) [French] La mesure de la distribution de neutrons par detecteurs a activation dans le coeur de Proserpine exige de proteger ces detecteurs contre tout contact avec la solution de plutonium. Les detecteurs sont places dans des gaines en polyvinyle ou en polyethylene. Ce rapport decrit le procede utilise pour decontaminer ces gaines. On a retenu un lavage a l'acide nitrique suivi du revetement d'une meme couche de plexiglass, ce qui permet d'eviter la contamination des detecteurs. (auteur)

  15. EXPECTED IMPACT OF HANFORD PROCESSING ORGANICS OF PLUTONIUM DURING TANK WASTE SLUDGE RETRIEVAL

    International Nuclear Information System (INIS)

    TROYER, G.L.; WINTERS, W.I.

    2004-01-01

    This document evaluates the potential for extracting plutonium from Hanford waste tanks into residual organic solvents and how this process may have an impact on criticality specifications during the retrieval of wastes. The two controlling factors for concentrating plutonium are the solubility of the plutonium in the wastes and the extraction efficiency of the potential organic extractants that may be found in these wastes. Residual Hanford tank sludges contain plutonium in solid forms that are expected to be primarily insoluble Pu(IV) hydroxides. Evaluation of thermodynamic Pourbaix diagrams, documentation on solubility studies of various components in waste tank matrices, and actual analysis of plutonium in tank supernates all indicate that the solubility of Pu in the alkaline waste is on the order of 10 -6 M. Based on an upper limit plutonium solubility of 10 -5 M in high pH and a conservative distribution coefficient for organic extractants of a 0 for plutonium in 30% TBP at 0.07 M HNO 3 ), the estimated concentration for plutonium in the organic phase would be -7 M. This is well below the process control criteria. A significant increase in plutonium solubility or the E a o would have to occur to raise this concentration to the 0.01 M concern level for organics. Measured tank chemical component values, expected operating conditions, and the characteristics of the expected chemistry and extraction mechanisms indicate that concentration of plutonium from Hanford tank residual sludges to associated process organic extractants is significantly below levels of concern

  16. Benchmark Evaluation of Plutonium Nitrate Solution Arrays

    International Nuclear Information System (INIS)

    Marshall, M.A.; Bess, J.D.

    2011-01-01

    In October and November of 1981 thirteen approach-to-critical experiments were performed on a remote split table machine (RSTM) in the Critical Mass Laboratory of Pacific Northwest Laboratory (PNL) in Richland, Washington, using planar arrays of polyethylene bottles filled with plutonium (Pu) nitrate solution. Arrays of up to sixteen bottles were used to measure the critical number of bottles and critical array spacing with a tight fitting Plexiglas(reg s ign) reflector on all sides of the arrays except the top. Some experiments used Plexiglas shells fitted around each bottles to determine the effect of moderation on criticality. Each bottle contained approximately 2.4 L of Pu(NO3)4 solution with a Pu content of 105 g Pu/L and a free acid molarity H+ of 5.1. The plutonium was of low 240Pu (2.9 wt.%) content. These experiments were performed to fill a gap in experimental data regarding criticality limits for storing and handling arrays of Pu solution in reprocessing facilities. Of the thirteen approach-to-critical experiments eleven resulted in extrapolations to critical configurations. Four of the approaches were extrapolated to the critical number of bottles; these were not evaluated further due to the large uncertainty associated with the modeling of a fraction of a bottle. The remaining seven approaches were extrapolated to critical array spacing of 3-4 and 4-4 arrays; these seven critical configurations were evaluation for inclusion as acceptable benchmark experiments in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. Detailed and simple models of these configurations were created and the associated bias of these simplifications was determined to range from 0.00116 and 0.00162 ± 0.00006 ?keff. Monte Carlo analysis of all models was completed using MCNP5 with ENDF/BVII.0 neutron cross section libraries. A thorough uncertainty analysis of all critical, geometric, and material parameters was performed using parameter

  17. Dynamic process model of a plutonium oxalate precipitator

    International Nuclear Information System (INIS)

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

    1980-01-01

    A dynamic model of a plutonium oxalate precipitator is developed to provide a means of predicting plutonium inventory on a continuous basis. The model is based on state-of-the-art crystallization equations, which describe nucleation and growth phenomena. The model parameters were obtained through the use of batch experimental data. The model has been used to study the approach to steady state, to investigate the response to input transients, and to simulate the control of the precipitation process. 12 refs

  18. 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)

  19. Test and evaluation of the in-line plutonium solution K-absorption-edge densitometer at the Savannah River Plant. Phase I. Off-line testing results

    International Nuclear Information System (INIS)

    Smith, H.A. Jr.; Marks, T.; Johnson, S.S.

    1982-04-01

    An in-line, plutonium-solution, K-edge absorption densitometer has been developed at Los Alamos and is currently undergoing test and evaluation at the Savannah River Plant (SRP). The first phase of the test and evaluation (off-line instrument calibration and solution assays) was completed, and preparations are under way to install the instrument in-line, as soon as process schedules permit. Calibration data in the design concentration range of 25 to 40 g Pu/L demonstrate routine achievement of densitometry assay precisions of 0.5% or better in 40 min. Plutonium assays at concentrations outside the calibration range were investigated in an effort to define better the limitations of the instrument and address other possible assay situations at SRP. Densitometry precisions obtained for 40-min assays range from 3% to 5 g Pu/L down to 0.4% at 70 g Pu/L. At higher plutonium concentrations, the precision deteriorated due to increasing gamma-ray absorption by the solution. In addition, with actinide concentrations above approximately 100 g/L, the assay accuracy also suffered because of enhanced small-angle scattering effects in the large sample cell. Measurements on mixed U/Pu solutions demonstrated the feasibility of accurate plutonium assays with correction for the large uranium matrix contributions being determined from the measurement data. The 239 240 Pu weight fractions and 241 Pu/ 239 Pu and 238 Pu/ 239 Pu isotopic ratios can be determined. In a mockup of the in-line solution plumbing system, all assay sequences, error conditions, and interlock criteria were exercised and verified to be working properly

  20. 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

  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. The Plutonium Temperature Effect Experimental Program

    Energy Technology Data Exchange (ETDEWEB)

    Haeck, Wim; Leclaire, Nicolas; Letang, Eric [IRSN, Fontenay-aux-Roses (France); Girault, Emmanuel; Fouillaud, Patrick [CEA, VALDUC (France)

    2008-07-01

    Various theoretical studies have shown that highly diluted plutonium solutions could have a positive temperature effect but (up to now) no experimental program has confirmed this effect. The main goal of the French Plutonium Temperature Effect Experimental Program (or PU+ in short) is to effectively show that such a positive temperature effect exists for diluted plutonium solutions. The experiments were conducted in the 'Apparatus B' facility at the CEA Valduc research centre in France and involved several sub-critical approach type of experiments using plutonium nitrate solutions with concentrations of 14.3, 15 and 20 g/l at temperatures ranging from 20 to 40 deg. C. A total number of 14 phase I experiments (consisting of independent subcritical approaches) have been performed (5 at 20 g/l, 4 at 15 g/l and 5 at 14.3 g/l) between 2006 and 2007. The impact of the uncertainties on the solution acidity and the plutonium concentration makes it difficult to clearly demonstrate the positive temperature effect, requiring an additional phase II experiment (in which the use of the same plutonium solution was ensured) from 22 to 28 deg. C performed in July 2007. This experiment has shown the existence of a positive temperature effect approx +2 pcm/deg. C (from 22 to 28 deg. C for a plutonium concentration of 14.3 g/l). (authors)

  3. 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

  4. Development of an expert system for analysis of plutonium processing operations

    International Nuclear Information System (INIS)

    Boeringter, S.T.; Fasel, J.H.; Kornreich, D.E.

    2001-01-01

    At Los Alamos National Laboratory (LANL) an expert system has been developed for the analysis and assessment of plutonium processing operations. This system is based upon an object-oriented simulation environment specifically developed for the needs of nuclear material processing. The simulation environment, called the ''Process Modeling System'' (ProMoS), contains a library of over 250 plutonium-based unit process operations ranging from analytical chemistry, oxide operations, recycle and recovery, waste management, and component fabrication. (author)

  5. Civil plutonium management

    International Nuclear Information System (INIS)

    Sicard, B.; Zaetta, A.

    2004-01-01

    During 1960 and 1970 the researches on the plutonium recycling in fast neutrons reactors were stimulated by the fear of uranium reserves diminishing. At the beginning of 1980, the plutonium mono-recycling for water cooled reactors is implementing. After 1990 the public opinion concerning the radioactive wastes management and the consequences of the disarmament agreements between Russia and United States, modified the context. This paper presents the today situation and technology associated to the different options and strategical solutions of the plutonium management: the plutonium use in the world, the neutronic characteristics, the plutonium effect on the reactors characteristics, the MOX behavior in the reactors, the MOX fabrication and treatment, the possible improvements to the plutonium use, the concepts performance in a nuclear park. (A.L.B.)

  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. PFP solution stabilization

    International Nuclear Information System (INIS)

    Aftanas, B.L.

    1996-01-01

    This Functional Design Criteria (FDC) addresses remediation of the plutonium-bearing solutions currently in inventory at the Plutonium Finishing Plant (PFP). The recommendation from the Environmental Impact Statement (EIS) is that the solutions be treated thermally and stabilized as a solid for long term storage. For solutions which are not discardable, the baseline plan is to utilize a denitration process to stabilize the solutions prior to packaging for storage

  9. Plutonium controversy

    International Nuclear Information System (INIS)

    Gofman, J.W.

    1976-01-01

    If the world chooses to seek a solution to the energy dilemma through nuclear energy, the element plutonium will become an article of commerce to be handled in quantities of thousands of tonnes annually. Plutonium is a uniquely potent inhalation carcinogen, the potential induction of lung cancer dwarfing other possible toxic effects. For reasons to be presented here, it is the author's opinion that plutonium's carcinogenicity has been very seriously underestimated. If one couples the corrected carcinogenicity with the probable degree of industrial containment of the plutonium, it appears that the commercialization of a plutonium-based energy economy is not an acceptable option for society. Sagan's statement that ''the experience of 30 years supports the contention that plutonium can be used safely'' is manifestly indefensible. No meaningful epidemiological study of plutonium-exposed workers for that 30-year period has ever been done. Since thousands of those possibly exposed have left the industry and are not even available to follow-up, it is doubtful that any meaningful study of ''the experience of 30 years'' will ever be accomplished

  10. Evaluation of chloride-ion-specific electrodes as in situ chemical sensors for monitoring total chloride concentration in aqueous solutions generated during the recovery of plutonium from molten salts used in plutonium electrorefining operations

    International Nuclear Information System (INIS)

    Smith, W.H.

    1992-10-01

    Two commercially available chloride-ion-specific electrodes (CLISEs), a solid-state type and a membrane type, were evaluated as potential in situ chemical sensors for determining total chloride ion concentration in mixed sodium chloride/potassium chloride/hydrochloric acid solutions generated during the recovery of plutonium from molten salts used in plutonium electrorefining operations. Because the response of the solid-state CLISE was closer than was the response of the membrane-type CLISE to the theoretical response predicted by the Nernst equation, the solid-state CLISE was selected for further evaluation. A detailed investigation of the characteristics of the chloride system and the corresponding CLISE response to concentration changes suggested four methods by which the CLISE could be used either as a direct, in situ sensor or as an indirect sensor through which an analysis could be performed on-line with a sample extracted from the process solution

  11. 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

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

    International Nuclear Information System (INIS)

    Harty, D.P.

    1993-12-01

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

  13. Plutonium determination by isotope dilution

    International Nuclear Information System (INIS)

    Lucas, M.

    1980-01-01

    The principle is to add to a known amount of the analysed solution a known amount of a spike solution consisting of plutonium 242. The isotopic composition of the resulting mixture is then determined by surface ionization mass spectrometry, and the plutonium concentration in the solution is deduced, from this measurement. For irradiated fuels neutronic studies or for fissile materials balance measurements, requiring the knowledge of the ratio U/Pu or of concentration both uranium and plutonium, it is better to use the double spike isotope dilution method, with a spike solution of known 233 U- 242 Pu ratio. Using this method, the ratio of uranium to plutonium concentration in the irradiated fuel solution can be determined without any accurate measurement of the mixed amounts of sample and spike solutions. For fissile material balance measurements, the uranium concentration is determined by using single isotope dilution, and the plutonium concentration is deduced from the ratio Pu/U and U concentration. The main advantages of isotope dilution are its selectivity, accuracy and very high sensitivity. The recent improvements made to surface ionization mass spectrometers have considerably increased the precision of the measurements; a relative precision of about 0.2% to 0.3% is obtained currently, but it could be reduced to 0.1%, in the future, with a careful control of the experimental procedures. The detection limite is around 0.1 ppb [fr

  14. Plutonium and americium separation from salts

    International Nuclear Information System (INIS)

    Hagan, P.G.; Miner, F.J.

    1976-01-01

    Salts or materials containing plutonium and americium are dissolved in hydrochloric acid, heated, and contacted with an alkali metal carbonate solution to precipitate plutonium and americium carbonates which are thereafter readily separable from the solution

  15. Nuclear fuel technology - Determination of milligram amounts of plutonium in nitric acid solutions - Potentiometric titration with potassium dichromate after oxidation by Ce(IV) and reduction by Fe(II)

    International Nuclear Information System (INIS)

    2000-01-01

    This International Standard describes a precise and accurate analytical method for determining 1 mg to 5 mg of plutonium per millilitre in nitric acid solutions. The method is very selective for plutonium. It is suitable for the direct determination of plutonium in materials ranging from pure product solutions, to solutions of mixed nuclear materials with a uranium/plutonium ratio up to 20:1. However, potential application to the assay of plutonium in solutions of irradiated nuclear fuels and solutions of mixed nuclear materials with uranium/plutonium ratios of 20:1 to 33:1 has not yet been documented. The method recommends that the aliquot be weighed and that the titration burettes be calibrated gravimetrically in order to obtain adequate precision and accuracy. This does not preclude using any alternative technique which can be shown to give an equivalent accuracy. As the reproducibility of the reaction conditions is important to maintain good performance, extensive automatization of the procedure is beneficial

  16. Vapor-liquid equilibria for nitric acid-water and plutonium nitrate-nitric acid-water solutions

    International Nuclear Information System (INIS)

    Maimoni, A.

    1980-01-01

    The liquid-vapor equilibrium data for nitric acid and nitric acid-plutnonium nitrate-water solutions were examined to develop correlations covering the range of conditions encountered in nuclear fuel reprocessing. The scanty available data for plutonium nitrate solutions are of poor quality but allow an order of magnitude estimate to be made. A formal thermodynamic analysis was attempted initially but was not successful due to the poor quality of the data as well as the complex chemical equilibria involved in the nitric acid and in the plutonium nitrate solutions. Thus, while there was no difficulty in correlating activity coefficients for nitric acid solutions over relatively narrow temperature ranges, attempts to extend the correlations over the range 25 0 C to the boiling point were not successful. The available data were then analyzed using empirical correlations from which normal boiling points and relative volatilities can be obtained over the concentration ranges 0 to 700 g/l Pu, 0 to 13 M nitric acid. Activity coefficients are required, however, if estimates of individual component vapor pressures are needed. The required ternary activity coefficients can be approximated from the correlations

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

    International Nuclear Information System (INIS)

    Reilly, Sean Douglas; Smith, Paul Herrick; Jarvinen, Gordon D.; Prochnow, David Adrian; Schulte, Louis D.; DeBurgomaster, Paul Christopher; Fife, Keith William; Rubin, Jim; Worl, Laura Ann

    2016-01-01

    Understanding the water solubility of plutonium and uranium compounds and residues at TA-55 is necessary to provide a technical basis for appropriate criticality safety, safety basis and accountability controls. Individual compound solubility was determined using published solubility data and solution thermodynamic modeling. Residue solubility was estimated using a combination of published technical reports and process knowledge of constituent compounds. The scope of materials considered includes all compounds and residues at TA-55 as of March 2016 that contain Pu-239 or U-235 where any single item in the facility has more than 500 g of nuclear material. This analysis indicates that the following materials are not appreciably soluble in water: plutonium dioxide (IDC=C21), plutonium phosphate (IDC=C66), plutonium tetrafluoride (IDC=C80), plutonium filter residue (IDC=R26), plutonium hydroxide precipitate (IDC=R41), plutonium DOR salt (IDC=R42), plutonium incinerator ash (IDC=R47), uranium carbide (IDC=C13), uranium dioxide (IDC=C21), U 3 O 8 (IDC=C88), and uranium filter residue (IDC=R26). This analysis also indicates that the following materials are soluble in water: plutonium chloride (IDC=C19) and uranium nitrate (IDC=C52). Equilibrium calculations suggest that PuOCl is water soluble under certain conditions, but some plutonium processing reports indicate that it is insoluble when present in electrorefining residues (R65). Plutonium molten salt extraction residues (IDC=R83) contain significant quantities of PuCl 3 , and are expected to be soluble in water. The solubility of the following plutonium residues is indeterminate due to conflicting reports, insufficient process knowledge or process-dependent composition: calcium salt (IDC=R09), electrorefining salt (IDC=R65), salt (IDC=R71), silica (IDC=R73) and sweepings/screenings (IDC=R78). Solution thermodynamic modeling also indicates that fire suppression water buffered with a commercially-available phosphate

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-13

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

  19. Plutonium recovery from spent glass fiber paper fine air filter

    International Nuclear Information System (INIS)

    Rovnyj, S.I.; Guzhavin, V.I.; Pyatin, N.P.; Evlanov, D.S.

    2002-01-01

    Investigations into the realizing technology of plutonium recovery from waste glass paper filters of fine purification were conducted. Two process schemes involving the nitro-fluoro-acid treatment of glass paper in the mixture of nitric and hydrofluoric acids and the previous alkali treatment of glass paper with the following nitro-fluoro-acid leaching of plutonium from pulp by the mixture of nitric and hydrofluoric acids were developed. Alkali, nitrate solutions and insoluble precipitants were analyzed for plutonium content [ru

  20. Dissolution of aerosol particles collected from nuclear facility plutonium production process

    International Nuclear Information System (INIS)

    Ning Xu; Martinez, Alex; Schappert, Michael; Montoya, D.P.; Martinez, Patrick; Tandon, Lav

    2016-01-01

    A simple, robust analytical chemistry method has been developed to dissolve plutonium containing particles in a complex matrix. The aerosol particles collected on Marple cascade impactor substrates were shown to be dissolved completely with an acid mixture of 12 M HNO 3 and 0.1 M HF. A pressurized closed vessel acid digestion technique was utilized to heat the samples at 130 deg C for 16 h to facilitate the digestion. The dissolution efficiency for plutonium particles was 99 %. The resulting particle digestate solution was suitable for trace elemental analysis and isotope composition determination, as well as radiochemistry measurements. (author)

  1. Analysis of Americium in Transplutonium Process Solutions

    International Nuclear Information System (INIS)

    Ferguson, R.B.

    2001-01-01

    One of the more difficult analyses in the transplutonium field is the determination of americium at trace levels in a complex matrix such as a process dissolver solution. Because of these conditions a highly selective separation must precede the measurement of americium. The separation technique should be mechanically simple to permit remote operation with master-slave manipulators. For subsequent americium measurement by the mass spectroscopic isotopic-dilution technique, plutonium and curium interferences must also have been removed

  2. The plutonium fuel cycles

    International Nuclear Information System (INIS)

    Pigford, T.H.; Ang, K.P.

    1975-01-01

    The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)

  3. 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

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

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

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

  5. 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

  6. Separation of neptunium from uranium and plutonium in the Purex process

    International Nuclear Information System (INIS)

    Kolarik, Z.; Schuler, R.

    1984-01-01

    The possibility of removing neptunium from the Purex process in the first extraction cycle was investigated. Butyraldehyde was found to reduce Np(VI) to Np(V), but not Pu(IV) to Pu(III). Up to 99.7% Np can be separated from uranium and plutonium in the 1A extractor or, much more favourably, in an additional partitioning extractor. Hydroxylamine nitrate can be used for reducing Np(VI) to Np(V) in a uranium purification cycle at a high U concentration in the feed solution. Here the decontamination factor for Np can be as high as 2300 and is lowered if iron is present in the feed. (author)

  7. Extraction and purification of plutonium by a tertiary amine; Extraction et purification du plutonium par une amine tertiaire

    Energy Technology Data Exchange (ETDEWEB)

    Trentinian, M de; Chesne, A [Commissariat a l' Energie Atomique, Fontenay aux Roses, Section de Chimie des Actimides (France).Centre d' Etudes Nucleaires; Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    Trilaurylamine diluted with a paraffinic solvent (dodecane) was studied as part of the research dealing with the separation and purification of plutonium. The physical properties (solubility of nitrates in the amine as a function of temperature) and the resistance to radiations of this substance were examined. The extraction characteristics of nitric solutions of plutonium, uranium and certain fission products are given as a function of the following factors: concentration of the various ions in solution, valency states. A method of plutonium purification based on these results is presented. (author) [French] La trilaurylamine diluee par un solvant paraffinique (dodecane) a ete etudiee dans le cadre des recherches concernant la separation et la purification du plutonium. Une etude des caracteres physiques (solubilite des nitrates dans l'amine en fonction de la temperature) s'ajoute a celle de la tenue aux radiations de ce corps. Les caracteristiques d'extraction de solutions nitriques de plutonium, uranium, et certains produits de fission, sont donnes en fonction des facteurs suivants: concentration des differents ions en solution, etats de valence. On presente une methode de purification du plutonium basee sur ces resultats. (auteur)

  8. 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

  9. Determination of plutonium in nitric acid solutions using energy dispersive L X-ray fluorescence with a low power X-ray generator

    Energy Technology Data Exchange (ETDEWEB)

    Py, J. [Laboratoire Chrono-Environnement, UMR CNRS 6249, Université de Franche-Comté, 16 route de Gray, F-25030 Besançon (France); Commissariat à l’Énergie Atomique, Centre de Valduc, F-21120 Is-sur-Tille (France); Groetz, J.-E., E-mail: jegroetz@univ-fcomte.fr [Laboratoire Chrono-Environnement, UMR CNRS 6249, Université de Franche-Comté, 16 route de Gray, F-25030 Besançon (France); Hubinois, J.-C.; Cardona, D. [Commissariat à l’Énergie Atomique, Centre de Valduc, F-21120 Is-sur-Tille (France)

    2015-04-21

    This work presents the development of an in-line energy dispersive L X-ray fluorescence spectrometer set-up, with a low power X-ray generator and a secondary target, for the determination of plutonium concentration in nitric acid solutions. The intensity of the L X-rays from the internal conversion and gamma rays emitted by the daughter nuclei from plutonium is minimized and corrected, in order to eliminate the interferences with the L X-ray fluorescence spectrum. The matrix effects are then corrected by the Compton peak method. A calibration plot for plutonium solutions within the range 0.1–20 g L{sup −1} is given.

  10. Study of accurate volume measurement system for plutonium nitrate solution

    Energy Technology Data Exchange (ETDEWEB)

    Hosoma, T. [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works

    1998-12-01

    It is important for effective safeguarding of nuclear materials to establish a technique for accurate volume measurement of plutonium nitrate solution in accountancy tank. The volume of the solution can be estimated by two differential pressures between three dip-tubes, in which the air is purged by an compressor. One of the differential pressure corresponds to the density of the solution, and another corresponds to the surface level of the solution in the tank. The measurement of the differential pressure contains many uncertain errors, such as precision of pressure transducer, fluctuation of back-pressure, generation of bubbles at the front of the dip-tubes, non-uniformity of temperature and density of the solution, pressure drop in the dip-tube, and so on. The various excess pressures at the volume measurement are discussed and corrected by a reasonable method. High precision-differential pressure measurement system is developed with a quartz oscillation type transducer which converts a differential pressure to a digital signal. The developed system is used for inspection by the government and IAEA. (M. Suetake)

  11. 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.)

  12. Preparation of plutonium fluoride to obtain metal of high purity; Preparation de fluorures de plutonium pour l'obtention de metal de haute purete

    Energy Technology Data Exchange (ETDEWEB)

    Faugeras, P; Brut, A; Helou, R [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    In the process of treating irradiated uranium, plutonium can be separated from the majority of the fission products and from the uranium by TBP extraction cycles. The high purity necessary for metallurgical and nuclear physics experiments led us to consider more elaborate purification processes, and a specially adapted method of fluoride preparation. The first part of the paper describes purification cycles of plutonium in solution on ion exchange resins, and the results are given. The second part contains the description and results of the fluoride preparation method. (author) [French] Dans le processus du traitement de l'uranium irradie, les cycles d'extraction au TBP permettent la separation du plutonium de la majorite des produits de fission et de l'uranium. La haute purete exigee pour les experiences de metallurgie et de physique nucleaire nous a conduit a envisager des purifications plus poussee et un mode de confection des fluorures specialement adapte. La premiere partie de l'expose decrit et donne les resultats de cycles de purification du plutonium en solution sur des resines echangeuses d'ions. La seconde partie decrit et donne les resultats du mode de confection des fluorures. (auteur)

  13. Improving Efficiency with 3-D Imaging: Technology Essential in Removing Plutonium Processing Equipment from Plutonium Finishing Plant Gloveboxes

    International Nuclear Information System (INIS)

    Crow, Stephen H.; Kyle, Richard N.; Minette, Michael J.

    2008-01-01

    The Plutonium Finishing Plant at Hanford, Washington began operations in 1949 to process plutonium and plutonium products. Its primary mission was to produce plutonium metal, fabricate weapons parts, and stabilize reactive materials. These operations, and subsequent activities, were performed in remote production lines, consisting primarily of hundreds of gloveboxes. Over the years these gloveboxes and processes have been continuously modified. The plant is currently inactive and Fluor Hanford has been tasked to clean out contaminated equipment and gloveboxes from the facility so it can be demolished in the near future. Approximately 100 gloveboxes at PFP have been cleaned out in the past four years and about 90 gloveboxes remain to be cleaned out. Because specific commitment dates for this work have been established with the State of Washington and other entities, it is important to adopt work practices that increase the safety and speed of this effort. The most recent work practice to be adopted by Fluor Hanford D and D workers is the use of 3-D models to improve the efficiency of cleaning out radioactive gloveboxes at the plant. The use of 3-D models has significantly improved the work planning process by providing workers with a clear image of glovebox construction and composition, which is then used to determine cleanout methods and work sequences. The 3-D visual products enhance safety by enabling workers to more easily identify hazards and implement controls. In addition, the ability to identify and target the removal of radiological materials early in the D and D process provides substantial dose reduction for the workers

  14. Management of Russian military plutonium

    International Nuclear Information System (INIS)

    Zaleski, C.P.

    1996-01-01

    The objective of this paper is to propose and discuss a solution which enables storing as quickly as possible all weapons-grade plutonium from Russian military program in a way which would prevent diversion. Two main conditions apply to this solution. First, it should be achieved in a manner acceptable to Russian government, notably by preserving plutonium for possible future energy production, and second, the economics of the total system should be good enough to ensure no charge or limited charge for the storage of plutonium. A proposal is made to store plutonium in a specially designed fast reactor or specially designed reactor core. This solution could be favorable in comparison to other solutions applying the above mentioned goal and conditions. Additionally the proposed solution would have the following side advantages: utilizing available personnel and installations of the Russian nuclear complex; providing possible basis for decommissioning of older and less safe Russian reactors; giving experience of construction and operation of a series of sodium-cooled fast reactors. The major problem however is the need for large capital investment with the risk of getting no adequate return on investment due to difficult political and economic situation in Russia

  15. Cycle downstream: the plutonium question; Aval du cycle la question du plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Zask, G [Electricite de France, EDF/DAC, 75 - Paris (France); Rome, M [Electricite de France, EDF, Service Etudes et Projets Thermiques et Nucleaires, 92 - Courbevoie (France); Delpech, M [CEA Cadarache, Dept. d' Etudes des Reacteurs/SPRC, 13 - Saint-Paul-lez-Durance (France); and others

    1998-06-29

    This day, organized by the SFEN, took place at Paris the 4 june 1998. Nine papers were presented. They take stock on the plutonium physics and its utilization as a nuclear fuel. This day tried to bring information to answer the following questions: do people have to keep the plutonium in the UOX fuel or in the MOX fuel in order to use it for future fast reactors? Do people have to continue obstinately the plutonium reprocessing in the MOX for the PWR type reactors? Will it be realized a underground disposal? Can it be technically developed plutonium incinerators and is it economically interesting? The plutonium physics, the experimental programs and the possible solutions are presented. (A.L.B.)

  16. Solution assay instrument operations manual

    International Nuclear Information System (INIS)

    Li, T.K.; Marks, T.; Parker, J.L.

    1983-09-01

    An at-line solution assay instrument (SAI) has been developed and installed in a plutonium purification and americium recovery process area in the Los Alamos Plutonium Processing Facility. The instrument was designed for accurate, timely, and simultaneous nondestructive analysis of plutonium and americium in process solutions that have a wide range of concentrations and americium/plutonium ratios and for routine operation by process technicians who lack instrumentation background. The SAI, based on transmission-corrected, high-resolution gamma-ray spectroscopy, has two measurement stations attached to a single multichannel analyzer/computer system. To ensure the quality of assay results, the SAI has an internal measurement control program, which requires daily and weekly check runs and monitors key aspects of all assay runs. For a 25-ml sample, the assay precision is 5 g/l within a 2000-s count time

  17. Some aspects of a technology of processing weapons grade plutonium to nuclear fuel

    International Nuclear Information System (INIS)

    Bibilashvili, Y.; Glagovsky, E.M.; Zakharkin, B.S.; Orlov, V.K.; Reshetnikov, F.G.; Rogozkin, B.G.; Soloni-N, M.I.

    2000-01-01

    The concept by Russia to use fissile weapons-grade materials, which are being recovered from nuclear pits in the process of disarmament, is based on an assessment of weapons-grade plutonium as an important energy source intended for use in nuclear power plants. However, in the path of involving plutonium excessive from the purposes of national safety into industrial power engineering there are a lot of problems, from which effectiveness and terms of its disposition are being dependent upon. Those problems have political, economical, financial and environmental character. This report outlines several technology problems of processing weapons-grade metallic plutonium into MOX-fuel for reactors based on thermal and fast neutrons, in particular, the issue of conversion of the metal into dioxide from the viewpoint of fabrication of pelletized MOX-fuel. The processing of metallic weapons-grade plutonium into nuclear fuel is a rather complicated and multi-stage process, every stage of which is its own production. Some of the stages are absent in production of MOX-fuel, for instance the stage of the conversion, i.e. transferring of metallic plutonium into dioxide of the ceramic quality. At this stage of plutonium utilization some tasks must be resolved as follows: I. As a result of the conversion, a material purified from ballast and radiogenic admixtures has to be obtained. This one will be applied to fabricate pelletized MOX-fuel going from morphological, physico-mechanical and technological properties. II. It is well known that metallic gallium, which is used as an alloying addition in weapons-grade plutonium, actively reacts with multiple metals. Therefore, an important issue is to study the effect of gallium on the technology of MOX-fuel production, quality of the pellets, as well as the interaction of gallium oxide with zirconium and steel shells of fuel elements depending upon the content of gallium in the fuel. The rate of the interaction of gallium oxide

  18. The extraction of plutonium with triethylene glycol dichloride

    International Nuclear Information System (INIS)

    Aikin, A.M.; Moss, M.; Bruce, T.

    1951-03-01

    The extraction of plutonium by triethylene glycol dichloride (trigly) has been investigated briefly. The effect of (1) the valence state of the plutonium, (2) the concentration of nitric acid, (3) the concentration of ammonium nitrate and (4) the conditioning of the trigly was measured. The solubility of plutonium IV in trigly was found to be 70 mgms/ml. Solutions of plutonium in trigly and in concentrated nitric acid solutions have been examined spectrophotometrically. (author)

  19. The extraction of plutonium with triethylene glycol dichloride

    Energy Technology Data Exchange (ETDEWEB)

    Aikin, A M; Moss, M; Bruce, T

    1951-03-15

    The extraction of plutonium by triethylene glycol dichloride (trigly) has been investigated briefly. The effect of (1) the valence state of the plutonium, (2) the concentration of nitric acid, (3) the concentration of ammonium nitrate and (4) the conditioning of the trigly was measured. The solubility of plutonium IV in trigly was found to be 70 mgms/ml. Solutions of plutonium in trigly and in concentrated nitric acid solutions have been examined spectrophotometrically. (author)

  20. Photometric estimation of plutonium in product solutions and acid waste solutions using flow injection analysis technique

    International Nuclear Information System (INIS)

    Dhas, A.J.A.; Dharmapurikar, G.R.; Kumaraguru, K.; Vijayan, K.; Kapoor, S.C.; Ramanujam, A.

    1995-01-01

    Flow injection analysis technique is employed for the measurement of plutonium concentrations in product nitrate solutions by measuring the absorbance of Pu(III) at 565 nm and of Pu(IV) at 470 nm, using a Metrohm 662 photometer, with a pyrex glass tube of 2 nm (ID) inserted in the light path of the detector serving as a flow cell. The photometer detector never comes in contact with radioactive solution. In the case of acid waste solutions Pu is first purified by extraction chromatography with 2-ethyl hexyl hydrogen 2 ethyl hexyl phosphonate (KSM 17)- chromosorb and the Pu in the eluate in complexed with Arsenazo III followed by the measured of absorbance at 665 nm. Absorbance of reference solutions in the desired concentration ranges are measured to calibrate the system. The results obtained agree with the reference values within ±2.0%. (author). 3 refs., 1 tab

  1. Assay of low-level plutonium effluents

    International Nuclear Information System (INIS)

    Hsue, S.T.; Hsue, F.; Bowersox, D.F.

    1981-01-01

    In the plutonium recovery section at the Los Alamos National Laboratory, an effluent solution is generated that contains low plutonium concentration and relatively high americium concentration. Nondestructive assay of this solution is demonstrated by measuring the passive L x-rays following alpha decay. Preliminary results indicate that an average deviation of 30% between L x-ray and alpha counting can be achieved for plutonium concentrations above 10 mg/L and Am/Pu ratios of up to 3; for plutonium concentrations less than 10 mg/L, the average deviation is 40%. The sensitivity of the L x-ray assay is approx. 1 mg Pu/L

  2. A method for the determination of free nitric acid in aqueous plutonium nitrate solutions - potassium fluoride method

    International Nuclear Information System (INIS)

    Mair, M.A.

    1988-06-01

    Plutonium IV and VI, and certain other hydrolysable metals which may be present, are converted to non-interfering species by the addition of the sample to potassium fluoride solution. The free acid is then titrated with standard sodium hydroxide solution using phenolphthalein as an indicator. (author)

  3. Design safety features of containments used for handling plutonium in Reprocessing Plants

    International Nuclear Information System (INIS)

    Aherwal, P.; Achuthan, P.V.

    2016-01-01

    The plutonium present in spent fuel is separated from the associated uranium and fission products using solvent extraction cycles in process cells. Product plutonium nitrate solution containing trace concentrations of uranium and fission products is treated in the reconversion facility through a precipitation-calcination route and converted to sinterable grade plutonium oxide (PuO 2 ). All chemical operations involving materials with high plutonium content, both in solid and solution forms are carried out in glove boxes. Glove box provides an effective isolation from radioactive materials handled and acts as a barrier between the operator and the source of radiation. These glove boxes are interconnected for sequential operations and the interconnected glove box trains are installed within secondary enclosures called double skin which provides double barrier protection to operators

  4. An MCNP model of glove boxes in a plutonium processing facility

    International Nuclear Information System (INIS)

    Dooley, D.E.; Kornreich, D.E.

    1998-01-01

    Nuclear material processing usually occurs simultaneously in several glove boxes whose primary purpose is to contain radioactive materials and prevent inhalation or ingestion of radioactive materials by workers. A room in the plutonium facility at Los Alamos National Laboratory has been slated for installation of a glove box for storing plutonium metal in various shapes during processing. This storage glove box will be located in a room containing other glove boxes used daily by workers processing plutonium parts. An MCNP model of the room and glove boxes has been constructed to estimate the neutron flux at various locations in the room for two different locations of the storage glove box and to determine the effect of placing polyethylene shielding around the storage glove box. A neutron dose survey of the room with sources dispersed as during normal production operations was used as a benchmark to compare the neutron dose equivalent rates calculated by the MCNP model

  5. An improved, computer-based, on-line gamma monitor for plutonium anion exchange process control

    International Nuclear Information System (INIS)

    Pope, N.G.; Marsh, S.F.

    1987-06-01

    An improved, low-cost, computer-based system has replaced a previously developed on-line gamma monitor. Both instruments continuously profile uranium, plutonium, and americium in the nitrate anion exchange process used to recover and purify plutonium at the Los Alamos Plutonium Facility. The latest system incorporates a personal computer that provides full-feature multichannel analyzer (MCA) capabilities by means of a single-slot, plug-in integrated circuit board. In addition to controlling all MCA functions, the computer program continuously corrects for gain shift and performs all other data processing functions. This Plutonium Recovery Operations Gamma Ray Energy Spectrometer System (PROGRESS) provides on-line process operational data essential for efficient operation. By identifying abnormal conditions in real time, it allows operators to take corrective actions promptly. The decision-making capability of the computer will be of increasing value as we implement automated process-control functions in the future. 4 refs., 6 figs

  6. Adaptation of the IBM ECR [electric cantilever robot] robot to plutonium processing applications

    International Nuclear Information System (INIS)

    Armantrout, G.A.; Pedrotti, L.R.; Halter, E.A.; Crossfield, M.

    1990-12-01

    The changing regulatory climate in the US is adding increasing incentive to reduce operator dose and TRU waste for DOE plutonium processing operations. To help achieve that goal the authors have begun adapting a small commercial overhead gantry robot, the IBM electric cantilever robot (ECR), to plutonium processing applications. Steps are being taken to harden this robot to withstand the dry, often abrasive, environment within a plutonium glove box and to protect the electronic components against alpha radiation. A mock-up processing system for the reduction of the oxide to a metal was prepared and successfully demonstrated. Design of a working prototype is now underway using the results of this mock-up study. 7 figs., 4 tabs

  7. The problem of utilization of the military uranium and plutonium

    International Nuclear Information System (INIS)

    Feoktistov, L.P.

    1995-01-01

    The problem on military uranium and plutonium is considered from the viewpoint of their utilization as a source of fissionable materials for NPPs. The solution consists in combining spherical geometry of critical mass with enriched center and the uranium burnup expansion recess. It is necessary thereby to obtain the minimum plutonium consumption in order to draw in unlimited quaintness of uranium-238 in the burnup process. It is estimated that hundred reactors with the total capacity of several hundred gigawatt may be involved into operation with the help of military plutonium. Refs. 2

  8. Decontamination of plutonium-contaminated surfaces

    International Nuclear Information System (INIS)

    Bertrand, J.; Clouet d'Orval, Ch.; Tachon, J.

    1958-01-01

    The measure of the neutron distribution in the core of 'Proserpine', by means of activation detectors, requires no contact between the plutonium sulfate solution and the detectors. These detectors are put into PVC or polyethylene bags. This report describes the process used to decontaminate these bags. A washing by nitric acid followed by coating with plexiglass is kept, with this process we have no contamination on the detectors. (author) [fr

  9. Recent improvements in plutonium gamma-ray analysis using MGA

    International Nuclear Information System (INIS)

    Ruhter, W.D.; Gunnink, R.

    1992-06-01

    MGA is a gamma-ray spectrum analysis program for determining relative plutonium isotopic abundances. It can determine plutonium isotopic abundances better than 1% using a high-resolution, low-energy, planar germanium detector and measurement times ten minutes or less. We have modified MGA to allow determination of absolute plutonium isotopic abundances in solutions. With calibration of a detector using a known solution concentration in a well-defined sample geometry, plutonium solution concentrations can be determined. MGA can include analysis of a second spectrum of the high-energy spectrum to include determination of fission product abundances relative to total plutonium. For the high-energy gamma-ray measurements we have devised a new hardware configuration, so that both the low- and high-energy gamma-ray detectors are mounted in a single cryostat thereby reducing weight and volume of the detector systems. We describe the detector configuration, and the performance of the MGA program for determining plutonium concentrations in solutions and fission product abundances

  10. Test procedure for anion exchange testing with Argonne 10-L solutions

    International Nuclear Information System (INIS)

    Compton, J.A.

    1995-01-01

    Four anion exchange resins will be tested to confirm that they will sorb and release plutonium from/to the appropriate solutions in the presence of other cations. Certain cations need to be removed from the test solutions to minimize adverse behavior in other processing equipment. The ion exchange resins will be tested using old laboratory solutions from Argonne National Laboratory; results will be compared to results from other similar processes for application to all plutonium solutions stored in the Plutonium Finishing Plant

  11. Zirconolite glass-ceramics for plutonium immobilization: The effects of processing redox conditions on charge compensation and durability

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yingjie, E-mail: yzx@ansto.gov.au; Gregg, Daniel J.; Kong, Linggen; Jovanovich, Miodrag; Triani, Gerry

    2017-07-15

    Zirconolite glass-ceramic samples doped with plutonium have been prepared via hot isostatic pressing. The effects of processing redox and plutonium loadings on plutonium valences, the presence of cation vacancies, zirconolite phase compositions, microstructures and durability have been investigated. Either tetravalent or trivalent plutonium ions may be incorporated on the Ca-site of CaZrTi{sub 2}O{sub 7} zirconolite with the Ca-site cation vacancies and the incorporation of Al{sup 3+} ions on the Ti-site for charge compensation. Plutonium and gadolinium (as a neutron absorber) are predominantly partitioned in zirconolite phases leading to the formation of chemically durable glass-ceramics suitable for the immobilization of impure plutonium wastes arising from the nuclear fuel cycle. - Highlights: •Plutonium validations of zirconolite glass-ceramics. •Effects of processing redox and plutonium loading. •Zirconolite phase compositions and plutonium valences. •Cation vacancies and chemical durability.

  12. Determination of plutonium 241 in solutions of nuclear wastes

    International Nuclear Information System (INIS)

    Raymond, A.; Bilcot, J.B.; Poletiko, C.

    1990-09-01

    Determination of plutonium 241 in nuclear wastes is important because of long period and high energy of some daughter products. In this report are presented two quantitative analysis methods using both scintillation techniques: A complete method, in any case, by selective extraction of plutonium on an anionic resin allowing simultaneous determination of Pu 241 and the sum of other plutonium isotopes; a simplified method when alpha activity is higher than beta/gamma activity by liquid extraction with TTA. These methods are applied for analysis of 4 waste types: cement encapsulated wastes, bitumen encapsulated wastes, incineration ashes, leaching of encapsulated incineration ashes. In these 4 examples, Pu 241 activity is equal or higher than the sum of alpha plutonium isotope activity. Separation efficiency, measured from Pu 239 or with Pu 236 as tracer, is between 90 and 99% [fr

  13. 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.

  14. 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

  15. Separation of americium and plutonium from nuclear wastes by the TRUEX process

    International Nuclear Information System (INIS)

    Leonard, R.A.; Vandegrift, G.F.; Manry, C.W.

    1986-01-01

    Americium and plutonium can be removed from a transuranic (TRU) waste stream to <10 nCi/g by the TRUEX process. The resulting waste is nontransuranic, greatly reducing disposal costs. An overview is given of the TRUEX process and of centrifugal contactors used to implement this process. Then, a plan for the deployment of TRUEX at the Hanford Site is discussed. Finally, details are given on the proposed use of TRUEX to treat the liquid wastes from the Plutonium Finishing Plant at the Hanford Site

  16. Solvent wash solution

    International Nuclear Information System (INIS)

    Neace, J.C.

    1986-01-01

    This patent describes a process for removing diluent degradation products from a solvent extraction solution comprising an admixture of an organic extractant for uranium and plutonium and a non-polar organic liquid diluent, which has been used to recover uranium and plutonium from spent nuclear fuel. Comprising combining a wash solution consisting of: (a) water; and (b) a positive amount up to about, an including, 50 volume percent of at least one highly-polar water-miscible organic solvent, based on the total volume of the water and the highly-polar organic solvent, with the solvent extraction solution after uranium and plutonium values have been stripped from the solvent extraction solution, the diluent degradation products dissolving in the highly-polar organic solvent and the extractant and diluent of the extraction solution not dissolving in the highly-polar organic solvent, and separating the highly-polar organic solvent and the extraction solution to obtain a purified extraction solution

  17. Comparison of different methods of determining plutonium content and isotopic composition

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    At Rockwell Hanford Operations, several different methods are used to determine plutonium content and isotopic composition. These include alpha particle energy analysis, calorimetry/gamma-ray analysis, mass spectrometry, and low energy gamma-ray assay. Each is used in a process control environment and has its advantages and disadvantages in terms of sample matrix, sample preparation, concentration, error ranges, detection limits, and turn around time. Of the methods discussed, special attention is paid to the Plutonium Isotopics Solution Counter, a low energy gamma-ray assay system designed to provide plutonium and americium content and isotopic composition of Pu-238 through Pu-241 and Am-241. It is qualitatively and quantitatively compared to the other methods. A brief description of sample types which the Solution Counter analyzes is presented

  18. Resin bead-thermal ionization mass spectrometry for determination of plutonium concentration in irradiated fuel dissolver solution

    International Nuclear Information System (INIS)

    Paul, Sumana; Shah, R.V.; Aggarwal, S.K.; Pandey, A.K.

    2015-01-01

    Determination of isotopic composition (IC) and concentration of plutonium (Pu) is necessary at various stages of nuclear fuel cycle which involves analysis of complex matrices like dissolver solution of irradiated fuel, nuclear waste stream etc. Mass spectrometry, e.g. thermal ionization mass spectrometry (TIMS) and inductively coupled plasma mass spectrometry (ICP-MS) are commonly used for determination of IC and concentration of plutonium. However, to circumvent matrix interferences, efficient separation as well as preconcentration of Pu is required prior to mass spectrometric analysis. Purification steps employing ion-exchange resins are widely used for the separation of Pu from dissolver solution or from mixture of other actinides e.g. U, Am. However, an alternative way is to selectively preconcentrate Pu on a resin bead, followed by direct loading of the bead on the filament of thermal ionization mass spectrometer

  19. Some plutonium IV polymers properties in Purex process

    International Nuclear Information System (INIS)

    Scoazec, H.; Pasquiou, J.Y.; Germain, M.

    1990-01-01

    The metabolism of plutonium polymers in fuel reprocessing using the Purex process with tributylphosphate as solvent, and its practical consequence in real operation conditions are examined. Precipitation with dibutylphosphoric acid, a solvent degradation product, occurs both in extraction and stripping units when polymers are present. (author)

  20. Application of microwaves in the denitration of nitric solutions of uranium and/or plutonium

    International Nuclear Information System (INIS)

    Quesada, C.A.; Adelfang, P.

    1990-01-01

    A method for the conversion of nitric solutions of uranium and/or plutonium that would be an alternative more economic and operatively simpler than the conventional processes is the direct denitration by means of microwaves and vacuum application. This conversion method has the following technical advantages: a) the process is simple, which allows a stable operation; b) neither the addition of chemical reagents nor the dilution of the starting solution are required, thereby the volume of residual liquids is small as compared with other processes; c) one fraction of the evaporation residues is nitric acid which can be reused. The development (on laboratory scale) of this conversion process was initiated. In this first stage, a description of the employed equipment is presented. An example of one of the evaporation and denitration batches and obtained products are fully described. The operative experience leads to deduce that the equipment is satisfactory, due to the following characteristics: 1) it permits an easy manipulation within the glove boxes; 2) the projections, coming out from the reactor, are retained completely; 3) the microwaves oven and the vacuum pump are effectively protected from the corrosive vapors. It is concluded that the employed experimental device is adequate to obtain the necessary materials for the reduction, pressing and sinterability studies. This equipment is adopted for the integral development of sintered pellets fabrication process. (Author) [es

  1. Applications of molten salts in plutonium processing

    International Nuclear Information System (INIS)

    Bowersox, D.F.; Christensen, D.C.; Williams, J.D.

    1987-01-01

    Plutonium is efficiently recovered from scrap at Los Alamos by a series of chemical reactions and separations conducted at temperatures ranging from 700 to 900 0 C. These processes usually employ a molten salt or salt eutectic as a heat sink and/or reaction medium. Salts for these operations were selected early in the development cycle. The selection criteria are being reevaluated. In this article we describe the processes now in use at Los Alamos and our studies of alternate salts and eutectics

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

    International Nuclear Information System (INIS)

    Heinonen, O.J.

    1981-01-01

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

  5. Continued studies of the gastrointestinal absorption of plutonium by rodents

    International Nuclear Information System (INIS)

    Larsen, R.P.; Bhattacharyya, M.H.; Oldham, R.D.; Moretti, E.S.; Spaletto, M.I.

    1982-01-01

    In the mouse the gastrointestinal absorption of hexavalent plutonium (the form present in chlorinated drinking water) is (1) a factor of about ten lower in the fed animal than in the fasted one (0.015 vs 0.15%), (2) independent of plutonium concentration over a range that broadly brackets the MPC for plutonium in drinking water, and (3) independent of the time of day the solution is administered to fasted animals. Other factors related to the determination of G.I. absorption which have been investigated are: (1) the adsorption of plutonium onto teeth of animals during both gavage and ad libitum administrations, (2) the formation of polymeric tetravalent plutonium during and subsequent to solution preparation, and (3) the relationship between the metabolic behavior of plutonium solutions, administered both intragastrically and intravenously, and their ultrafilterability

  6. 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.

  7. Uncertainty propagation for the coulometric measurement of the plutonium concentration in CRM126 solution provided by JAEA

    Energy Technology Data Exchange (ETDEWEB)

    Morales-Arteaga, Maria [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-11-07

    This GUM WorkbenchTM propagation of uncertainty is for the coulometric measurement of the plutonium concentration in a Pu standard material (C126) supplied as individual aliquots that were prepared by mass. The C126 solution had been prepared and as aliquoted as standard material. Samples are aliquoted into glass vials and heated to dryness for distribution as dried nitrate. The individual plutonium aliquots were not separated chemically or otherwise purified prior to measurement by coulometry in the F/H Laboratory. Hydrogen peroxide was used for valence adjustment.

  8. Development of isotope dilution gamma-ray spectrometry for plutonium analysis

    Energy Technology Data Exchange (ETDEWEB)

    Li, T.K.; Parker, J.L. (Los Alamos National Lab., NM (United States)); Kuno, Y.; Sato, S.; Kurosawa, A.; Akiyama, T. (Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan))

    1991-01-01

    We are studying the feasibility of determining the plutonium concentration and isotopic distribution of highly radioactive, spent-fuel dissolver solutions by employing high-resolution gamma-ray spectrometry. The study involves gamma-ray plutonium isotopic analysis for both dissolver and spiked dissolver solution samples, after plutonium is eluted through an ion-exchange column and absorbed in a small resin bead bag. The spike is well characterized, dry plutonium containing {approximately}98% of {sup 239}Pu. By using measured isotopic information, the concentration of elemental plutonium in the dissolver solution can be determined. Both the plutonium concentration and the isotopic composition of the dissolver solution obtained from this study agree well with values obtained by traditional isotope dilution mass spectrometry (IDMS). Because it is rapid, easy to operate and maintain, and costs less, this new technique could be an alternative method to IDMS for input accountability and verification measurements in reprocessing plants. 7 refs., 4 figs., 4 tabs.

  9. CONTAMINATED PROCESS EQUIPMENT REMOVAL FOR THE DECOMMISSIONG AND DECONTAMINATION OF THE 232-Z CONTAMINATED WASTE RECOVERY PROCESS FACILITY AT THE PLUTONIUM FINISHING PLANT

    International Nuclear Information System (INIS)

    HOPKINS, A.M.; MINETTE, M.J.; KLOS, D.B.

    2007-01-01

    This paper describes the unique challenges encountered and subsequent resolutions to accomplish the deactivation and decontamination of a plutonium ash contaminated building. The 232-Z Contaminated Waste Recovery Process Facility at the Plutonium Finishing Plant was used to recover plutonium from process wastes such as rags, gloves, containers and other items by incinerating the items and dissolving the resulting ash. The incineration process resulted in a light-weight plutonium ash residue that was highly mobile in air. This light-weight ash coated the incinerator's process equipment, which included gloveboxes, blowers, filters, furnaces, ducts, and filter boxes. Significant airborne contamination (over 1 million derived air concentration hours [DAC]) was found in the scrubber cell of the facility. Over 1300 grams of plutonium held up in the process equipment and attached to the walls had to be removed, packaged and disposed. This ash had to be removed before demolition of the building could take place

  10. Proserpine - plutonium 239 - Proserpine - uranium 235 - comparison of experimental results; Proserpine - plutonium 239 - proserpine - uranium 235 - comparaison de resultats experimentaux

    Energy Technology Data Exchange (ETDEWEB)

    Brunet, J P; Caizergues, R; Clouet D' Orval, Ch; Kremser, J; Moret-Bailly, J; Verriere, Ph [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The Proserpine homogeneous reactor is constituted by a tank, 25 cm dia, 30 cm high, surrounded by a composite reflector made of beryllium oxide and graphite. In this tank can be made critical plutonium or 90 per cent enriched uranium solutions, the fissile substances being in the form of a dissolved salt. In varying the concentration of the solution, critical masses were studied as a function of the level of the liquid in the tank. The minimum critical mass is 256 {+-} 2 grs for plutonium and 409 {+-} 3 grs for uranium 235. In the range of the critical concentrations which were studied, the neutronic properties of fissionable solutions of plutonium and enriched uranium were compared for identical geometries. (authors) [French] Proserpine est un reacteur homogene comportant une cuve de diametre 25 cm, de hauteur 30 cm, entouree d'un reflecteur composite d'oxyde de beryllium et de graphite. On y a rendu critiques des solutions de plutonium ou d'uranium enrichi a 90 pour cent, le produit fissile se trouvant sous la forme d'un sel dissous. En faisant varier la concentration de la solution, on a etudie les masses critiques en fonction de la hauteur du liquide dans la cuve. La masse- critique minimum est, pour le plutonium de 256 {+-} 2 g, pour l'uranium 235 de 409 {+-} 3 g. Dans la gamme des concentrations critiques etudiees, on a compare, dans des conditions de geometrie identique, les proprietes neutroniques des solutions fissiles de plutonium et d'uranium enrichi. (auteurs)

  11. Cycle downstream: the plutonium question

    International Nuclear Information System (INIS)

    Zask, G.; Rome, M.; Delpech, M.

    1998-01-01

    This day, organized by the SFEN, took place at Paris the 4 june 1998. Nine papers were presented. They take stock on the plutonium physics and its utilization as a nuclear fuel. This day tried to bring information to answer the following questions: do people have to keep the plutonium in the UOX fuel or in the MOX fuel in order to use it for future fast reactors? Do people have to continue obstinately the plutonium reprocessing in the MOX for the PWR type reactors? Will it be realized a underground disposal? Can it be technically developed plutonium incinerators and is it economically interesting? The plutonium physics, the experimental programs and the possible solutions are presented. (A.L.B.)

  12. 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.

  13. 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

  14. 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)

  15. 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.

  16. The plutonium danger

    International Nuclear Information System (INIS)

    Ruiter, W. de

    1983-01-01

    Nobody can ignore the fact that plutonium is potentially very dangerous and the greatest danger concerning it lies in the spreading of nuclear weapons via nuclear energy programmes. The following seven different attitudes towards this problem are presented and discussed: 1) There is no connection between peaceful and military applications; 2) The problem cannot be prevented; 3) A technical solution must be found; 4) plutonium must be totally inaccessible to countries involved in acquiring nuclear weapons; 5) The use of plutonium for energy production should only occur in one multinational centre; 6) Dogmas in the nuclear industry must be enfeebled; 7) All developments in this area should stop. (C.F.)

  17. Comparison of different methods of determining plutonium content and isotopic composition

    International Nuclear Information System (INIS)

    Dowell, M.R.W.

    1985-05-01

    At Rockwell Hanford Operations, several different methods are used to determine plutonium content and isotopic composition. These include alpha particle energy analysis, calorimetry/gamma-ray analysis, mass spectrometry, and low energy ray assay. Each is used in a process control environment and has its advantages and disadvantages in terms of sample matrix, sample preparation, concentration, error ranges, detection limits, and turn around time. Of the methods discussed, special attention is paid to the Plutonium Isotopics Solution Counter, a low energy gamma ray assay system designed to provide plutonium and americium content and isotopic composition of Pu-238 through Pu-241 and Am-241. It is qualitatively and quantitatively compared to the other methods. A brief description of sample types which the Solution Counter analyzes is presented. 4 refs., 4 tabs

  18. Evaluation of TASTEX task H: measurement of plutonium isotopic abundances by gamma-ray spectrometry

    International Nuclear Information System (INIS)

    Gunnink, R.; Prindle, A.L.; Asakura, Y.; Masui, J.; Ishiguro, N.; Kawasaki, A.; Kataoka, S.

    1981-10-01

    This report describes a computer-based gamma spectrometer system that was developed for measuring isotopic and total plutonium concentrations in nitric acid solutions. The system was installed at the Tokai reprocessing plant where it is undergoing testing and evaluation as part of the Tokai Advanced Safeguards Exercise (TASTEX). Objectives of TASTEX Task H, High-Resolution Gamma Spectrometer for Plutonium Isotopic Analysis, the methods and equipment used, the installation and calibration of the system, and the measurements obtained from several reprocessing campaigns are discussed and described. In general, we find that measurements for gamma spectroscopy agree well with those of mass spectrometry and of other chemical analysis. The system measures both freshly processed plutonium from the product accountability tank and aged plutonium solutions from storage tanks. 14 figures, 15 tables

  19. 233-S Plutonium Concentration Facility data quality objectives

    International Nuclear Information System (INIS)

    Encke, D.B.

    1996-08-01

    This document is a summary of the decision-making associated with the Data Quality Objective process that pertains to the characterization activities in the 233-S Plutonium Concentration Facility at the Hanford Site in Richland, Washington. The 233-S Plutonium Concentration Facility is located adjacent to, and north of, the REDOX Plant. The facility was used to concentrate the plutonium nitrate product solution from the REDOX facility. The 233-S Pipe Gallery, Control Room, SWP Change Room, Toilet, Equipment Room and the Electrical Cubicle are currently scheduled for decontamination and cleanout to support future demolition (D and D). Identification of the radiological contamination and presence of hazardous materials is needed to allow for disposal of the D and D debris

  20. Gamma radiolysis of alkaline aqueous solutions of neptunium and plutonium ions

    International Nuclear Information System (INIS)

    Pikaev, A.K.; Gogolev, A.V.; Shilov, V.P.

    1998-01-01

    Full text: The paper is a brief review of data obtained by the authors from the study on redox reactions of neptunium and plutonium ions upon γ radiolysis of their aerated alkaline aqueous solutions. It includes the information on radiolytic reduction of Np(V), Np(VI) and Pu(VI) ions under various experimental conditions. It was found that the values of Np(VI) and Pu(VI) reduction yields do not depend on alkali concentration. The values considerably increase in the presence of some organic compounds (EDTA and formate were investigated). The formation of the Np(V) peroxo complex was observed in the γ radiolysis of alkaline aqueous solutions of Np(VI) and Np(V) in the presence of nitrate. The mechanism of radiolytic redox reactions of the ions is discussed in some detail

  1. PRODUCTION OF PLUTONIUM METAL

    Science.gov (United States)

    Lyon, W.L.; Moore, R.H.

    1961-01-17

    A process is given for producing plutonium metal by the reduction of plutonium chloride, dissolved in alkali metal chloride plus or minus aluminum chloride, with magnesium or a magnesium-aluminum alloy at between 700 and 800 deg C and separating the plutonium or plutonium-aluminum alloy formed from the salt.

  2. 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)

  3. Engineering report (conceptual design) PFP solution stabilization

    Energy Technology Data Exchange (ETDEWEB)

    Witt, J.B.

    1997-07-17

    This Engineering Report (Conceptual Design) addresses remediation of the plutonium-bearing solutions currently in inventory at the Plutonium Finishing Plant (PFP). The recommendation from the Environmental Impact Statement (EIS) is that the solutions be treated thermally and stabilized as a solid for long term storage. For solutions which are not discardable, the baseline plan is to utilize a denitration process to stabilize the solutions prior to packaging for storage.

  4. Engineering report (conceptual design) PFP solution stabilization

    International Nuclear Information System (INIS)

    Witt, J.B.

    1997-01-01

    This Engineering Report (Conceptual Design) addresses remediation of the plutonium-bearing solutions currently in inventory at the Plutonium Finishing Plant (PFP). The recommendation from the Environmental Impact Statement (EIS) is that the solutions be treated thermally and stabilized as a solid for long term storage. For solutions which are not discardable, the baseline plan is to utilize a denitration process to stabilize the solutions prior to packaging for storage

  5. Processing plutonium-contaminated soil on Johnston Atoll

    International Nuclear Information System (INIS)

    Moroney, K.; Moroney, J. III; Turney, J.

    1994-01-01

    This article describes a cleanup project to process plutonium- and americium-contaminated soil on Johnston Atoll for volume reduction. Thermo Analytical's (TMA's) segmented gate system (SGS) for this remedial operation has been in successful on-site operation since 1992. Topics covered include the basis for development, a description of the Johnston Atoll; the significance of results; the benefits of the technology; applicability to other radiologically contaminated sites. 7 figs., 1 tab

  6. 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. â

  7. THE DEACTIVATION, DECONTAMINATION AND DECOMMISSIONING OF THE PLUTONIUM FINISHING PLANT, A FORMER PLUTONIUM PROCESSING FACILITY AT DOE'S HANFORD SITE

    International Nuclear Information System (INIS)

    CHARBONEAU, S.L.

    2006-01-01

    The Plutonium Finishing Plant (PFP) was constructed as part of the Manhattan Project during World War II. The Manhattan Project was developed to usher in the use of nuclear weapons to end the war. The primary mission of the PFP was to provide plutonium used as special nuclear material (SNM) for fabrication of nuclear devices for the war effort. Subsequent to the end of World War II, the PFP's mission expanded to support the Cold War effort through plutonium production during the nuclear arms race and later the processing of fuel grade mixed plutonium-uranium oxide to support DOE's breeder reactor program. In October 1990, at the close of the production mission for PFP, a shutdown order was prepared by the Department of Energy (DOE) in Washington,; DC--and issued to the Richland DOE field office. Subsequent to the shutdown order, a team from the Defense Nuclear Facilities Safety Board (DNFSB) analyzed the hazards at PFP associated with the continued storage of certain forms of plutonium solutions and solids. The assessment identified many discrete actions that were required to stabilize the different plutonium forms into stable form and repackage the material in high integrity containers. These actions were technically complicated and completed as part of the PFP nuclear material stabilization project between 1995 and early 2005. The completion of the stabilization project was a necessary first step in deactivating PFP. During stabilization, DOE entered into negotiations with the U.S. Environmental Protection Agency (EPA) and the State of Washington and established milestones for the Deactivation and Decommissioning (DandD) of the PFP. The DOE and its contractor, Fluor Hanford (Fluor), have made great progress in deactivating, decontaminating and decommissioning the PFP at the Hanford Site as detailed in this paper. Background information covering the PFP DandD effort includes descriptions of negotiations with the State of Washington concerning consent

  8. Safely disposing and controlling the various forms of excess military plutonium

    International Nuclear Information System (INIS)

    Albright, D.

    1991-01-01

    The growing surplus of plutonium will continue to pose safety, health, and verification problems. Although long term storage and disposal of plutonium seems technically feasible, or at least comparable in technical difficulty to commercial spent fuel disposal, significant political obstacles within the government and the public, may make it difficult to solve this problem. Although options to build verifiable warhead dismantlement facilities or to recycle plutonium in reactors and thus convert separated plutonium into irradiated fuel are straight forward concepts, their realization remains difficult for economic and political reasons. The plutonium recycle option also raises additional proliferation concerns about its impact on civilian nuclear programs. In the absence of a long term solution, the United States can implement various storage or interim disposal options that involve minimal processing, but that ease verification problems and provide adequate safety and protection of public health

  9. Study of plutonium sorption in aluminia column in the system HNO3-HF

    International Nuclear Information System (INIS)

    Araujo, J.A. de.

    1977-01-01

    The column chromatographic method using alumina has been applied successfully to study the sorption-desorption behavior of plutonium traces in HNO 3 -HF and HNO 3 -HF-UO 2 (NO 3 ) 2 systems, aiming to elaborate a process for recovering plutonium traces from reprocessing wastes, mainly in existing solutions where uranium is presented in macro quantities. Basically, the method consists in the sorption of plutonium by percolating a solution containing HNO 3 (0,1 to 0,8M) or uranyl nitrate (1-50 gU/l) and HF(0,1 to 0,3M) through an Al 2 O 3 collumn. The plutonium is fixed on Al 2 O 3 whereas the uranyl ions is collected in the efluent. The adsorption of Pu-III, Pu-IV and Pu-VI in the presence of HF was determined and Pu-IV can be almost completely sorbed. The Pu-IV is eluted by reduction to Pu-III in the column using 3 M HNO 3 -0,005M FeSO 4 at 50 0 C as elutrient. This method is very simple and can be applied for separation and purification of plutonium (traces) from uranyl nitrate or others coming solutions from wet chemistry of irradiated fuels [pt

  10. 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.

  11. Shielding calculational system for plutonium

    International Nuclear Information System (INIS)

    Zimmerman, M.G.; Thomsen, D.H.

    1975-08-01

    A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)

  12. An environmentally benign plutonium processing future at Los Alamos

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1993-01-01

    In recent years, the U.S. Department of Energy (DOE) has elevated environmental restoration and waste management to major mission areas, and it has established the reduction of wastes from DOE facilities as a major objective. The DOE facilities must now comply with all environmental regulations, including special regulations required of federal facilities. In recognition of this shift in philosophy, the plutonium processing facility at Los Alamos National Laboratory (LANL) has adopted the goal of becoming a facility that processes plutonium in a way that produces only environmentally benign waste. Becoming a facility with zero radionuclide and mixed-waste discharge is an extremely challenging goal and one that requires the technical contributions of a multidisciplinary team of experts. While all the technologies necessary to achieve this goal are not yet available, an extensive knowledge base does exist that can be applied to solving the remaining problems. Working toward this goal is a worthwhile endeavor, not only for LANL, but for the nuclear complex of the future

  13. Work surface for soluble plutonium

    International Nuclear Information System (INIS)

    Silver, G.L.

    2005-01-01

    A three-dimensional work surface for aqueous plutonium is illustrated. It is constructed by means of estimating work as a function of the ambient pH and redox potential in a plutonium solution. The surface is useful for illustrating the chemistry of disproportionation reactions. Work expressions are easier to use than work integrals. (author)

  14. Six-kilogram-scale electrorefining of plutonium metal

    International Nuclear Information System (INIS)

    Mullins, L.J.; Morgan, A.N.; Apgar, S.A. III; Christensen, D.C.

    1982-09-01

    The electrorefining of metallic plutonium scrap to produce high purity metal has been an established procedure at Los Alamos since 1964. This is a batch process and was limited to 4-kg plutonium because of criticality safety considerations. Improvements in critical mass measurements have permitted us to develop a process for 6-kg plutonium. The 6-kg process is now operational. The increased size of the process, together with other improvements which have been made, makes plutonium electrorefining the principal industrial tool for processing and purifying metallic plutonium scrap

  15. An Improved Plutonium Trifluoride Precipitation Flowsheet

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, H.D.

    2001-06-26

    This report discusses results of the plutonium trifluoride two-stage precipitation study. A series of precipitation experiments was used to identify the significant process variables affecting precipitation performance. A mathematical model of the precipitation process was developed which is based on the formation of plutonium fluoride complexes. The precipitation model relates all process variables, in a single equation, to a single parameter which can be used to control the performance of the plutonium trifluoride precipitation process. Recommendations have been made which will optimize the FB-Line plutonium trifluoride precipitation process.

  16. An Improved Plutonium Trifluoride Precipitation Flowsheet

    International Nuclear Information System (INIS)

    Harmon, H.D.

    2001-01-01

    This report discusses results of the plutonium trifluoride two-stage precipitation study. A series of precipitation experiments was used to identify the significant process variables affecting precipitation performance. A mathematical model of the precipitation process was developed which is based on the formation of plutonium fluoride complexes. The precipitation model relates all process variables, in a single equation, to a single parameter which can be used to control the performance of the plutonium trifluoride precipitation process. Recommendations have been made which will optimize the FB-Line plutonium trifluoride precipitation process

  17. Solidification of radioactive aqueous solution

    Energy Technology Data Exchange (ETDEWEB)

    Aikawa, Hideaki; Kato, Kiyoshi; Wadachi, Yoshiki

    1970-09-07

    A process for solidifying a radioactive waste solution is provided, using as a solidifying agent a mixture of calcined gypsum and burnt vermiculite. The quantity ratio of the mixture is preferred to be 1:1 by volume. The quantity of impregnation is 1/2 of the volume of the total quantity of the solidifying agent. In embodiments, 10 liters of plutonium waste solution was mixed with a mixture of 1:1 calcined gypsum and burnt vermiculite contained in a 20-liter cylindrical steel container lined with asphalt. The plutonium waste solution from the laboratory was neutralized with a caustic soda aqueous solution to prevent explosion due to the nitration of organic compounds. The neutralization is not always necessary. A market available dental gypsum was calcined at 400 to 500/sup 0/C and a vermiculite from Illinois was burnt at 1,100/sup 0/C to prepare the agents. The time required for the impregnation with 10 liters of plutonium solution was four minutes. After impregnation, the temperature rose to 40/sup 0/C within 30 minutes to one hour. Next, it was cooled to room temperature by standing for 3-4 hours. Solidification time was about 1 hour. The Japan Atomic Energy Research Insitute had treated and disposed about 1,000 tons of plutonium waste by this process as of August 19, 1970.

  18. Plutonium scrap waste processing based on aqueous nitrate and chloride media

    International Nuclear Information System (INIS)

    Navratil, J.D.

    1985-01-01

    A brief review of plutonium scrap aqueous waste processing technology at Rocky Flats is given. Nitric acid unit operations include dissolution and leaching, anion exchange purification and precipitation. Chloride waste processing consists of cation exchange and carbonate precipitation. Ferrite and carrier precipitation waste treatment processes are also described. 3 figs

  19. Plutonium titration by controlled potential coulometry; Dosage du plutonium par coulometrie a potentiel impose

    Energy Technology Data Exchange (ETDEWEB)

    Leguay, N.

    2011-07-01

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

  20. Plutonium Plant, Trombay

    International Nuclear Information System (INIS)

    Yadav, J.S.; Agarwal, K.

    2017-01-01

    The journey of Indian nuclear fuel reprocessing started with the commissioning of Plutonium Plant (PP) at Trombay on 22"n"d January, 1965 with an aim to reprocess the spent fuel from research reactor CIRUS. The basic process chosen for the plant was Plutonium Uranium Reduction EXtraction (PUREX) process. In seventies, the plant was subjected to major design modifications and replacement of hardware, which later met the additional demand from research reactor DHRUVA. The augmented plutonium plant has been operating since 1983. Experience gained from this plant was very much helpful to design future reprocessing plant in the country

  1. HF effect on dissociation kinetics of plutonium and neptunium complexes with 1,2-diaminocyclohexanetetraacetic acid in nitric acid solutions

    International Nuclear Information System (INIS)

    Nikitina, S.A.; Stepanov, A.V.

    1982-01-01

    Dissociation kinetics of Pusup((4)) and Np sup((4)) complexes with DCTA were investigated in HNO 3 solutions in the presence of HF and arsenazo 3. It was found that HF or NaF produced a differentiating effect on the reactivity of the complexes at [HNO 3 ]=1-6 mol/l as well as inhibiting effect at [HNO 3 ]=0.01 mol/l. Conditions of the differential kinetic analysis of plutonium and neptunium in the mixture and differential spectrophotometric analysis of uranium (6) during the camouflage of neptunium (4) and plutonium (4) were determined

  2. 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)

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

    International Nuclear Information System (INIS)

    Venault, L.

    1998-01-01

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

  4. 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

  5. Plutonium solution analyzer. Revised February 1995

    International Nuclear Information System (INIS)

    Burns, D.A.

    1995-02-01

    A fully automated analyzer has been developed for plutonium solutions. It was assembled from several commercially available modules, is based upon segmented flow analysis, and exhibits precision about an order of magnitude better than commercial units (0.5%--0.05% RSD). The system was designed to accept unmeasured, untreated liquid samples in the concentration range 40--240 g/l: and produce a report with sample identification, sample concentrations, and an abundance of statistics. Optional hydraulics can accommodate samples in the concentration range 0.4--4.0 g/y. Operating at a typical rate of 30 to 40 samples per hour, it consumes only 0.074 ml of each sample and standard, and generates waste at the rate of about 1.5 ml per minute. No radioactive material passes through its multichannel peristaltic pump (which remains outside the glovebox, uncontaminated) but rather is handled by a 6-port, 2-position chromatography-type loop valve. An accompanying computer is programmed in QuickBASIC 4.5 to provide both instrument control and data reduction. The program is truly user-friendly and communication between operator and instrument is via computer screen displays and keyboard. Two important issues which have been addressed are waste minimization and operator safety (the analyzer can run in the absence of an operator, once its autosampler has been loaded)

  6. In-line analytical methods for fuel reprocessing streams : Part IV -Neutron monitoring for plutonium

    International Nuclear Information System (INIS)

    Rao, V.K.; Bhargava, V.K.; Marathe, S.G.; Iyer, R.H.; Ramaniah, M.V.; Srinivasan, N.

    1975-01-01

    A neutron monitoring assembly consisting of a stainless steel housing packed with beryllium oxide chips, a paraffin moderator, a ring of fifteen BF 3 counters and an all stainless steel continuous flow system for circulating plutonium solutions has been fabricated and tested for monitoring plutonium concentrations in flow solutions. The method is based on the detection and measurement of neutron flux produced when alpha particles from plutonium interact with beryllium by the nuclear reactoon 9 4 Be(α,n) 12 6 C. The unit was successfully tested for the estimation of plutonium concentrations upto 10 g/1 in solutions of plutonium and plutonium solutions mixed with uranium and fission products. The unit gave an accuracy of 10-15%. Details of the construction and working of the system are discussed. (author)

  7. 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.

  8. Plutonium storage phenomenology

    International Nuclear Information System (INIS)

    Szempruch, R.

    1995-12-01

    Plutonium has been produced, handled, and stored at Department of Energy (DOE) facilities since the 1940s. Many changes have occurred during the last 40 years in the sources, production demands, and end uses of plutonium. These have resulted in corresponding changes in the isotopic composition as well as the chemical and physical forms of the processed and stored plutonium. Thousands of ordinary food pack tin cans have been used successfully for many years to handle and store plutonium. Other containers have been used with equal success. This paper addressees the exceptions to this satisfactory experience. To aid in understanding the challenges of handling plutonium for storage or immobilization the lessons learned from past storage experience and the necessary countermeasures to improve storage performance are discussed

  9. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

    2007-02-07

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  10. 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.

  11. 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

  12. 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.)

  13. A solvent proceed for the extraction of the irradiate uranium and plutonium in the reactor core; Un procede par solvant pour l'extraction du plutonium de l'uranium irradie dans les piles

    Energy Technology Data Exchange (ETDEWEB)

    Goldschmidt, B; Regnaut, P; Prevot, I [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    Description of the conditions of plutonium, fission products and of uranium separation by selective extraction of the nitrates by organic solvent, containing a simultaneous extraction of plutonium and uranium, followed by a plutonium re-extraction after reduction, and an uranium re-extraction. The rates of decontamination being insufficient in this first stage, we also describes the processes of decontamination permitting separately to get the rates wanted for uranium and plutonium. Finally, we describes the beginning of the operation that consists in a nitric dissolution of the active uranium while capturing the products of gaseous fission, as well as the final concentration of the products of fission in a concentrated solution. (authors) [French] Description des conditions de separation du plutonium, des produits de fission et de l'uranium au moyen d'une extraction selective des nitrates par solvant organique, comprenant une extraction simultanee du plutonium et de l'uranium, suivie d'une reextraction du plutonium apres reduction, et d'une reextraction de l'uranium. Les taux de decontamination etant insuffisants dans ce premier stade, on decrit egalement les processus de decontamination permettant separement d'obtenir les taux desires pour l'uranium et le plutonium. Enfin, on decrit aussi le debut de l'operation qui consiste en une dissolution nitrique de l'uranium actif en captant les produits de fission gazeux, ainsi que la concentration finale des produits de fission sous forme de solution concentree. (auteurs)

  14. Diluent paraffin nature and plutonium(IV) organic solution lamination: new results and new approach

    International Nuclear Information System (INIS)

    Renard, E.V.; Ivanchenko, V.A.; Chizhov, A.A.; Neumojev, N.V.

    1994-01-01

    The knowledge of the relation between the diluent composition and structure and the critical (maximum achievable) concentration (CC(Me)) of metals, including plutonium(IV), in the organic phase is an actual goal of a radiochemical extraction technology (PUREX process). Using γ-spectrometry analysis with high accuracy, the parameter CC(Pu) has been determined in application to Pu(IV) nitrate solution in 30% (vol.) TBP diluent solutions. n-Paraffins C 10 C 16 , iso-paraffins (mono- and dimethylderivatives), iso-paraffins with C-quaternare atoms (hydrogenated mixtures of tetra- and pentapropylene isomers) have been used as diluents. Regular correlations between CC(Pu) parameters and some individual (and mixture) diluent structure characteristics, including practically linear inversely proportional decrease of CC(Pu) with C-atom number increase (in molecules of n-C n H 2n+2 ), symbatically development of this relation to both the paraffin types (n- and iso-) have been found. The general straight proportional relation between CC(Pu)-parameter and fluidity (F=1/η, where η-dynamic viscosity at given temperature) has been discovered for all paraffins investigated - both individual and mixtures - at different temperatures. ((orig.))

  15. 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.

  16. Chloride-catalyzed corrosion of plutonium in glovebox atmospheres

    International Nuclear Information System (INIS)

    Burgess, M.; Haschke, J.M.; Allen, T.H.; Morales, L.A.; Jarboe, D.M.; Puglisi, C.V.

    1998-04-01

    Characterization of glovebox atmospheres and the black reaction product formed on plutonium surfaces shows that the abnormally rapid corrosion of components in the fabrication line is consistent with a complex salt-catalyzed reaction involving gaseous hydrogen chloride (HCl) and water. Analytical data verify that chlorocarbon and HCl vapors are presented in stagnant glovebox atmospheres. Hydrogen chloride concentrations approach 7 ppm at some locations in the glovebox line. The black corrosion product is identified as plutonium monoxide monohydride (PuOH), a product formed by hydrolysis of plutonium in liquid water and salt solutions at room temperature. Plutonium trichloride (PuCl 3 ) produced by reaction of HCl at the metal surface is deliquescent and apparently forms a highly concentrated salt solution by absorbing moisture from the glovebox atmosphere. Rapid corrosion is attributed to the ensuing salt-catalyzed reaction between plutonium and water. Experimental results are discussed, possible involvement of hydrogen fluoride (HF) is examined, and methods of corrective action are presented in this report

  17. Process control and safeguards system plutonium inventory conrol for MOX fuel facility

    International Nuclear Information System (INIS)

    Mishima, T.; Aoki, M.; Muto, T.; Amanuma, T.

    1979-01-01

    The plutonium inventory control (PINC) system is a real-time material accountability control system that is expected to be applied to a new large-scale plutonium fuel production facility for both fast breeder reactor and heavy water reactor at the Power Reactor and Nuclear Development Corporation. The PINC is basically a system for material control but is expected to develop into a whole facility control system, including criticality control, process control, quality control, facility protection, and so forth. Under PINC, every process and storage area is divided into a unit area, which is the smallest unit for both accountability and process control. Item and material weight automatically are accounted for at every unit area, and data are simultaneously treated by a computer network system. Sensors necessary for the system are being developed. 9 figures

  18. Impurity diffusion of cobalt in plutonium

    International Nuclear Information System (INIS)

    Charissoux, Christian; Calais, Daniel.

    1975-01-01

    The sectioning method for investigation of the impurity diffusion of 60 Co in the fcc and bcc phases of plutonium gives the following results: 344-426 deg C: D=1.2x10 -2 exp(-12700/RT)cm 2 /s in delta Pu(fcc); 484-621 deg C: D=1.4x10 -3 exp(-9900/RT)cm 2 /s in epsilon Pu(bcc). Cobalt is a fast diffuser in plutonium; the diffusion coefficient being unaffected by phase changes delta'→delta; delta'→epsilon, the diffusion mechanism is probably dissociative in both phases, the solute becoming interstitial by: solute (substitution) reversible solute (interstitial) + vacancy [fr

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

    Energy Technology Data Exchange (ETDEWEB)

    Venault, L

    1998-07-01

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

  20. Study of an automatic dosing of neptunium in the industrial process of separation neptunium 237-plutonium 238

    International Nuclear Information System (INIS)

    Ros, Pierre

    1973-01-01

    The objective is to study and to adapt a method of automatic dosing of neptunium to the industrial process of separation and purification of plutonium 238, while taking the information quality and economic aspects into account. After a recall of some generalities on the production of plutonium 238, and the process of separation plutonium-neptunium, the author addresses the dosing of neptunium. The adopted measurement technique is spectrophotometry (of neptunium, of neptunium peroxide) which is the most flexible and economic to adapt to automatic control. The author proposes a project of chemical automatic machine, and discusses the complex (stoichiometry, form) and some aspects of neptunium dosing (redox reactions, process control) [fr

  1. The use of absorption spectroscopy of plutonium to minimize waste streams

    International Nuclear Information System (INIS)

    Vaughn, R.B.; Berg, J.; Cisneros, M.

    1997-01-01

    Through the use of absorption spectroscopy we are better able to understand the chemical reactions of plutonium and other actinide elements in solution. In many cases such an understanding can minimize the generation of waste streams by suggesting more optimal chemical conditions for separating these radioactive elements from their host matrix. Many processes are developed using an empirical approach with little understanding of what is actually taking place. One such example is the anion exchange process for Plutonium purification. Various resins have been tested in various solutions and workable outcomes have been produced. However, absorption spectroscopy provides an understanding of why ion exchange works and can determine which compounds complex best with actinides in order to obtain a more efficient and effective separations process. This presentation will touch on the chemistry involved, the spectroscopic instrumentation, and the environmental impacts. Primarily the talk will focus on the chemical technicians involvement in the day to day research, the obstacles encountered, and the environment in which this research was conducted

  2. Solubility of plutonium and waste evaporation

    International Nuclear Information System (INIS)

    Karraker, D.G.

    1993-01-01

    Chemical processing of irradiated reactor elements at the Savannah River Site separates uranium, plutonium and fission products; fission products and process-added chemicals are mixed with an excess of NaOH and discharged as a basic slurry into large underground tanks for temporary storage. The slurry is composed of base-insoluble solids that settle to the bottom of the tank; the liquid supemate contains a mixture of base-soluble chemicals--nitrates, nitrites aluminate, sulfate, etc. To conserve space in the waste tanks, the supemate is concentrated by evaporation. As the evaporation proceeds, the solubilities of some components are exceeded, and these species crystallize from solution. Normally, these components are soluble in the hot solution discharged from the waste tank evaporator and do not crystallize until the solution cools. However, concern was aroused at West Valley over the possibility that plutonium would precipitate and accumulate in the evaporator, conceivably to the point that a nuclear accident was possible. There is also a concern at SRS from evaporation of sludge washes, which arise from washing the base-insoluble solids (open-quote sludge close-quote) with ca. 1M NaOH to reduce the Al and S0 4 -2 content. The sludge washes of necessity extract a low level of Pu from the sludge and are evaporated to reduce their volume, presenting the possibility of precipitating Pu. Measurements of the solubility of Pu in synthetic solutions of similar composition to waste supernate and sludge washes are described in this report

  3. HF effect on dissociation kinetics of plutonium and neptunium complexes with 1,2-diaminocyclohexanetetraacetic acid in nitric acid solutions

    Energy Technology Data Exchange (ETDEWEB)

    Nikitina, S.A.; Stepanov, A.V.

    1982-01-01

    Dissociation kinetics of Pusup((4)) and Np sup((4)) complexes with DCTA were investigated in HNO/sub 3/ solutions in the presence of HF and arsenazo 3. It was found that HF or NaF produced a differentiating effect on the reactivity of the complexes at (HNO/sub 3/)=1-6 mol/l as well as inhibiting effect at (HNO/sub 3/)=0.01 mol/l. Conditions of the differential kinetic analysis of plutonium and neptunium in the mixture and differential spectrophotometric analysis of uranium (6) during the camouflage of neptunium (4) and plutonium (4) were determined.

  4. EDF research scenarios for closing the Plutonium cycle

    International Nuclear Information System (INIS)

    Le Mer, Joël; Garzenne, Claude; Lemasson, David

    2013-01-01

    Conclusion: → There are various solutions to plutonium fuel closure; → Natural uranium consumption is reduced: • Full generation IV fleet is obviously the most efficient; • Symbiotic fleet makes a better use of its advanced reactors. → Plutonium inventory reaches an equilibrium between 700 tons and 1150 tons. • The multi-recycling of spent MOX fuel must be a long term solution in order to reduce significantly the plutonium inventory. → Spent fuel storage is reduced when MOX spent fuel are reprocessed but sodium pools are challenging. → Fast reactors are not the only solution to use MOX spent fuel: • HCPWR is a roundabout solution: – the reduction of natural uranium is limited; – the high level waste production is high. – The reprocessing plant capacity must be increased during deployment phase → R&D must be continued to improve HCPWR design

  5. Assessment of risk due to vehicle accident for the plutonium solution transfer from H-area to F-area

    International Nuclear Information System (INIS)

    Sarrack, A.G.

    1996-09-01

    Transporting radioactive material onsite (intrasite transfers) via truck or train must be performed in a safe manner. Adequate safety is assured for each transfer, as documented in the corresponding Onsite Safety Assessment (OSA). One aspect of the OSA is to show that the package to be used for the transfer meets onsite acceptance criteria. The activity being analyzed in this report is the movement of plutonium solution with greater than 20 curies, all reasonable mitigative controls will be implemented to minimize the likelihood of an accidental release, and a probabilistic analysis will be used to evaluate the risk associated with the move. The purpose of this report is to document the evaluation of risk due to vehicle accident associated with transporting plutonium solution from H-area to F-area. Included in the report is a list of the required mitigative controls which reduce the predicted accident and release frequencies to those reported in the summary

  6. 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

  7. PLUTONIUM LOADING CAPACITY OF REILLEX HPQ ANION EXCHANGE COLUMN - AFS-2 PLUTONIUM FLOWSHEET FOR MOX

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E.; King, W.; O' Rourke, P.

    2012-07-26

    Radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the dependence of column loading performance on the feed composition in the H-Canyon dissolution process for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). These loading experiments show that a representative feed solution containing {approx}5 g Pu/L can be loaded onto Reillex{trademark} HPQ resin from solutions containing 8 M total nitrate and 0.1 M KF provided that the F is complexed with Al to an [Al]/[F] molar ratio range of 1.5-2.0. Lower concentrations of total nitrate and [Al]/[F] molar ratios may still have acceptable performance but were not tested in this study. Loading and washing Pu losses should be relatively low (<1%) for resin loading of up to 60 g Pu/L. Loading above 60 g Pu/L resin is possible, but Pu wash losses will increase such that 10-20% of the additional Pu fed may not be retained by the resin as the resin loading approaches 80 g Pu/L resin.

  8. 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)

  9. 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

    Waste (HLW) suitable for final disposal in the deep geological formations without requiring any further reprocessing treatment (once-through solution). The fuel pellets, similar to those currently employed in the commercial LWR's will come from the ceramic mixed powders technology or from Gel Supported Precipitation (GSP) microsphere process. The neutronic calculations show that commercial PWRs partially fed with Pu-inert matrix fuel and operated in a once-through cycle, have good plutonium elimination capabilities. The plutonium still remaining in the spent fuel is quality-poor, difficult to be recovered and then highly proliferation resistant. Radiotoxicity levels versus time in inert matrix spent fuel, do not show any increase with respect to standard unreprocessed spent fuel. A most relevant open issue is, however, the technological development and qualification of this new fuel. (author)

  10. 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

  11. Continuous plutonium(IV) oxalate precipitation, filtration, and calcination process. [From product streams from Redox, Purex, or Recuplex solvent extraction plants

    Energy Technology Data Exchange (ETDEWEB)

    Beede, R L

    1956-09-27

    A continuous plutonium (IV) oxalate precipitation, filtration, and calcination process has been developed. Continuous and batch decomposition of the oxalate in the filtrates has been demonstrated. The processes have been demonstrated in prototype equipment. Plutonium (IV) oxalate was precipitated continuously at room temperature by the concurrent addition of plutonium (IV) nitrate feed and oxalic acid into the pan of a modified rotary drum filter. The plutonium (IV) oxalate was calcined to plutonium dioxide, which could be readily hydrofluorinated. Continuous decomposition of the oxalate in synthetic plutonium (IV) oxalate filtrates containing plutonium (IV) oxalate solids was demonstrated using co-current flow in a U-shaped reactor. Feeds containing from 10 to 100 g/1 Pu, as plutonium (IV) nitrate, and 1.0 to 6.5 M HNO/sub 3/, respectively, can be processed. One molar oxalic acid is used as the precipitant. Temperatures of 20 to 35/sup 0/C for the precipitation and filtration are satisfactory. Plutonium (IV) oxalate can be calcined at 300 to 400/sup 0/C in a screw-type drier-calciner to plutonium dioxide and hydrofluorinated at 450 to 550/sup 0/C. Plutonium dioxide exceeding purity requirements has been produced in the prototype equipment. Advantages of continuous precipitation and filtration are: uniform plutonium (IV) oxalate, improved filtration characteristics, elimination of heating and cooling facilities, and higher capacities through a single unit. Advantages of the screw-type drier-calciner are the continuous production of an oxide satisfactory for feed for the proposed plant vibrating tube hydrofluorinator, and ease of coupling continuous precipitation and filtration to this proposed hydrofluorinator. Continuous decomposition of oxalate in filtrates offers advantages in decreasing filtrate storage requirements when coupled to a filtrate concentrator. (JGB)

  12. Hold-up monitoring system for plutonium process tanks

    International Nuclear Information System (INIS)

    Zhu Rongbao; Jin Huimin; Tan Yajun

    1994-01-01

    The development of hold-up monitoring system for plutonium process tanks and a calculation method for α activities deposited in containers and inner walls of pipe are described. The hardware of monitoring system consists of a portable HPGe detector, a φ50 mm x 60 mm NaI(Tl) detector, γ-ray tungsten collimators, ORTEC92X Spectrum Master and an AST-286 computer. The software of system includes Maestro Tm for Window3 and a PHOUP1 hold-up application software for user. The Monte-Carlo simulation calculation supported by MCNP software is performed for the probability calculation of all the unscattering γ-rays reaching to the detection positions from the source terms deposited in the complicated tanks. A measurement mean value for different positions is used to minimize the effect of heterogeneous distribution of source term. The sensitivity is better than 3.7 x 10 6 Bq/kg (steel) for a plutonium simulation source on a 3-8 mm thick steel plate surrounded by 0.8 x 10 -10 C/kg·s γ field from long-life fission products

  13. Recovery of plutonium from solvent wash solutions

    International Nuclear Information System (INIS)

    Kyser, E.A.

    1992-01-01

    A number of potential alternatives to the acid hydrolysis recovery of Pu were investigated. The most promising alternative for short-term use appears to be an anion exchange process that would eliminate the long boiling times and the multiple-pass concentration steps needed with the solvent extraction process because it separates the Pu from the dibutyl phosphate (DBP) while at the same time concentrating the Pu. However, restart of the Primary Recovery Column (PRC) to process this solution would require significant administrative effort. The original boiling recovery by acid hydrolysis followed by solvent extraction is probably the most expedient way to process the Pu-DBP-carbonate solution currently stored in tank 13.5 even with its long processing times and dilute product concentration. Anion exchange of a heat stabilized acidified solution is a more efficient process, but requires restart of the PRC. Extended-boiling acid hydrolysis or anion exchange of a heat stabilized acidified solution provide two well developed alternatives for recovery of the Pu from the tank 13.5 carbonate. Further work defining additional recovery processes is not planned at this time

  14. A rapid method of dosing plutonium in radioactive effluents; Methode de dosage rapide du plutonium dans les effluents radioactifs (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Scheidhauer, J; Messainguiral, L [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

    1961-07-01

    The plutonium is first separated by a lanthanum fluoride precipitation. The precipitated fluorides are dissolved in normal nitric acid solution in the presence of aluminium nitrate. The plutonium transformed to the tetravalent state is then extracted with thenoyltrifluoroacetone and returned to the aqueous phase with 10 N nitric acid. After evaporation on a watch glass the residue is calcined on a Meker burner and counted using a counting system fitted with a zinc sulphide scintillator. When necessary, the calcium is eliminated at the beginning of the dosage by a fluoride precipitation, the plutonium being oxidised to the valency IV. (authors) [French] Le plutonium est d'abord separe par entrainement au fluorure de lanthane. Le precipite des fluorures est remis en solution en milieu acide nitrique normal, en presence de nitrate d'aluminium. Le plutonium amene a la valence IV est alors extrait par la thenoyltrifluoroacetone et remis en phase aqueuse dans l'acide nitrique 10 N. Apres evaporation sur verre de montre, le residu est calcine sur bec Meker et compte sur un ensemble de comptage equipe d'un scintillateur au sulfure de zinc. Lorsque cela est necessaire, le calcium est elimine, au debut du dosage, par precipitation du fluorure, le plutonium etant oxyde a la valence VI. (auteurs)

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

    International Nuclear Information System (INIS)

    Boisde, G.; Mus, G.; Tachon, M.

    1985-06-01

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

  16. Preliminary process simulation and analysis of GMODS: Processing of plutonium surplus materials

    International Nuclear Information System (INIS)

    Ferrada, J.J.; Nehls, J.W. Jr.; Welch, T.D.; Giardina, J.L.; Forsberg, C.W.; Maliyekkel, A.T.

    1996-01-01

    To address growing concerns in the areas of arms control, control of fissile materials, waste management, and environment and health, the US Department of Energy is studying and evaluating various options for the control and disposal of surplus fissile materials (SFMs). One of the options under consideration is the Glass Material Oxidation and Dissolution System (GMODS) which directly converts plutonium-bearing materials such as metals, ceramics, and organics into a durable-high-quality glass for long-term storage or a waste form for disposal. This study undertook the development of a computer simulation of the GMODS process using FLOW. That computer simulation was used to perform an assessment of how GMODS would handle the treatment of plutonium, rich scrap (RS) and lead scrap (LS), and identify critical process parameters. Among the key process parameters affecting the glass formation were processing temperatures, additives, and the effects of varying them on the final product. This assessment looked at the quantity of glass produced, the quality of the final glass form, and the effect of blending different groups of the feed streams on the glass produced. The model also provided a way to study the current process assumptions and determine in which areas more experimental studies are required. The simulation showed that the glass chemistry postulated in the models is workable. It is expected that the glass chemistry assumed during the modeling process can be verified by the results of the laboratory experiments that are currently being conducted relating to the GMODS process.Further waste characterization, especially of the SFM waste streams not studied in this report, will provide more nearly accurate results and give a more detailed evaluation of the GMODS process

  17. Preliminary process simulation and analysis of GMODS: Processing of plutonium surplus materials

    Energy Technology Data Exchange (ETDEWEB)

    Ferrada, J.J.; Nehls, J.W. Jr.; Welch, T.D.; Giardina, J.L.; Forsberg, C.W. [Oak Ridge National Lab., TN (United States); Maliyekkel, A.T. [Oak Ridge Associated Universities, TN (United States)

    1996-01-02

    To address growing concerns in the areas of arms control, control of fissile materials, waste management, and environment and health, the US Department of Energy is studying and evaluating various options for the control and disposal of surplus fissile materials (SFMs). One of the options under consideration is the Glass Material Oxidation and Dissolution System (GMODS) which directly converts plutonium-bearing materials such as metals, ceramics, and organics into a durable-high-quality glass for long-term storage or a waste form for disposal. This study undertook the development of a computer simulation of the GMODS process using FLOW. That computer simulation was used to perform an assessment of how GMODS would handle the treatment of plutonium, rich scrap (RS) and lead scrap (LS), and identify critical process parameters. Among the key process parameters affecting the glass formation were processing temperatures, additives, and the effects of varying them on the final product. This assessment looked at the quantity of glass produced, the quality of the final glass form, and the effect of blending different groups of the feed streams on the glass produced. The model also provided a way to study the current process assumptions and determine in which areas more experimental studies are required. The simulation showed that the glass chemistry postulated in the models is workable. It is expected that the glass chemistry assumed during the modeling process can be verified by the results of the laboratory experiments that are currently being conducted relating to the GMODS process.Further waste characterization, especially of the SFM waste streams not studied in this report, will provide more nearly accurate results and give a more detailed evaluation of the GMODS process.

  18. 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)

  19. 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

  20. 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

  1. Annual report for FY 1976 on project AN0115A: the migration of plutonium and americium in the lithosphere

    International Nuclear Information System (INIS)

    Fried, S.; Friedman, A.M.; Hines, J.J.; Atcher, R.W.; Quarterman, L.A.; Volesky, A.

    1976-12-01

    Studies have been carried out on the migration of plutonium and americium in solutions flowing through porous and crushed rock and through fissures. The migration process can be described in terms of the surface absorption of these elements. In addition, chemical effects on the absorption have been observed. One of these effects is possibly due to the presence of a plutonium polymer that migrates at a more rapid rate than normal plutonium

  2. Volume measurement system for plutonium nitrate solution and its uncertainty to be used for nuclear materials accountancy proved by demonstration over fifteen years

    International Nuclear Information System (INIS)

    Hosoma, Takashi

    2010-10-01

    An accurate volume measurement system for plutonium nitrate solution stored in an accountability tank with dip-tubes has been developed and demonstrated over fifteen years at the Plutonium Conversion Development Facility of the Japan Atomic Energy Agency. As a result of calibrations during the demonstration, it was proved that measurement uncertainty practically achieved and maintained was less than 0.1% (systematic character) and 0.15% (random) as one sigma which was half of the current target uncertainty admitted internationally. It was also proved that discrepancy between measured density and analytically determined density was less than 0.002 g·cm -3 as one sigma. These uncertainties include effects by long term use of the accountability tank where cumulative plutonium throughput is six tons. The system consists of high precision differential pressure transducers and a dead-weight tester, sequentially controlled valves for periodical zero adjustment, dampers to reduce pressure oscillation and a procedure to correct measurement biases. The sequence was also useful to carry out maintenances safely without contamination. Longevity of the transducer was longer than 15 years. Principles and essentials to determine solution volume and weight of plutonium, measurement biases and corrections, accurate pressure measurement system, maintenances and diagnostics, operational experiences, evaluation of measurement uncertainty are described. (author)

  3. Plutonium isotopic assay of reprocessing product solutions in the KfK K-edge densitometer

    International Nuclear Information System (INIS)

    Eberle, H.; Ottmar, H.; Matussek, P.

    1985-04-01

    The KfK K-edge densiometer, designed for accurate element concentration measurements using the technique of X-ray absorptiometry at the K absorption edge, provides as an additional option the possibility to determine the isotopic composition of freshly separated plutonium from an gamma-spectrometric analysis of its self-radiation. This report describes the underlying methodology and experimental procedures for the isotopic analysis in the K-edge densitometer. The paper also presents and discusses the experimental results so far obtained from routine measurements on reprocessing product solutions. (orig.)

  4. Determination of uranium and plutonium in metal conversion products from electrolytic reduction process

    International Nuclear Information System (INIS)

    Lee, Chang Heon; Suh, Moo Yul; Joe, Kih Soo; Sohn, Se Chul; Jee, Kwang Young; Kim, Won Ho

    2005-01-01

    Chemical characterization of process materials is required for the optimization of an electrolytic reduction process in which uranium dioxide, a matrix of spent PWR fuels, is electrolytically reduced to uranium metal in a medium of LiCl-Li 2 O molten at 650 .deg. C. A study on the determination of fissile materials in the uranium metal products containing corrosion products, fission products and residual process materials has been performed by controlled-potential coulometric titration which is well known in the field of nuclear science and technology. Interference of Fe, Ni, Cr and Mg (corrosion products), Nd (fission product) and LiCl molten salt (residual process material) on the determination of uranium and plutonium, and the necessity of plutonium separation prior to the titration are discussed in detail. Under the analytical condition established already, their recovery yields are evaluated along with analytical reliability

  5. An advanced purex process based on salt-free reductants

    Energy Technology Data Exchange (ETDEWEB)

    He, Hui; Ye, Guoan; Tang, Hongbin; Zheng, Weifang; Li, Gaoliang; Lin, Rushan [China Institute of Atomic Energy, Beijing (China). Dept. of Radiochemistry

    2014-04-01

    An advanced plutonium and uranium recovery process has been established based on two organic reductants, N,N-dimethylhydroxylamine (DMHAN) and methylhydrazine (MH), as U/Pu separation reagents. This Advanced Purex process based on Organic Reductants (APOR) is composed of three cycles, including U/Pu co-decontamination/separation cycle, uranium purification cycle and plutonium purification cycle. Using DMHAN and MH as plutonium stripping reagents in the U/Pu co-decontamination/separation cycle and plutonium purification cycle, the APOR process exhibits high performance with following highlights: (1) the process is much simpler because of the elimination of Tc scrubbing operation and the supplement extraction operation, (2) high efficiency of U/Pu separation can be achieved in the first cycle, (3) plutonium product solution of high concentration can be obtained in the Pu purification cycle with a simple extraction operation instead of circumfluent extraction or evaporation of the plutonium solution. (orig.)

  6. 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

  7. 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

  8. METHOD OF PRODUCING PLUTONIUM TETRAFLUORIDE

    Science.gov (United States)

    Tolley, W.B.; Smith, R.C.

    1959-12-15

    A process is presented for preparing plutonium tetrafluoride from plutonium(IV) oxalate. The oxalate is dried and decomposed at about 300 deg C to the dioxide, mixed with ammonium bifluoride, and the mixture is heated to between 50 and 150 deg C whereby ammonium plutonium fluoride is formed. The ammonium plutonium fluoride is then heated to about 300 deg C for volatilization of ammonium fluoride. Both heating steps are preferably carried out in an inert atmosphere.

  9. Evaluation of indigenous anion exchange resins for plutonium purification

    International Nuclear Information System (INIS)

    Kumaresan, R.; Sabharwal, K.N.; Srinivasan, T.G.; Vasudeva Rao, P.R.; Thite, B.S.; Ajithlal, R.T.; Sinalkar, Nitin; Dharampurikar, G.R.; Janardhanan, C.; Michael, K.M.; Vijayan, K.; Jambunathan, U.; Dey, P.K.

    2004-01-01

    Preliminary data with pure plutonium nitrate solution indicate that indigenous anion exchange resin can be used for the purification and concentration of plutonium. However, further studies are required to be conducted on larger scale with actual plant feed solutions before arriving to final conclusions. This includes repeated loading and elution cycles studies with the same bed and evaluation of the performance after each cycle

  10. Plutonium distribution in various components of natural organic matters and their role in plutonium migration in soils

    Energy Technology Data Exchange (ETDEWEB)

    Pavlotskaya, F.I.; Goryachenkova, T.A.

    1987-09-01

    The purpose of this work was to ascertain the links of plutonium with various components of the organic matter of different types of soils and their role in its migration in soils. The test objects were typical soils of forest and forest-steppe zones: sod-podzolic, gray forest, and leached chernozem contaminated with plutonium under laboratory conditions and stored in the air-dried state for three years, as also chernozem leached from an experimental site where agricultural plants were grown for a long time. The plutonium content in the fractions isolated from the contaminated soils was determined by its direct coprecipitation with microgram quantities of cerium hydroxide. The chemical yield of plutonium was determined by adding to a solution aliquot, equal to the analysis aliquot in volume, a known amount of plutonium and by its isolation under identical conditions. The ..gamma..-radiation of plutonium isolated on a nuclear filter with a pore size less than or equal to 15 ..mu..m was measured on a Protoka type gas flow counter.

  11. 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

  12. The distribution of plutonium-241 in rodents

    International Nuclear Information System (INIS)

    Priest, N.D.

    1977-01-01

    Plutonium-241 citrate solution at pH 6.5 was injected intravenously or intraperitoneally into hamsters and rats at a dose of 50 MBq kg -1 (1.35 mCi kg -1 ). The animals were killed 1 day or 1 week later, and tissues were removed for autoradiography and radiochemical analysis. Plutonium-241 was distributed in rats in the same way as plutonium-239, and is a suitable isotope for high-resolution tissue-section autoradiography. Plutonium deposits in cells consisted of a nuclear and a cytoplasmic component. In the hamster kidney cells, the amount associated with the nucleus was about 55 per cent of the total cellular plutonium at 24 hours after injection. Six days later, it was only about 30 per cent. Plutonium deposits were also characterized in hepatocytes, in the interstitial cells of the testes, in the cells of ovarian follicles, in chondrocytes and in bone cells, including osteoblasts and osteocytes. In bone there appeared to be both an extracellular and intracellular deposit. No evidence was found of substantial incorporation of plutonium into the mineral phase of bone. (author)

  13. 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

  14. Plutonium Proliferation: The Achilles Heel of Disarmament

    International Nuclear Information System (INIS)

    Leventhal, Paul

    2001-01-01

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

  15. International plutonium policy

    International Nuclear Information System (INIS)

    1979-02-01

    As any other fissile material, the plutonium raises several utilization problems, particularly as far as diversion by sub-national groups or proliferation are concerned. The purpose of this paper is to show that these problems can be given reasonable solutions avoiding over penalties on energy production

  16. Alternative method of portable irradiation of manganese sulphate solution by an plutonium-beryllium source for manganese sulphate bath efficiency measurements

    International Nuclear Information System (INIS)

    Silva, Fellipe Souza da; Martins, Marcelo Marques; Pereira, Walsan Wagner

    2016-01-01

    This study intends to create an alternative irradiation system from a Plutonium-Beryllium source for manganese sulphate solution using the Monte Carlo code. Thus seeking to eliminate the issue of institutes that do not have reactors or particle accelerators in its infrastructure, in order to optimize and provide independence for them to carry out efficiency measurements of MnSO_4 solution in their own locality. The Monte Carlo simulations defined the technical features of this new system so that the solution reaches the maximum neutron capture by manganese in solution. (author)

  17. 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.)

  18. Proserpine - plutonium 239 - Proserpine - uranium 235 - comparison of experimental results

    International Nuclear Information System (INIS)

    Brunet, J.P.; Caizergues, R.; Clouet D'Orval, Ch.; Kremser, J.; Moret-Bailly, J.; Verriere, Ph.

    1964-01-01

    The Proserpine homogeneous reactor is constituted by a tank, 25 cm dia, 30 cm high, surrounded by a composite reflector made of beryllium oxide and graphite. In this tank can be made critical plutonium or 90 per cent enriched uranium solutions, the fissile substances being in the form of a dissolved salt. In varying the concentration of the solution, critical masses were studied as a function of the level of the liquid in the tank. The minimum critical mass is 256 ± 2 grs for plutonium and 409 ± 3 grs for uranium 235. In the range of the critical concentrations which were studied, the neutronic properties of fissionable solutions of plutonium and enriched uranium were compared for identical geometries. (authors) [fr

  19. The establishment of in-process plutonium mass equation in Rokkasho Reprocessing Plant

    International Nuclear Information System (INIS)

    Yamaya, Kosuke; Ebata, Takashi; Yamazaki, Yoshihiro; Kawai, Akio; Iwamoto, Tomonori

    2008-01-01

    At Rokkasho Reprocessing Plant (RRP), Active Test (AT) using actual spent fuels for the final confirmation of the equipment and the system has been performed toward the commercial operation. From the safeguards viewpoint, performance of material accountancy equipment is confirmed and data for evaluating parameters of the inspection equipment is obtained by making use of the AT period. RRP is applied to Near Real Time material Accountancy (NRTA). Under the NRTA scheme, the inventory at a cut-off time during process operation needs to be accounted for. There are some un-measurable inventories of plutonium in the process, which will be calculated from inventory estimation equations. The amount of these plutonium inventories calculated from the equations is so large that it is essential to improve the inventory estimation equations to be quite accurate. Therefore, correctness of the inventory estimation equations is evaluated by using process operation data obtained during AT. This paper describes the results of evaluating the inventory estimation equations by using the process operation data and the NRTA procedure under continuous operating condition as well. (author)

  20. An investigation to compare the performance of methods for the determination of free acid in highly concentrated solutions of plutonium and uranium nitrate

    International Nuclear Information System (INIS)

    Crossley, D.

    1980-08-01

    An investigation has been carried out to compare the performance of the direct titration method and the indirect mass balance method, for the determination of free acid in highly concentrated solutions of uranium nitrate and plutonium nitrate. The direct titration of free acid with alkali is carried out in a fluoride medium to avoid interference from the hydrolysis of uranium or plutonium, while free acid concentration by the mass balance method is obtained by calculation from the metal concentration, metal valency state, and total nitrate concentration in a sample. The Gran plot end-point prediction technique has been used extensively in the investigation to gain information concerning the hydrolysis of uranium and plutonium in fluoride media and in other complexing media. The use of the Gran plot technique has improved the detection of the end-point of the free acid titration which gives an improvement in the precision of the determination. The experimental results obtained show that there is good agreement between the two methods for the determination of free acidity, and that the precision of the direct titration method in a fluoride medium using the Gran plot technique to detect the end-point is 0.75% (coefficient of variation), for a typical separation plant plutonium nitrate solution. The performance of alternative complexing agents in the direct titration method has been studied and is discussed. (author)

  1. Materials measurement and accounting in an operating plutonium conversion and purification process. Phase I. Process modeling and simulation

    International Nuclear Information System (INIS)

    Thomas, C.C. Jr.; Ostenak, C.A.; Gutmacher, R.G.; Dayem, H.A.; Kern, E.A.

    1981-04-01

    A model of an operating conversion and purification process for the production of reactor-grade plutonium dioxide was developed as the first component in the design and evaluation of a nuclear materials measurement and accountability system. The model accurately simulates process operation and can be used to identify process problems and to predict the effect of process modifications

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

    International Nuclear Information System (INIS)

    Ottmar, H.; Eberle, H.; Michel-Piper, I.; Kuhn, E.; Johnson, E.

    1985-11-01

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

  3. Non-destructive assay system for uranium and plutonium in reprocessing input solutions. Hybrid K-edge/XRF Densitometer. JASPAS JC-11 final report

    International Nuclear Information System (INIS)

    Surugaya, N.; Abe, K.; Kurosawa, A.; Ikeda, H.; Kuno, Y.

    1997-05-01

    As a part of JASPAS programme, a non-radioactive assay system for the accountability of uranium and plutonium in input dissolver solutions of a spent fuel reprocessing plant, called Hybrid K-edge/XRF Densitometer, has been developed at the Tokai Reprocessing plant (TRP) since 1991. The instrument is the one of the hybrid type combined K-edge densitometry (KED) and X-ray fluorescence (XRF) analysis. The KED is used to determine the uranium concentration and the XRF is used to determine the U/Pu ratio. These results give the plutonium concentration in consequence. It is considered that the instrument has the capability of timely on-site verification for input accountancy. The instrument had been installed in the analytical hot cell at the TRP and the experiments comparing with Isotope Dilution Mass Spectrometry (IDMS) method have been carried out. As the results of measurements for the actual input solutions in the acceptance and performance tests, it was typically confirmed that the precision for determining uranium concentration by the KED was within 0.2%, whereas the XRF for plutonium performed within 0.7%. This final report summarizes the design information and performance data so as to end the JASPAS programme. (author)

  4. Plutonium Disposition Now exclamation point

    International Nuclear Information System (INIS)

    Buckner, M.R.

    1995-01-01

    A means for use of existing processing facilities and reactors for plutonium disposition is described which requires a minimum capital investment and allows rapid implementation. The scenario includes interim storage and processing under IAEA control, and fabrication into MOX fuel in existing or planned facilities in Europe for use in operating reactors in the two home countries. Conceptual studies indicate that existing Westinghouse four-loop designs can safety dispose of 0.94 MT of plutonium per calendar year. Thus, it would be possible to consume the expected US excess stockpile of about 50 MT in two to three units of this type, and it is highly likely that a comparable amount of the FSU excess plutonium could be deposed of in a few VVER-1000's. The only major capital project for this mode of plutonium disposition would be the weapons-grade plutonium processing which could be done in a dedicated international facility or using existing facilities in the US and FSU under IAEA control. This option offers the potential for quick implementation at a very low cost to the governments of the two countries

  5. Plutonium Immobilization Can Loading Conceptual Design

    Energy Technology Data Exchange (ETDEWEB)

    Kriikku, E.

    1999-05-13

    'The Plutonium Immobilization Facility will encapsulate plutonium in ceramic pucks and seal the pucks inside welded cans. Remote equipment will place these cans in magazines and the magazines in a Defense Waste Processing Facility (DWPF) canister. The DWPF will fill the canister with glass for permanent storage. This report discusses the Plutonium Immobilization can loading conceptual design and includes a process block diagram, process description, preliminary equipment specifications, and several can loading issues. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas.'

  6. Plutonium Immobilization Can Loading Conceptual Design

    International Nuclear Information System (INIS)

    Kriikku, E.

    1999-01-01

    'The Plutonium Immobilization Facility will encapsulate plutonium in ceramic pucks and seal the pucks inside welded cans. Remote equipment will place these cans in magazines and the magazines in a Defense Waste Processing Facility (DWPF) canister. The DWPF will fill the canister with glass for permanent storage. This report discusses the Plutonium Immobilization can loading conceptual design and includes a process block diagram, process description, preliminary equipment specifications, and several can loading issues. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas.'

  7. 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.)

  8. Progress on plutonium stabilization

    Energy Technology Data Exchange (ETDEWEB)

    Hurt, D. [Defense Nuclear Facilities Safety Board, Washington, DC (United States)

    1996-05-01

    The Defense Nuclear Facilities Safety Board has safety oversight responsibility for most of the facilities where unstable forms of plutonium are being processed and packaged for interim storage. The Board has issued recommendations on plutonium stabilization and has has a considerable influence on DOE`s stabilization schedules and priorities. The Board has not made any recommendations on long-term plutonium disposition, although it may get more involved in the future if DOE develops plans to use defense nuclear facilities for disposition activities.

  9. Progress on plutonium stabilization

    International Nuclear Information System (INIS)

    Hurt, D.

    1996-01-01

    The Defense Nuclear Facilities Safety Board has safety oversight responsibility for most of the facilities where unstable forms of plutonium are being processed and packaged for interim storage. The Board has issued recommendations on plutonium stabilization and has has a considerable influence on DOE's stabilization schedules and priorities. The Board has not made any recommendations on long-term plutonium disposition, although it may get more involved in the future if DOE develops plans to use defense nuclear facilities for disposition activities

  10. Project Management Plan Solution Stabilization

    International Nuclear Information System (INIS)

    SATO, P.K.

    1999-01-01

    This plan presents the overall objectives, description, justification and planning for the Plutonium Finishing Plant (PFP) Solutions Stabilization subproject. The intent of this plan is to describe how this project will be managed and integrated with other facility stabilization and deactivation activities. This plan supplements the overall integrated plan presented in the Integrated Project Management Plan (IPMP) for the Plutonium Finishing Plant Stabilization and Deactivation Project, HNF-3617. This project plan is the top-level definitive project management document for the PFP Solution Stabilization subproject. It specifies the technical, schedule, requirements and the cost baselines to manage the execution of the Solution Stabilization subproject. Any deviations to the document must be authorized through the appropriate change control process

  11. Project Management Plan Solution Stabilization

    Energy Technology Data Exchange (ETDEWEB)

    SATO, P.K.

    1999-08-31

    This plan presents the overall objectives, description, justification and planning for the Plutonium Finishing Plant (PFP) Solutions Stabilization subproject. The intent of this plan is to describe how this project will be managed and integrated with other facility stabilization and deactivation activities. This plan supplements the overall integrated plan presented in the Integrated Project Management Plan (IPMP) for the Plutonium Finishing Plant Stabilization and Deactivation Project, HNF-3617. This project plan is the top-level definitive project management document for the PFP Solution Stabilization subproject. It specifies the technical, schedule, requirements and the cost baselines to manage the execution of the Solution Stabilization subproject. Any deviations to the document must be authorized through the appropriate change control process.

  12. 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)

  13. Analytical techniques for in-line/on-line monitoring of uranium and plutonium in process solutions : a brief literature survey

    International Nuclear Information System (INIS)

    Marathe, S.G.; Sood, D.D.

    1991-01-01

    In-line/on-line monitoring of various parameters such as uranium-plutonium-fission product concentration, acidity, density etc. plays an important role in quickly understanding the efficiency of processes in a reprocessing plant. Efforts in studying and installation of such analytical instruments are going on since more than three decades with adaptation of newer methods and technologies. A review on the developement of in-line analytical instrumentation was carried out in this laboratory about two decades ago. This report presents a very short literature survey of the work in the last two decades. The report includes an outline of principles of the main techniques employed in the in-line/on-line monitoring. (author). 77 refs., 6 tabs

  14. 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.)

  15. The Macdonald and Savage titrimetric procedure scaled down to 4 mg sized plutonium samples. P. 1

    International Nuclear Information System (INIS)

    Kuvik, V.; Lecouteux, C.; Doubek, N.; Ronesch, K.; Jammet, G.; Bagliano, G.; Deron, S.

    1992-01-01

    The original Macdonald and Savage amperometric method scaled down to milligram-sized plutonium samples was further modified. The electro-chemical process of each redox step and the end-point of the final titration were monitored potentiometrically. The method is designed to determine 4 mg of plutonium dissolved in nitric acid solution. It is suitable for the direct determination of plutonium in non-irradiated fuel with a uranium-to-plutonium ratio of up to 30. The precision and accuracy are ca. 0.05-0.1% (relative standard deviation). Although the procedure is very selective, the following species interfere: vanadyl(IV) and vanadate (almost quantitatively), neptunium (one electron exchange per mole), nitrites, fluorosilicates (milligram amounts yield a slight bias) and iodates. (author). 15 refs.; 8 figs.; 7 tabs

  16. Modeling of Diffusion of Plutonium in Other Metals and of Gaseous Species in Plutonium-Based Systems

    International Nuclear Information System (INIS)

    Cooper, Bernard R.; Gayanath W. Fernando; Beiden, S.; Setty, A.; Sevilla, E.H.

    2004-01-01

    Establish standards for temperature conditions under which plutonium, uranium, or neptunium from nuclear wastes permeates steel, with which it is in contact, by diffusion processes. The primary focus is on plutonium because of the greater difficulties created by the peculiarities of face-centered-cubic-stabilized (delta) plutonium (the form used in the technology generating the waste)

  17. The research of technology and equipment for a microwave denitration process of the uranyl nitrate solution

    International Nuclear Information System (INIS)

    Bao Weimin; Wang Xuejun; Ma Xuquan; Shi Miaoyi; Zhang Zhicheng; Bao Zhu Tian.

    1991-01-01

    In order to improve the present process of converting the plutonium nitrate into oxide powder in the nuclear fuel cycle, a new conversion process for the direct denitration using microwave heating has been developed. Microwave denitration is based on intramolecular polarization of a material in electric field and has no need of a process of heat transfer during microwave heating, so that the whole material can be heated quickly and uniformly. The thermal decomposition reactions of Pu, U, Th and RE nitrate have been analyzed and compared. The uranyl nitrate solution was chosen as imitative plutonium nitrate solution. The performance parameters ε r tanδ of U, Th and RE nitrate and oxide in microwave field were measured. The data obtained show that all of them could absorb microwave energy well and cause heating decomposition reactions. The microwave denitration test unit was designed and made. Denitration tests for rare-earths nitrate and uranyl nitrate solutions were performed. It could be completed in one step that the uranyl nitrate solution was evaporated, dryed and denitrated in a vessel. The denitrated products are a porous lump and easy to scrape off from the denitration vessel. The main forms of the products UO 3 ·0.8H 2 O and U 3 O 8 which have excellent powder properties. The capacity of the denitration unit is 1.3 kg UO 3 /h. According to the experimental results the simplicity, feasibility and good repeatability of the process have been fully proved. The unit operates easily and is adaptable to conversion of nitrate in nuclear fuel cycle. (author)

  18. Separation of Americium from plutonium, Annex 3; Prilog 3: Odvajanje amercijuma od plutonijuma

    Energy Technology Data Exchange (ETDEWEB)

    Cvjeticanin, D; Milic, N; Janicijevic, P; Ratkovic, S [Institute of Nuclear Sciences Boris Kidric, Laboratorija za visoku aktivnost, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    Since there was the possibility of working with plutonium milligram quantities, it was possible to study plutonium with contents of americium, which was expected in the about two years old plutonium solutions. Method for separation of the micro quantities of americium and plutonium was needed as well as a multichannel alpha-pulse analyzer. Method for separation of americium from plutonium by thenol trifluoro-acetone (TTA) and anion exchange was adopted.

  19. 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

  20. Plutonium decontamination studies using Reverse Osmosis

    International Nuclear Information System (INIS)

    Plock, C.E.; Travis, T.N.

    1980-01-01

    Water in batches of 45 gallons each, from a creek crossing the Rocky Flats Plant, was transferred to the Reverse Osmosis (RO) laboratory for experimental testing. The testing involved using RO for plutonium decontamination. For each test, the water was spiked with plutonium, had its pH adjusted, and was then processed by RO. At a water recovery level of 87%, the plutonium decontamination factors ranged from near 100 to 1200, depending on the pH of the processed water

  1. 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

  2. Adsorption and diffusion of plutonium in soil

    International Nuclear Information System (INIS)

    Relyea, J.F.; Brown, D.A.

    1978-01-01

    The behavior of plutonium in soil--water systems was studied by measuring its apparent diffusion coefficient in the aqueous and solid phases and by finding the adsorption--desorption relationships between soil and solution. Apparent diffusion coefficients of plutonium in soil were measured using a quick-freeze method. Aqueous diffusion was studied in a capillary-tube diffusion cell. Adsorption studies were done by equilibrating a tagged soil--water mixture on a rotary shaker before centrifuging and sampling. As expected from high adsorption coefficients (Kd) (300--10,000), the apparent diffusion coefficients were low compared with normal soil cations (1.4 x 10 -8 cm 2 /sec in a sandy soil to less than 2.4 x 10 -11 cm 2 /sec in a silt loam). The Kd of plutonium in aqueous solution containing the chelate ethylenediaminetetraacetic acid (EDTA) was reduced compared with the Kd in dilute HNO 3 . As the EDTA concentration was increased, the Kd was decreased. The chelate diethylenetriaminepentaacetic acid (DTPA) reduced the Kd more than EDTA at comparable concentrations. The aqueous diffusion coefficients varied from 3.1 x 10 -7 cm 2 /sec in a solution extracted from the silt loam up to 2.7 x 10 -5 cm 2 /sec in a solution extracted from the sandy soil

  3. Separation of traces of traces of trans-plutonium elements in weight quantities of rare earths

    International Nuclear Information System (INIS)

    SORET, Christian

    1969-08-01

    The author reports the separation of trans-plutonium elements and their dosing in a mixture of fission products. In some situations dosing is performed on both rare earths and trans-plutonium elements. The chemical separation process is a chromatographic method of exchange on an anionic resin in concentrated lithium chloride. He proposes a brief overview of separation processes, describes the separation mechanism, and then reports preliminary studies of the influence of increasing quantities of rare earths and the influence of increasing heights of resin bed in order to determine the best conditions of separation. He describes the preparation of resin and of the column, the introduction of the fixing solution at the top of the column, the preparation of lithium chloride solutions. He presents the adjustment and measurement devices, and the calculation of the resin minimum volume. Results are then presented and discussed. The operation mode is addressed: devices, reagents, preparation techniques (preparation of lithium chloride solutions) [fr

  4. Investigation of plutonium (4) hydroxoformates

    International Nuclear Information System (INIS)

    Andryushin, V.G.; Belov, V.A.; Galaktionov, S.V.; Kozhevnikov, P.B.; Matyukha, V.A.; Shmidt, V.S.

    1982-01-01

    Deposition processes of plutonium (4) hydroxoformates in the system Pu(NO 3 ) 4 -HNO 3 -HCoOH-N6 4 OH-H 2 O have been studied in pH range 0.2-10.7 at total plutonium concentration in the system 100 g/l. It is shown that under the conditions plutonium (4) hydrolysis takes place with the formation of hydroxoformates. A local maximum of plutonium (4) hydroxoformate solubility in the range pH=3.8-4.8, which is evidently conditioned by the formation of soluble formate complex of plutonium in the region, is pointed out. The basic plutonium (4) formates of the composition PuOsub(x)(OH)sub(y)(COOH)sub(4-2x-y)xnHsub(2)O, where 1,3 >=x >= 0.7, 1.7 >= y >= 1.0 and n=1.5-7.0, are singled out, their thermal stability being studied. Density of the crystals and plutonium dioxide, formed during their thermal decomposition, is measured. It is established that for plutonium (4) hydroxoformates common regularities of the influence of salt composition (OH - -, CHOO - - and H 2 O-group numbers in the mulecule) on position of temperature decomposition effects and on the density of compounds, which have been previously found during the study of thorium and plutonium hydroxosalts are observed. It is shown that the density of plutonium dioxide decreases with the increase of hydration and hydrolysis degree of the initial plutonium hydroxoformate

  5. Optimisation of parameters for co-precipitation of uranium and plutonium - results of simulation studies

    International Nuclear Information System (INIS)

    Pandey, N.K.; Velvandan, P.V.; Murugesan, S.; Ahmed, M.K.; Koganti, S.B.

    1999-01-01

    Preparation of plutonium oxide from plutonium nitrate solution generally proceeds via oxalate precipitation route. In a nuclear fuel reprocessing scheme this step succeeds the partitioning step (separation of uranium and plutonium). Results of present studies confirm that it is possible to avoid partitioning step and recover plutonium and uranium as co-precipitated product. This also helps in minimising the risk of proliferation of fissile material. In this procedure, the solubility of uranium oxalate in nitric acid is effectively used. Co-precipitation parameters are optimised with simulated solutions of uranium nitrate and thorium nitrate (in place of plutonium). On the basis of obtained results a reconversion flow-sheet is designed and reported here. (author)

  6. 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

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

    International Nuclear Information System (INIS)

    Brownson, D.A.; Hanson, D.J.; Blackman, H.S.

    1993-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-06-01

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

  9. HGMF of 10-L solutions

    International Nuclear Information System (INIS)

    Larkin, K.A.

    1994-01-01

    This test plan describes the activities associated with the High Gradient Magnetic Filtration (HGMF) of plutonium-bearing solutions (10-L). The 10-L solutions were received from Argonne National Laboratories in 1972, are highly acidic, and are considered unstable. The purpose of the testing is to show that HGMF is an applicable method of removing plutonium precipitates from solution. The plutonium then can be stored safely in a solid form

  10. Modelling of uranium/plutonium splitting in purex process

    International Nuclear Information System (INIS)

    Boullis, B.; Baron, P.

    1987-06-01

    A mathematical model simulating the highly complex uranium/plutonium splitting operation in PUREX process has been achieved by the french ''Commissariat a l'Energie Atomique''. The development of such a model, which includes transfer and redox reactions kinetics for all the species involved, required an important experimental work in the field of basis chemical data acquisition. The model has been successfully validated by comparison of its results with those of specific trials achieved (at laboratory scale), and with the available results of the french reprocessing units operation. It has then been used for the design of french new plants splitting operations

  11. Plutonium microstructures. Part 1

    International Nuclear Information System (INIS)

    Cramer, E.M.; Bergin, J.B.

    1981-09-01

    This report is the first of three parts in which Los Alamos and Lawrence Livermore National Laboratory metallographers exhibit a consolidated set of illustrations of inclusions that are seen in plutonium metal as a consequence of inherent and tramp impurities, alloy additions, and thermal or mechanical treatments. This part includes illustrations of nonmetallic and intermetallic inclusions characteristic of major impurity elements as an aid to identifying unknowns. It also describes historical aspects of the increased purity of laboratory plutonium samples, and it gives the composition of the etchant solutions and describes the etching procedure used in the preparation of each illustrated sample. 25 figures

  12. Waste minimization and the goal of an environmentally benign plutonium processing facility: A strategic plan

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1994-02-01

    To maintain capabilities in nuclear weapons technologies, the Department of Energy (DOE) has to maintain a plutonium processing facility that meets all the current and emerging standards of environmental regulations. A strategic goal to transform the Plutonium Processing Facility at Los Alamos into an environmentally benign operation is identified. A variety of technologies and systems necessary to meet this goal are identified. Two initiatives now in early stages of implementation are described in some detail. A highly motivated and trained work force and a systems approach to waste minimization and pollution prevention are necessary to maintain technical capabilities, to comply with regulations, and to meet the strategic goal

  13. 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.

  14. 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

  15. Plutonium and americium behavior in coral atoll environments

    International Nuclear Information System (INIS)

    Noshkin, V.E.; Wong, K.M.; Jokela, T.A.; Brunk, J.L.; Eagle, R.J.

    1984-01-01

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

  16. Strategies for the plutonium utilization

    International Nuclear Information System (INIS)

    Zouain, D.M.; Lima, J.O.V.; Sakamoto, L.H.

    1981-11-01

    A review of the activities involving plutonium (its recycle, utilization and technological status and perspectives) is done. These informations are useful for an economic viability study for the plutonium utilization in thermal reactors (recycling) and in fast breeders reactor (FBR), trying to collect the major number of informations about these subjects. The initial phase describes the present status and projections of plutonium accumulation and requirements. Then, the technological process are described and some strategies are analyzed. (E.G.) [pt

  17. 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.

  18. 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

  19. Differential spectrophotometric determination of plutonium

    International Nuclear Information System (INIS)

    Lecat, J.

    1980-01-01

    Differential spectrophotometric method is used for determination of plutonium reduced to oxydation state III+ by ascorbic acid, at 560 nm. Concentration of solutions is 4 g/l and accuracy of the method is better than 0,3% [fr

  20. 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

  1. Pyrochemical recovery of plutonium fluoride reduction slag

    International Nuclear Information System (INIS)

    Christensen, D.C.; Rayburn, J.A.

    1983-07-01

    A process was developed for the pyrochemical recovery of plutonium from residues resulting from the PuF 4 reduction process. The process involves crushing the CaF 2 slag and dissolving it at 800 0 C in a CaCl 2 solvent. The plutonium, which exists either as finely divided metal or as incompletely reduced fluoride salt, is reduced to metal and/or allowed to coalesce as a massive button in the bottom of the reaction crucible. The recovery of plutonium in a 1-day cycle averaged 96%; all of the resulting residues were discardable

  2. Spectrochemical determination of tantalum in plutonium metal using direct current plasma emission spectrometry

    International Nuclear Information System (INIS)

    Fadeff, S.K.; Morris, W.F.

    1983-01-01

    Tantalum is determined by direct current plasma spectrometry after separation of plutonium from solution as PuF 3 . After centrifugation of the PuF 3 precipitate, a low level of plutonium remains in solution in sufficient quantity to cause spectral interferences. It is necessary to determine the plutonium by dc plasma spectrometry and apply a correction to determine low tantalum concentrations with good accuracy and precision. Tantalum can be determined down to 0.4 ppM in solution with a relative standard deviation of 10 percent. Better precision can be achieved at higher concentrations. The procedure is simple and convenient for glovebox work. 5 references, 1 figure, 1 table

  3. Criticality of mixtures of plutonium and high enriched uranium

    International Nuclear Information System (INIS)

    Grolleau, E.; Lein, M.; Leka, G.; Maidou, B.; Klenov, P.

    2003-01-01

    This paper presents a criticality evaluation of moderated homogeneous plutonium-uranium mixtures. The fissile media studied are homogeneous mixtures of plutonium and high enriched uranium in two chemical forms: aqueous mixtures of metal and mixtures of nitrate solutions. The enrichment of uranium considered are 93.2wt.% 235 U and 100wt.% 235 U. The 240 Pu content in plutonium varies from 0wt.% 240 Pu to 12wt.% 240 Pu. The critical parameters (radii and masses of a 20 cm water reflected sphere) are calculated with the French criticality safety package CRISTAL V0. The comparison of the calculated critical parameters as a function of the moderator-to-fuel atomic ratio shows significant ranges in which high enriched uranium systems, as well as plutonium-uranium mixtures, are more reactive than plutonium systems. (author)

  4. 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

  5. The chemistry of tributyl phosphate at elevated temperatures in the Plutonium Finishing Plant Process Vessels

    International Nuclear Information System (INIS)

    Barney, G.S.; Cooper, T.D.

    1994-01-01

    Potentially violent chemical reactions of the tributyl phosphate solvent used by the Plutonium Finishing Plant at the Hanford Site were investigated. There is a small probability that a significant quantity of this solvent could be accidental transferred to heated process vessels and react there with nitric acid or plutonium nitrate also present in the solvent extraction process. The results of laboratory studies of the reactions show that exothermic oxidation of tributyl phosphate by either nitric acid or actinide nitrates is slow at temperatures expected in the heated vessels. Less than four percent of the tributyl phosphate will be oxidized in these vented vessels at temperatures between 125 degrees C and 250 degrees C because the oxidant will be lost from the vessels by vaporization or decomposition before the tributyl phosphate can be extensively oxidized. The net amounts of heat generated by oxidation with concentrated nitric acid and with thorium nitrate (a stand-in for plutonium nitrate) were determined to be about -150 and -220 joules per gram of tributyl phosphate initially present, respectively. This is not enough heat to cause violent reactions in the vessels. Pyrolysis of the tributyl phosphate occurred in these mixtures at temperatures of 110 degrees C to 270 degrees C and produced mainly 1-butene gas, water, and pyrophosphoric acid. Butene gas generation is slow at expected process vessel temperatures, but the rate is faster at higher temperatures. At 252 degrees C the rate of butene gas generated was 0.33 g butene/min/g of tributyl phosphate present. The measured heat absorbed by the pyrolysis reaction was 228 J/g of tributyl phosphate initially present (or 14.5 kcal/mole of tributyl phosphate). Release of flammable butene gas into process areas where it could ignite appears to be the most serious safety consideration for the Plutonium Finishing Plant

  6. The chemistry of tributyl phosphate at elevated temperatures in the Plutonium Finishing Plant Process Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Barney, G.S.; Cooper, T.D.

    1994-06-01

    Potentially violent chemical reactions of the tributyl phosphate solvent used by the Plutonium Finishing Plant at the Hanford Site were investigated. There is a small probability that a significant quantity of this solvent could be accidental transferred to heated process vessels and react there with nitric acid or plutonium nitrate also present in the solvent extraction process. The results of laboratory studies of the reactions show that exothermic oxidation of tributyl phosphate by either nitric acid or actinide nitrates is slow at temperatures expected in the heated vessels. Less than four percent of the tributyl phosphate will be oxidized in these vented vessels at temperatures between 125{degrees}C and 250{degrees}C because the oxidant will be lost from the vessels by vaporization or decomposition before the tributyl phosphate can be extensively oxidized. The net amounts of heat generated by oxidation with concentrated nitric acid and with thorium nitrate (a stand-in for plutonium nitrate) were determined to be about -150 and -220 joules per gram of tributyl phosphate initially present, respectively. This is not enough heat to cause violent reactions in the vessels. Pyrolysis of the tributyl phosphate occurred in these mixtures at temperatures of 110{degrees}C to 270{degrees}C and produced mainly 1-butene gas, water, and pyrophosphoric acid. Butene gas generation is slow at expected process vessel temperatures, but the rate is faster at higher temperatures. At 252{degrees}C the rate of butene gas generated was 0.33 g butene/min/g of tributyl phosphate present. The measured heat absorbed by the pyrolysis reaction was 228 J/g of tributyl phosphate initially present (or 14.5 kcal/mole of tributyl phosphate). Release of flammable butene gas into process areas where it could ignite appears to be the most serious safety consideration for the Plutonium Finishing Plant.

  7. 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

  8. Plutonium helps probe protein, superconductor

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    Scientists are finding that plutonium can be a useful research tool that may help them answer important questions in fields as diverse as biochemistry and solid-state physics. This paper reports that U.S. research involving plutonium is confined to the Department of Energy's national laboratories and centers around nuclear weapons technology, waste cleanup and disposal, and health effects. But at Los Alamos National Laboratory, scientists also are using plutonium to probe the biochemical behavior of calmodulin, a key calcium-binding protein that mediates calcium-regulated processes in biological systems. At Argonne National Laboratory, another team is trying to learn how a superconductor's properties are affected by the 5f electrons of an actinide like plutonium

  9. Supported extractant membranes for americium and plutonium recovery

    International Nuclear Information System (INIS)

    Muscatello, A.C.; Navratil, J.D.; Killion, M.E.; Price, M.Y.

    1987-01-01

    Solid supported liquid membranes(SLM) are useful in transferring and concentrating americium and plutonium from nitrate solutions. Specifically, DHDECMP(dihexyl-N,N-diethylcarbamoylmethylphosphonate) supported on Accurel or Celgard polypropylene hollow fibers assembled in modular form transfers >95% of the americium and >70% of the plutonium from high nitrate (6.9 M), low acid (0.1 M) feeds into 0.25 M oxalic acid stripping solution. Membranes supporting TBP (tri-n-butylphosphate) also transfer these metal ions. Maximum permeabilities were observed to be 1 x 10 -3 cm sec -1 , similar to the values for other systems. The feed:strip volume ratio shows an inverse relationship to the fraction of metal ion transferred. Cation exchangers may be used to concentrate americium from the strip solution

  10. 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

  11. 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)

  12. 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

  13. Detection of {alpha} particles using semiconductors. Application to the control of plutonium extraction; Detection des particules {alpha} par semiconducteurs application au controle de l'extraction du plutonium

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-03-01

    A study is made of a particles produced by thick sources, using either diffused junction or surface barrier semiconductor detectors for controlling continuously the plutonium extraction process. For this, a presenting apparatus is described in which the solutions to be analyzed flow in contact with the detector protected by a thin mica membrane. A method is described which gives a precise recording of the spectra and which thus allows the separation of two or more {alpha} emitters present in the same solution. This method has been applied to the measurement of {sup 239}Pu in the the presence of {sup 241}Am with an accuracy of {+-}5 per cent. In the second part of the report is considered the detection of plutonium in solutions of {beta} - {gamma} emitting fission products. Pile-up is reduced by using a fast amplification chain associated to totally depleted thin detectors. Under these conditions a few mg of {sup 239}Pu can be detected in solutions of fission products having an activity of 100 curies/liter. A method is given for discriminating {alpha} and {beta} particles, it is based on the difference in the collection times for the charges liberated by these particles in the detector. (author) [French] On etudie la detection de particules {alpha} issues de sources epaisses par detecteurs semiconducteurs a jonction diffusee ou a barriere de surface pour le controle continu du procede d'extraction du plutonium. A cet effet on decrit un appareil presentateur dans lequel les solutions a analyser circulent au contact du detecteur protege par une membrane mince de mica. On decrit une methode qui permet par le trace precis des spectres de separer deux ou plusieurs emetteurs {alpha} presents dans une meme solution. Cette methode a ete appliquee a la mesure du {sup 239}Pu en presence de {sup 241}Am avec une precision de {+-} 5 pour cent. Dans la deuxieme partie on traite de la detection du plutonium dans des solutions de produits de fission emetteurs {beta} and {gamma}. On

  14. Precipitation process for supernate decontamination

    International Nuclear Information System (INIS)

    Lee, L.M.; Kilpatrick, L.L.

    1982-11-01

    A precipitation and adsorption process has been developed to remove cesium, strontium, and plutonium from water-soluble, high-level radioactive waste. An existing waste tank serves as the reaction vessel and the process begins with the addition of a solution of sodium tetraphenylborate and a slurry of sodium titanate to the contained waste salt solution. Sodium tetraphenylborate precipitates the cesium and sodium titanate adsorbs the strontium and plutonium. The precipitate/adsorbate is then separated from the decontaminated salt solution by crossflow filtration. This new process offers significant capital savings over an earlier ion exchange process for salt decontamination. Chemical and small-scale engineering studies with actual waste are reported. The effect of many variables on the decontamination factors and filter performance are defined

  15. 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

  16. Automation of plutonium spectrophotometry

    International Nuclear Information System (INIS)

    Perez, J.J.; Boisde, G.; Goujon de Beauvivier, M.; Chevalier, G.; Isaac, M.

    1980-01-01

    Instrumentation was designed and constructed for automatic control of plutonium by molecular absorption spectrophotometry, on behalf of the reprocessing facilities, to meet two objectives: on-line measurement, of the valency state of plutonium, on by-pass, with the measured concentration covering the process concentration range up to a few mg.l -1 ; laboratory measurement of plutonium adjusted to valency VI, with operation carried out using a preparative system meeting the required containment specifications. For this two objectives, the photometer, optical cell connections are made by optical fibers resistant to β, γ radiation. Except this characteristic the devices are different according to the quality required for result [fr

  17. Determination of plutonium isotopic ratios and total concentration by gamma ray spectrometry

    International Nuclear Information System (INIS)

    Despres, Michele.

    1980-11-01

    A non-destructive method of analysis is being investigated for the control in situ of plutonium isotopic composition and total concentration in different matrix without preliminary calibration. The plutonium isotopic composition is determined by gamma-ray spectrometry using germanium detector systems. The same apparatus is used for direct measuring of the total plutonium concentration in solutions or solids by a differential attenuation technique based on two transmitted gamma rays with energies on both sides of the k shell absorption edge of plutonium [fr

  18. Kinetics of the reaction between plutonium (4) and neptunium (4) in nitric acid solution

    Energy Technology Data Exchange (ETDEWEB)

    Koltunov, V S; Zhuravleva, G I; Marchenko, V I

    1976-01-01

    The kinetics of the oxidation of neptunium(IV) to neptunium(V) by tetravalent plutonium ions in solutions of HNO/sub 3/ + NaNO/sub 3/ at constant (..mu.. = 2) and variable (..mu.. = 0.7-2.0) ionic strengths of the solution was investigated by a spectrophotometric method. It was established that in the range of concentrations (Np(IV)) = (4.25-10.6) x 10/sup 13/; (Pu(IV)) = (2.6-3.9)x10/sup -3/ M; (H/sup +/) 0.37-1.91 M, a first order is observed with respect to the reagents, while the order of the reaction with respect to H/sup +/ ions is equal to -3. The average value of the true rate constant of the reaction is k = 27.9+-1.3 M/sup 2/xmin/sup -1/ at ..mu..=2 and 39/sup 0/C. It was shown that with increasing analytical concentration of HNO/sub 3/ and NO/sub 3//sup -/ ions (in a mixture of HNO/sub 3/ +HClO/sub 4/), the value of K decreases. On the basis of an invetigation of the dependence of the reaction rate on the temperature in the interval 31-44.8/sup 0/, we calculated the values of the energy (E = 34.6 kcal/mole), enthalpy (..delta..H* = 34 kcal/mole), free energy (..delta..F* = 19.6 kcal/mole, entropy (..delta..S* = 49 entropy units) of activation of the reaction and the formal ionic entropy of the activated complex (PuOOHNp/sup 5 +/)*, S* = -87 entropy units. A reaction mechanism including an interaction of hydrolyzed neptunium and plutonium ions as the rate-determining step was proposed and discusses. The results obtained are compared with data for this reaction in perchloric acid wolution and for other similar redox reactions.

  19. Stability with temperature of mixed uranium plutonium monocarbides

    International Nuclear Information System (INIS)

    Riglet-Martial, Ch.; Dumas, J.C.; Piron, J.P.; Gueneau, Ch.

    2008-01-01

    Full text: Among the different advanced fuel materials of concern for Generation IV systems, the mixed carbide of uranium and plutonium fuel is considered as one of the key materials for Gas Fast Reactors (GFR) systems. For purposes of optimising its fabrication process as well as its performances in various operating conditions, the losses of gaseous plutonium specially at elevated temperatures have to be controlled and minimized. The paper is therefore concerned with a parametric analysis of the stability with temperature of mixed carbides of uranium and plutonium. Previous published experimental studies have shown that mixed (U ,Pu) carbides undergo a highly incongruent sublimation at high temperatures: the vapour phase in equilibrium with the solid is mainly composed of gaseous plutonium (P Pu /P total > 99 % ) while the contribution of gaseous U and C remains very low. The composition of the system U 1-z Pu z C 1+x ' (z =Pu/(U+Pu) and x C/(U+Pu)), the temperature (T) and the expansion volume (V) of the gas are the main parameters in the loss of gaseous Pu. The calculations are carried out using the SAGE (Solgasmix Advanced Gibbs Energy) software, by assuming ideal solid solutions between UC and PuC, as well as between U 2 C 3 and Pu 2 C 3 . The validity of the model is previously tested using published equilibrium vapour pressure data. This work gives rise to a large description of the variations of Pu losses from mixed uranium plutonium carbides and leads to some basic recommendations in connection with the use of this advanced fuel materials

  20. Two-stage precipitation of plutonium trifluoride

    International Nuclear Information System (INIS)

    Luerkens, D.W.

    1984-04-01

    Plutonium trifluoride was precipitated using a two-stage precipitation system. A series of precipitation experiments identified the significant process variables affecting precipitate characteristics. A mathematical precipitation model was developed which was based on the formation of plutonium fluoride complexes. The precipitation model relates all process variables, in a single equation, to a single parameter that can be used to control particle characteristics

  1. The separation of plutonium from uranium and fission products on zirconium phosphate columns

    Energy Technology Data Exchange (ETDEWEB)

    Gal, I; Ruvarac, A [Institute of Nuclear Sciences Boris Kidric, Laboratorija za visoku aktivnost, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    In recent years special attention has been given to the ion-exchange properties of zirconium phosphate and similar compounds in aqueous solutions. These inorganic cation exchangers are stable in oxidizing media and at elevated temperatures. Their resistance to ionizing radiation makes them particularly suitable for work with radioactive solutions. On account of this we considered ir worthwhile to investigate the separation of plutonium from uranium and fission products on zirconium phosphate columns. We were interested in nitric and solutions containing macro-amounts of uranium (a few grams per litre), and micro-amounts of plutonium and long-lived fission products. To obtain a better insight into the ion-exchange behaviour of the different ionic species towards zirconium phosphate, we first determined the dependence of the distribution coefficients of uranium, plutonium and fission product cations on the aqueous nitric acid concentration. Then, taking the distribution data as a guide, we separated plutonium on small glass columns filled with zirconium phosphate and calculated the decontamination factors (author)

  2. Selective removal/recovery of RCRA metals from waste and process solutions using polymer filtration trademark technology

    International Nuclear Information System (INIS)

    Smith, B.F.

    1997-01-01

    Resource Conservation and Recovery Act (RCRA) metals are found in a number of process and waste streams at many DOE, U.S. Department of Defense, and industrial facilities. RCRA metals consist principally of chromium, mercury, cadmium, lead, and silver. Arsenic and selenium, which form oxyanions, are also considered RCRA elements. Discharge limits for each of these metals are based on toxicity and dictated by state and federal regulations (e.g., drinking water, RCRA, etc.). RCRA metals are used in many current operations, are generated in decontamination and decommissioning (D ampersand D) operations, and are also present in old process wastes that require treatment and stabilization. These metals can exist in solutions, as part of sludges, or as contaminants on soils or solid surfaces, as individual metals or as mixtures with other metals, mixtures with radioactive metals such as actinides (defined as mixed waste), or as mixtures with a variety of inert metals such as calcium and sodium. The authors have successfully completed a preliminary proof-of-principle evaluation of Polymer Filtration trademark (PF) technology for the dissolution of metallic mercury and have also shown that they can remove and concentrate RCRA metals from dilute solutions for a variety of aqueous solution types using PF technology. Another application successfully demonstrated is the dilute metal removal of americium and plutonium from process streams. This application was used to remove the total alpha contamination to below 30 pCi/L for the wastewater treatment plant at TA-50 at Los Alamos National Laboratory (LANL) and from nitric acid distillate in the acid recovery process at TA-55, the Plutonium Facility at LANL (ESP-CP TTP AL16C322). This project will develop and optimize the PF technology for specific DOE process streams containing RCRA metals and coordinate it with the needs of the commercial sector to ensure that technology transfer occurs

  3. Selective removal/recovery of RCRA metals from waste and process solutions using polymer filtration{trademark} technology

    Energy Technology Data Exchange (ETDEWEB)

    Smith, B.F. [Los Alamos National Lab., NM (United States)

    1997-10-01

    Resource Conservation and Recovery Act (RCRA) metals are found in a number of process and waste streams at many DOE, U.S. Department of Defense, and industrial facilities. RCRA metals consist principally of chromium, mercury, cadmium, lead, and silver. Arsenic and selenium, which form oxyanions, are also considered RCRA elements. Discharge limits for each of these metals are based on toxicity and dictated by state and federal regulations (e.g., drinking water, RCRA, etc.). RCRA metals are used in many current operations, are generated in decontamination and decommissioning (D&D) operations, and are also present in old process wastes that require treatment and stabilization. These metals can exist in solutions, as part of sludges, or as contaminants on soils or solid surfaces, as individual metals or as mixtures with other metals, mixtures with radioactive metals such as actinides (defined as mixed waste), or as mixtures with a variety of inert metals such as calcium and sodium. The authors have successfully completed a preliminary proof-of-principle evaluation of Polymer Filtration{trademark} (PF) technology for the dissolution of metallic mercury and have also shown that they can remove and concentrate RCRA metals from dilute solutions for a variety of aqueous solution types using PF technology. Another application successfully demonstrated is the dilute metal removal of americium and plutonium from process streams. This application was used to remove the total alpha contamination to below 30 pCi/L for the wastewater treatment plant at TA-50 at Los Alamos National Laboratory (LANL) and from nitric acid distillate in the acid recovery process at TA-55, the Plutonium Facility at LANL (ESP-CP TTP AL16C322). This project will develop and optimize the PF technology for specific DOE process streams containing RCRA metals and coordinate it with the needs of the commercial sector to ensure that technology transfer occurs.

  4. 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.)

  5. 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)

  6. Plutonium dissolution from Rocky Flats Plant incinerator ash

    International Nuclear Information System (INIS)

    Delegard, C.H.

    1985-06-01

    Rockwell Hanford Operations (Rockwell) soon will commence recovery of plutonium from Rocky Flats Plant incinerator ash. In preparation for this processing, Rockwell undertook literature and laboratory studies to identify, select and optimize plutonium dissolution methods for treating the ash. Ash reburning, followed by dissolution in nitric acid containing calcium fluoride, was selected as the processing method for the ash. Recommended values of process parameters were identified. Using the selected process, 99.5% plutonium recovery was achieved, leaving about 12.7 wt % heel residue for an equal weight composite of the three ashes tested. 15 refs., 26 figs

  7. Search for plutonium salt deposits in the plutonium extraction batteries of the Marcoule plant (1963)

    International Nuclear Information System (INIS)

    Bouzigues, H.; Reneaud, J.M.

    1963-01-01

    This report describes a method and a special apparatus making it possible to detach the insoluble plutonium salt deposits in the extraction chain of an irradiated fuel treatment plant. The process chosen allows the detection, in the extraction batteries or in the highly active chemical engineering equipment, of plutonium quantities of a few grains. After four years operation it has been impossible to detect measurable quantities of plutonium in any part of the extraction chain. The results have been confirmed by visual examinations carried out with a specially constructed endoscope. (authors) [fr

  8. Use of the Isomass 54E thermal ionisation mass spectrometer at AEE Winfrith. Part II: plutonium analysis

    International Nuclear Information System (INIS)

    Knight, A.P.

    1982-06-01

    This report describes the application of the Isomass 54E Thermal ionisation mass spectrometer for the isotopic analysis of plutonium, and gives details of the development of a method for quantitative determination of plutonium in sample solutions using isotopic dilution. A computer program for the control of the 54E is also described. Isotope dilution was used to compare results of plutonium content of twelve zebra reactor fuel pellets with results obtained by titrimetry. The Isomass 54E is shown to be capable of high precision analysis of plutonium sample solutions for both isotopic content and total plutonium concentration using an isotope dilution technique. (U.K.)

  9. Plutonium metal preparation and purification at Los Alamos, 1984

    International Nuclear Information System (INIS)

    Christensen, D.C.; Williams, J.D.; McNeese, J.A.; Fife, K.W.

    1984-01-01

    Plutonium metal preparation and purification are well established at Los Alamos. Metal is prepared by calcothermic reduction of both PuF 4 and PuO 2 . Metal is purified by halide slagging, casting, and electrorefining. The product from the production sequence is ultrapure plutonium metal. All of the processes involve high temperature operation and all but casting involve molten salt media. Development efforts are fourfold: (1) recover plutonium values from residues; (2) reduce residue generation through process improvements and changes; (3) recycle of reagents, and (4) optimize and integrate all processes into a close-loop system. Plutonium residues are comprised of oxides, chlorides, colloidal metal suspensions, and impure metal heels. Pyrochemical recovery techniques are under development to address each residue. In addition, we are looking back at each residue generation step and are making process changes to reduce plutonium content in each residue. Reagent salt is the principle media used in pyrochemical processing. The regeneration and recycle of these reagents will both reduce our waste handling and operating expense. The fourth area, process optimization, involves both existing processes and new process developments. A status of efforts in all four of these areas will be summarized

  10. Contribution to the characterization of the ideality deviation of concentrated solutions of electrolytes: application to the case plutonium and uranium (IV) nitrates

    International Nuclear Information System (INIS)

    Charrin, N.

    1999-01-01

    The purpose of this work is to establish a base of binary data referring to the plutonium and uranium nitrates (IV) activity coefficients, which will permit to take account the medium effects in the process of liquid-liquid extraction set in action during the reprocessing of irradiated combustibles in a more scrupulous way. The first chapter sticks to establish the problematic of acquisition of actinides binary data at an oxidation state (IV) linked to two characteristics of this type of electrolyte its radioactive properties and its chemical properties. Its chemical properties bring us to define the fictitious binary data and to use an approach based on the thermodynamic concept of simple solutions, on the measurements of water activity of ternary or quaternary mixtures of the actinide, in nitric acid medium and on the binary data of nitric acid. The second chapter intended to propose reliable binary data concerning nitric acid. The validation of acquisition of fictitious binary data method suggested is undertaken. The electrolyte test is the thorium nitrate (IV). The very encouraging results has determined the carrying out of this work of research in that way. The third chapter is based on the experimental acquisition of uranium and plutonium nitrates (IV) binary data. It emphasises the importance given to the preparation of the studied mixtures which characteristics, very high actinide concentrations and low acidities, make them atypical solutions and without any referenced equivalents. The last chapter describes the exploitation which was made of the established binary data. The characteristic parameters of Pu(NO 3 ) 4 and U(NO 3 ) 4 of Pitzer model and of the specific interaction theory has been appraised. Then the application of' the concept of simple solutions to the calculation of the density or quaternary mixtures like Pu(NO 3 ) 4 / UO 2 (NO 3 ) 2 /HNO 3 / H 2 O was proposed. (author)

  11. Plutonium immobilization program - Cold pour Phase 1 test results

    International Nuclear Information System (INIS)

    Hamilton, L.

    2000-01-01

    The Plutonium Immobilization Project will disposition excess weapons grade plutonium. It uses the can-in-canister approach that involves placing plutonium-ceramic pucks in sealed cans that are then placed into Defense Waste Processing Facility canisters. These canisters are subsequently filled with high-level radioactive waste glass. This process puts the plutonium in a stable form and makes it unattractive for reuse. A cold (non-radioactive) glass pour program was performed to develop and verify the baseline design for the canister and internal hardware. This paper describes the Phase 1 scoping test results

  12. Plutonium Immobilization Program - Cold pour Phase 1 test results

    International Nuclear Information System (INIS)

    Hamilton, L.

    2000-01-01

    The Plutonium Immobilization Project will disposition excess weapons grade plutonium. It uses the can-in-canister approach that involves placing plutonium-ceramic pucks in sealed cans that are then placed into Defense Waste Processing Facility canisters. These canisters are subsequently filled with high-level radioactive waste glass. This process puts the plutonium in a stable form and makes it unattractive for reuse. A cold (non-radioactive) glass pour program was performed to develop and verify the baseline design for the canister and internal hardware. This paper describes the Phase 1 scoping test results

  13. Some studies on the extraction of plutonium from phosphate containing nitric acid solutions using DBDECMP as extractant

    International Nuclear Information System (INIS)

    Sagar, V.B.; Oak, M.S.; Pawar, S.M.; Sivaramakrishnan, C.K.; Patil, S.K.

    1991-01-01

    Extraction studies have been carried out to explore the feasibility of separation of Pu(IV) from phosphate containing analytical wastes generated in the laboratory. Distribution data on the extraction of Pu(IV) from DBDECMP (di-butyl,N,N-diethylcarbamoyl methyl phosphonate) in xylene an aqueous nitric acid and its mixture with sulfuric as well as with sulfuric and phosphoric acids were obtained. Based on the data obtained, the conditions for the recovery of plutonium from such water solutions are suggested. (author) 7 refs.; 1 fig.; 3 tabs

  14. Stabilization of Plutonium in Subsurface Environments via Microbial Reduction and Biofilm Formation

    International Nuclear Information System (INIS)

    Hakim Boukhalfa; Gary A. Icopini; Sean D. Reilly; Mary P. Neu

    2007-01-01

    Plutonium has a long half-life (2.4 x 10 4 years) and is of concern because of its chemical and radiological toxicity, high-energy alpha radioactive decay. A full understanding of its speciation and interactions with environmental processes is required in order to predict, contain, or remediate contaminated sites. Under aerobic conditions Pu is sparingly soluble, existing primarily in its tetravalent oxidation state. To the extent that pentavalent and hexavalent complexes and small colloidal species form they will increase the solubility and resultant mobility from contamination sources. There is evidence that in both marine environments and brines substantial fractions of the plutonium in solution is present as hexavalent plutonyl, PuO 2 2+

  15. Reanalysis of gastrointestinal absorption factors for plutonium and other actinide elements

    International Nuclear Information System (INIS)

    Bhattacharyya, M.H.; Larsen, R.P.; Toohey, R.E.; Moretti, E.S.; Oldham, R.D.; Spaletto, M.I.; Engel, M.C.

    1981-01-01

    This project studies the gastrointestinal absorption of plutonium and other actinide elements relevant to nuclear power production, at concentrations at or below their respective maximum permissible concentrations (MPC's) in drinking water, using high specific activity isotopes. The gastrointestinal absorption of plutonium is measured in mice, rats, and dogs exposed to plutonium either via drinking water or by gavage. Plutonium concentrations are determined in liver and eviscerated carcass at 6 days (mice and rats) or 4 weeks (dogs). Administered solutions are 1 x 10 -10 M in Pu (the molar concentration at MPC for 239 Pu) and contain one of several high specific activity isotopes ( 237 Pu, 47-day half-life; 236 Pu, 2.8-year half-life; 238 Pu, 86-year half-life). Fasted mice and rats, administered plutonium solutions that are: (1) low in concentration (10- 10 M); and (2) carefully prepared to assure a given oxidation state and to avoid hydrolysis and polymes, and major policy issues. The first HEED for near-term battery energy storage systems (lead/acid, nickel/zinc, and nickel/iron) astention being paid to potential releases of radionuclides at relatively short times after disposal

  16. Analysis of civilian processing programs in reduction of excess separated plutonium and high-enriched uranium

    International Nuclear Information System (INIS)

    Persiani, P.J.

    1995-01-01

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

  17. Status summary of chemical processing development in plutonium-238 supply program

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Emory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Benker, Dennis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wham, Robert M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); DePaoli, David W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Delmau, Laetitia Helene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sherman, Steven R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-10-01

    This document summarizes the status of development of chemical processing in the Plutonium-238 Supply Program (PSP) near the end of Demonstration 1. The objective of the PSP is “to develop, demonstrate, and document a production process that meets program objectives and to prepare for its operation†(Frazier et al. 2016). Success in the effort includes establishing capability using the current infrastructure to produce Np targets for irradiation in Department of Energy research reactors, chemically processing the irradiated targets to separate and purify the produced Pu and transferring the PuO2 product to Los Alamos National Laboratory (LANL) at an average rate of 1.5 kg/y.

  18. Safeguarding the Plutonium Fuel Cycle

    International Nuclear Information System (INIS)

    Johnson, S.J.; Lockwood, D.

    2013-01-01

    In developing a Safeguards Approach for a plutonium process facility, two general diversion and misuse scenarios must be addressed: 1) Unreported batches of undeclared nuclear material being processed through the plant and bypassing the accountancy measurement points, and 2) The operator removing plutonium at a rate that cannot be detected with confidence due to measurement uncertainties. This paper will look at the implementation of international safeguards at plutonium fuel cycle facilities in light of past lessons learned and current safeguards approaches. It will then discuss technical areas which are currently being addressed as future tools to improve on the efficiency of safeguards implementation, while maintaining its effectiveness. The discussion of new improvements will include: safeguards by design (SBD), process monitoring (PM), measurement and monitoring equipment, and data management. The paper is illustrated with the implementation of international safeguards at the Rokkasho Reprocessing Plant in Japan and its accountancy structure is detailed. The paper is followed by the slides of the presentation

  19. Remote handling in the Plutonium Immobilization Project: Plutonium conversion and first stage immobilization

    International Nuclear Information System (INIS)

    Brault, J.R.

    2000-01-01

    Since the break up of the Soviet Union at the end of the Cold War, the United States and Russia have been negotiating ways to reduce their nuclear stockpiles. Economics is one of the reasons behind this, but another important reason is safeguarding these materials from unstable organizations and countries. With the downsizing of the nuclear stockpiles, large quantities of plutonium are being declared excess and must be safely disposed of. The Savannah River Site (SRS) has been selected as the site where the immobilization facility will be located. Conceptual design and process development commenced in 1998. SRS will immobilize excess plutonium in a ceramic waste form and encapsulate it in vitrified high level waste in the Defense Waste Processing Facility (DWPF) canister. These canisters will then be interred in the national repository at Yucca Mountain, New Mexico. The facility is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. This paper will discuss the first two operations

  20. 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

  1. Search for plutonium salt deposits in the plutonium extraction batteries of the Marcoule plant (1963); Recherche de depots de sels de plutonium dans les batteries d'extraction du plutonium de l'usine de Marcoule (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Bouzigues, H; Reneaud, J M [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

    1963-07-01

    This report describes a method and a special apparatus making it possible to detach the insoluble plutonium salt deposits in the extraction chain of an irradiated fuel treatment plant. The process chosen allows the detection, in the extraction batteries or in the highly active chemical engineering equipment, of plutonium quantities of a few grains. After four years operation it has been impossible to detect measurable quantities of plutonium in any part of the extraction chain. The results have been confirmed by visual examinations carried out with a specially constructed endoscope. (authors) [French] Ce rapport decrit une methode et un montage special permettant de detacher les accumulations de sels de plutonium insolubles dans les chaines d'extraction d'une usine de traitement de combustible irradie. Le procede retenu permet de reperer, dans des batteries d'extraction ou dans l'appareillage de genie chimique fortement actif, des masses de plutonium de quelques grammes. Apres quatre annees de fonctionnement, il n'a pas ete possible de deceler des quantites ponderables de plutonium en aucun endroit de la chaine d'extraction. Ces resultats ont ete confirmes par les examens visuels effectues a l'aide d'un endoscope concu specialement pour cet usage. (auteurs)

  2. Sample preparation automation for dosing plutonium in urine

    International Nuclear Information System (INIS)

    Jeanmaire, Lucien; Ballada, Jean; Ridelle Berger, Ariane

    1969-06-01

    After having indicated that dosing urinary plutonium by using the Henry technique can be divided into three stages (plutonium concentration by precipitation, passing the solution on an anionic resin column and plutonium elution, and eluate evaporation to obtain a source of which the radioactivity is measured), and recalled that the automation of the second stage has been reported in another document, this document describes the automation of the first stage, i.e. obtaining from urine a residue containing the plutonium, and sufficiently mineralized to be analyzed by means of ion exchanging resins. Two techniques are proposed, leading to slightly different devices. The different operations to be performed are indicated. The different components of the apparatus are described: beakers, hot plate stirrers, reagent circuits, a system for supernatant suction, and a control-command circuit. The operation and use are then described, and results are given

  3. Solvent extraction process development for high plutonium fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Anil Kumar, R; Selvaraj, P G; Natarajan, R; Raman, V R [Reprocessing Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1994-06-01

    The purification of high plutonium bearing irradiated fuels using 30% TBP in dodecane diluent requires precise determination of concentration profiles during steady state, transient and process upset conditions. Mathematical models have been developed and a computer code is in use for determining Pu-U concentration profiles in a solvent extraction equipment in a typical reprocessing plant. The process parameters have been optimised for recovery of U and Pu and decontamination from the fission products. This computer code is used to analyse the extraction flow sheets of fuels of two typical Pu-U compositions encountered in Indian fast breeder programme. The analysis include the effect of uncertainty in equilibrium condition prediction by the model and the variation of flows of streams during plant operation. The studies highlight the margin available to avoid second organic phase formation and adjustments required in the process flowsheet. (author). 7 refs., 7 figs., 2 tabs.

  4. The removal of plutonium contaminants from Rocky Flats Plant soil

    International Nuclear Information System (INIS)

    Sunderland, N.R.

    1987-01-01

    This research was undertaken to determine if the TRUclean process could effectively remove radioactive elements from soils other than derived coral. This is an interim report prior to the project report and discusses the outcome of the tests of the Rocky Flats Plant (RFP) soil. The soil tested contained plutonium particulates in the micron and submicron range. Volume reduction and activity removal were accomplished with an overall efficiency of greater than 90%. The TRUclean process is a very practical and economical solution to soil contamination problems at the Rocky Flats Plant

  5. Transfer of plutonium across the human gut and its urinary excretion

    International Nuclear Information System (INIS)

    Popplewell, D.S.; Ham, G.J.; McCarthy, W.; Lands, C.

    1994-01-01

    The gastrointestinal absorption of 244 Pu(IV) has been measured in three male adult volunteers. The plutonium was in citrate solution and was taken with food. The work was carried out in two stages. The first stage measured urinary plutonium excretion up to 8 or 9 d after the oral intake. The second stage commenced about six months later with an intravenous injection of plutonium citrate and measurements of the urinary plutonium excretion. Results from the two routes of intake were used to calculate the fractional absorption (f 1 ) of ingested plutonium. The f 1 values were in the range (2-9) x 10 -4 . In theory it should be possible to measure the plutonium in the volunteers' urine throughout their lives. Measurements are continuing and the results show the excretion pattern up to nearly 2 y for one subject, and 6 months for the other two volunteers. (author)

  6. Long time contamination from plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Fueloep, M; Patzeltova, N; Ragan, P [Inst. of Preventive and Clinical Medicine, Bratislava (Slovakia); Matel, L [Comenius Univ., Bratislava (Slovakia). Department of Nuclear Chemistry

    1996-12-31

    Plutonium isotopes in the organism of the patient (who had participated in the liquidation works after the Chernobyl accident; for three month he had stayed in the epicenter, where he acted as a chauffeur driving a radioactive material to the place of destination) from urine were determined. For determination of the concentration of Pu-239, Pu-240 in urine a modified radiochemical method was used. After mineralization the sample was separated as an anion-nitrate complex with contact by the anion form of the resin in the column. The resin was washed by 8 M HNO{sub 3}, the 8 M HCl with 0.3 M HNO{sub 3} for removing the other radionuclides. The solution 0.36 M HCl with 0.01 M HF was used for the elution of plutonium. Using the lanthanum fluoride technique the sample was filtrated through a membrane filter. The plutonium was detected in the dry sample. The Pu-239 tracer was used for the evaluation of the plutonium separation efficiency. The alpha spectrometric measurements were carried out with a large area silicon detector. The samples were measured and evaluated in the energy region 4.98-5.18 MeV. The detection limit of alpha spectrometry measurements has been 0.01 Bq dm{sup -3}. The concentration of plutonium in the 24-hour urine was determined three times in the quarter year intervals. The results are: 54 mBq, 63.2 mBq, 53 mBq, with average 56,7 mBq. From the results of the analyses of plutonium depositions calculated according to ICRP 54 the intake of this radionuclide for the patient was 56.7 kBq. To estimate a committed effective dose (50 years) from the intake of plutonium was used a conversion factor 6.8.10{sup -5} Sv.Bq{sup -1} (class W). So the expressed committed effective dose received from the plutonium intake is 3.8 Sv. This number is relatively high and all the effective dose will be higher, because the patient was exposed to the other radionuclides too. (Abstract Truncated)

  7. Experimental Studies on Plutonium Kinetics in Marine Biota

    International Nuclear Information System (INIS)

    Fowler, S.; Heyraud, M.; Beasley, T.M.

    1976-01-01

    Laboratory experiments were undertaken to measure plutonium flux through marine organisms and to clarify the pathways by which this important element is cycled in the marine environment. The use of a specially prepared isotope, plutonium-237, allowed measurements to be made with standard Nal(Tl) scintillation techniques. Mussels, shrimp and worms were allowed to accumulate plutonium-237 from sea water for up to 25 days. Accumulation by shrimp was relatively slow and the degree of uptake was strongly influenced by moulting. Cast moults contained large fractions of the shrimps ' plutonium content, indicating the high affinity of plutonium for surface areas. Only small amounts of the isotope in the moult are lost to water; hence, moulting is considered to be an important biological parameter in the biogeochemical cycling of plutonium. Mussels attained higher concentration factors than shrimp with most of the accumulated isotope (> 80%) located in the shell. Byssus threads often contained large fractions of the mussels' plutonium-237 content and reached concentration factors as high as 4100. Worms readily accumulated plutonium-237 in either the +4 or +6 state, reaching concentration factors of approximately 200, Retention studies indicated a relatively slow loss of plutonium-237 from all animals studied. In the case of mussels, a computed half-time for a large fraction of the animals plutonium content was of the order of 2 years. The more rapid loss from shrimp (Tb 1/2 = 1.5 months) was due principally to the large fraction of plutonium lost at moult. Food chain studies with shrimp indicated that tissue build-up via plutonium ingestion would be a slow process. Total excretion was not entirely a result of passing contaminated food through the gut; approximately 15% of the ingested plutonium was removed from the contaminated food and subsequently excreted by processes other than defaecation of labelled food. Ratios of four different plutonium isotopes used in

  8. Experimental studies on plutonium kinetics in marine biota

    International Nuclear Information System (INIS)

    Fowler, S.; Heyraud, M.; Beasley, T.M.

    1975-01-01

    Laboratory experiments were undertaken to measure plutonium flux through marine organisms and to clarify the pathways by which this important element is cycled in the marine environment. The use of a specially prepared isotope, plutonium-237, allowed measurements to be made with standard NaI(Tl) scintillation techniques. Mussels, shrimp and worms were allowed to accumulate plutonium-237 from seawater for up to 25 days. Accumulation by shrimp was relatively slow and the degree of uptake was strongly influenced by moulting. Cast moults contained large fractions of the shrimps' plutonium content, indicating the high affinity of plutonium for surface areas. Only small amounts of the isotope in the moult are lost to water; hence, moulting is considered to be an important biological parameter in the biogeochemical cycling of plutonium. Mussels attained higher concentration factors than shrimp with most of the accumulated isotope (>80%) located in the shell. Byssus threads often contained large fractions of the mussels' plutonium-237 content and reached concentration factors as high as 4100. Worms readily accumulated plutonium-237 in either the +4 or +6 state, reaching concentration factors of approximately 200. Retention studies indicated a relatively slow loss of plutonium-237 from all animals studied. In the case of mussels, a computed half-time for a large fraction of the animals' plutonium content was of the order of 2 years. The more rapid loss from shrimp (Tbsub(1/2)=1.5 months) was due principally to the large fraction of plutonium lost at moult. Food chain studies with shrimp indicated that tissue build-up via plutonium ingestion would be a slow process. Total excretion was not entirely a result of passing contaminated food through the gut; approximately 15% of the ingested plutonium was removed from the contaminated food and subsequently excreted by processes other than defaecation of labelled food. Ratios of four different plutonium isotopes used in the

  9. Plutonium-236 traces determination in plutonium-238 by α spectrometry

    International Nuclear Information System (INIS)

    Acena, M.L.; Pottier, R.; Berger, R.

    1969-01-01

    Two methods are described in this report for the determination of plutonium-236 traces in plutonium-238 by a spectrometry using semi-conductor detectors. The first method involves a direct comparison of the areas under the peaks of the α spectra of plutonium-236 and plutonium-238. The electrolytic preparation of the sources is carried out after preliminary purification of the plutonium. The second method makes it possible to determine the 236 Pu/ 238 Pu ratio by comparing the areas of the α peaks of uranium-232 and uranium-234, which are the decay products of the two plutonium isotopes respectively. The uranium in the source, also deposited by electrolysis, is separated from a 1 mg amount of plutonium either by a T.L.A. extraction, or by the use of ion-exchange resins. The report ends with a discussion of the results obtained with plutonium of two different origins. (authors) [fr

  10. Transfer of plutonium and americium to grass vegetation as a function of radionuclide solid - solution portioning in soil

    International Nuclear Information System (INIS)

    Sokolik, G.; Ovsiannikova, S.; Ivanova, T.; Leinova, S.; Kimlenka, I.; Zakharenkov, V.; Zakharenkova, N.

    2004-01-01

    The aim of investigation is to determine the main parameters influencing the plutonium and americium migration in the soil plant system including concentration factor Cf and distribution coefficient K d . The C f factor characterising the ratio of radionuclide activity concentration in the plant specie (A p , Bq/kg) and root-inhabited layer of soil (A s , Bq/kg) has been used as a measure of biological availability of TUE. The K d coefficient estimating the ratio between radionuclide activity concentration in the equilibrium solid phase (A s.ph. ) and pore solution (A sol. , Bq/l) is considered as a measure of sorption ability of soil in respect to the radionuclide. The biological availability of 239,240 Pu and 241 Am for different grass species in various mineral and organic soils of natural and agrarian systems has been studied. The soils and grass vegetation were sampled in 1994 - 2001 in Bragin, Narovla, Khoiniki districts of Belarus (12 - 53 km from ChNPP). Since plant uptake depends primarily on radionuclide portion in the pore soil solution the proper solutions were separated from the soil samples of root-inhabited layer with the method of high-speed centrifugation. 239,240 Pu and 241 Am in the samples were determined radiochemically using alpha-spectrometer ALPHA-KING 676 A. Influence of composition of soil solution on the radionuclide soil plant transfer has been analysed. The interrelationships between the concentration factor (C f ), portion of radionuclide in the soil solution and coefficient K d have been considered. The results of investigations clearly demonstrated the dependence of TUE concentration factors for meadow sedge-herbaceous association of soil sorbing complex. As a rule, C f of americium is higher than that of plutonium. Differentiating of soils according to the C f value and the forecast of grass vegetation contamination by TUE in the different periods after catastrophe has been done. The levels of various soils contamination to receive

  11. Long time contamination from plutonium

    International Nuclear Information System (INIS)

    Fueloep, M.; Patzeltova, N.; Ragan, P.; Matel, L.

    1995-01-01

    Plutonium isotopes in the organism of the patient (who had participated in the liquidation works after the Chernobyl accident; for three month he had stayed in the epicenter, where he acted as a chauffeur driving a radioactive material to the place of destination) from urine were determined. For determination of the concentration of Pu-239, Pu-240 in urine a modified radiochemical method was used. After mineralization the sample was separated as an anion-nitrate complex with contact by the anion form of the resin in the column. The resin was washed by 8 M HNO 3 , the 8 M HCl with 0.3 M HNO 3 for removing the other radionuclides. The solution 0.36 M HCl with 0.01 M HF was used for the elution of plutonium. Using the lanthanum fluoride technique the sample was filtrated through a membrane filter. The plutonium was detected in the dry sample. The Pu-239 tracer was used for the evaluation of the plutonium separation efficiency. The alpha spectrometric measurements were carried out with a large area silicon detector. The samples were measured and evaluated in the energy region 4.98-5.18 MeV. The detection limit of alpha spectrometry measurements has been 0.01 Bq dm -3 . The concentration of plutonium in the 24-hour urine was determined three times in the quarter year intervals. The results are: 54 mBq, 63.2 mBq, 53 mBq, with average 56,7 mBq. From the results of the analyses of plutonium depositions calculated according to ICRP 54 the intake of this radionuclide for the patient was 56.7 kBq. To estimate a committed effective dose (50 years) from the intake of plutonium was used a conversion factor 6.8.10 -5 Sv.Bq -1 (class W). So the expressed committed effective dose received from the plutonium intake is 3.8 Sv. This number is relatively high and all the effective dose will be higher, because the patient was exposed to the other radionuclides too. For example the determination of the rate radionuclides Am-241/Pu-239,Pu-240 was 32-36 % in the fallout after the Chernobyl

  12. Geomorphology of plutonium in the Northern Rio Grande

    Energy Technology Data Exchange (ETDEWEB)

    Graf, W.L. [Arizona Univ., Tempe, AZ (United States). Dept., of Geography

    1993-03-01

    Nearly all of the plutonium in the natural environment of the Northern Rio Grande is associated with soils and sediment, and river processes account for most of the mobility of these materials. A composite regional budget for plutonium based on multi-decadal averages for sediment and plutonium movement shows that 90 percent of the plutonium moving into the system is from atmospheric fallout. The remaining 10 percent is from releases at Los Alamos. Annual variation in plutonium flux and storage exceeds 100 percent. The contribution to the plutonium budget from Los Alamos is associated with relatively coarse sediment which often behaves as bedload in the Rio Grande. Infusion of these materials into the main stream were largest in 1951, 1952, 1957, and 1968. Because of the schedule of delivery of plutonium to Los Alamos for experimentation and weapons manufacturing, the latter two years are probably the most important. Although the Los Alamos contribution to the entire plutonium budget was relatively small, in these four critical years it constituted 71--86 percent of the plutonium in bedload immediately downstream from Otowi.

  13. Geomorphology of plutonium in the Northern Rio Grande

    International Nuclear Information System (INIS)

    Graf, W.L.

    1993-03-01

    Nearly all of the plutonium in the natural environment of the Northern Rio Grande is associated with soils and sediment, and river processes account for most of the mobility of these materials. A composite regional budget for plutonium based on multi-decadal averages for sediment and plutonium movement shows that 90 percent of the plutonium moving into the system is from atmospheric fallout. The remaining 10 percent is from releases at Los Alamos. Annual variation in plutonium flux and storage exceeds 100 percent. The contribution to the plutonium budget from Los Alamos is associated with relatively coarse sediment which often behaves as bedload in the Rio Grande. Infusion of these materials into the main stream were largest in 1951, 1952, 1957, and 1968. Because of the schedule of delivery of plutonium to Los Alamos for experimentation and weapons manufacturing, the latter two years are probably the most important. Although the Los Alamos contribution to the entire plutonium budget was relatively small, in these four critical years it constituted 71--86 percent of the plutonium in bedload immediately downstream from Otowi

  14. Plutonium working group report on environmental, safety and health vulnerabilities associated with the Department's plutonium storage. Volume 2, Appendix A: Process and protocol

    International Nuclear Information System (INIS)

    1994-09-01

    This appendix contains documentation prepared by the Plutonium ES and H Vulnerability Working Group for conducting the Plutonium ES and H Vulnerability Assessment and training the assessment teams. It has the following five parts. (1) The Project Plan describes the genesis of the project, sets forth the goals, objectives and scope, provides definitions, the projected schedule, and elements of protocol. (2) The Assessment Plan provides a detailed methodology necessary to guide the many professionals who have been recruited to conduct the DOE-wide assessment. It provides guidance on which types and forms of plutonium are to be considered within the scope of the assessment, and lays out the assessment methodology to be used. (3) The memorandum from the Project to Operations Office Managers provides the protocol and direction for participation in the assessment by external stakeholders and members of the public; and the guidance for the physical inspection of plutonium materials in storage. (4) The memorandum from the Project to the assessment teams provides guidance for vulnerability screening criteria, vulnerability evaluation and prioritization process, and vulnerability quantification for prioritization. (5) The Team Training manual was used at the training session held in Colorado Springs on April 19--21, 1994 for all members of the Working Group Assessment Teams and for the leaders of the Site Assessment Teams. The goal was to provide the same training to all of the individuals who would be conducting the assessments, and thereby provide consistency in the conduct of the assessments and uniformity in reporting of the results. The training manual in Section A.5 includes supplemental material provided to the attendees after the meeting

  15. Plutonium in nature

    International Nuclear Information System (INIS)

    Madic, C.

    1994-01-01

    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

  16. Solubility of pllutonium in alkaline salt solutions

    International Nuclear Information System (INIS)

    Hobbs, D.T.; Edwards, T.B.

    1993-01-01

    Plutonium solubility data from several studies have been evaluated. For each data set, a predictive model has been developed where appropriate. In addition, a statistical model and corresponding prediction intervals for plutonium solubility as a quadratic function of the hydroxide concentration have been developed. Because of the wide range of solution compositions, the solubility of plutonium can vary by as much as three orders of magnitude for any given hydroxide concentration and still remain within the prediction interval. Any nuclear safety assessments that depend on the maximum amount of plutonium dissolved in alkaline salt solutions should use concentrations at least as great as the upper prediction limits developed in this study. To increase the confidence in the prediction model, it is recommended that additional solubility tests be conducted at low hydroxide concentrations and with all of the other solution components involved. To validate the model for application to actual waste solutions, it is recommended that the plutonium solubilities in actual waste solutions be determined and compared to the values predicted by the quadratic model

  17. Simultaneous determinations of uranium, thorium, and plutonium in soft tissues by solvent extraction and alpha-spectrometry

    International Nuclear Information System (INIS)

    Singh, N.P.; Zimmerman, C.J.; Lewis, L.L.; Wrenn, M.E.

    1984-01-01

    A radiochemical procedure for the simultaneous determination of uranium, thorium, and plutonium, in soft tissues has been developed. The weighed amounts of tissues, spiked with 232 U, 242 Pu, and 229 th tracers, are wet ashed. Uranium, thorium, and plutonium are coprecipitated with iron as hydroxides, dissolved in concentrated HCl and the acidity adjusted to 10 M. Uranium and plutonium are extracted into 20% TLA solution in xylene, leaving thorium in the aqueous phase. Plutonium is back-extracted by reducing to the trivalent state with 0.05 M NH 4 I solution in 8 M HCl, and uranium is back-extracted with 0.1 M HCl. Thorium is extracted into 20% TLA solution from 4 M HNO 3 and back-extracted with 10 M HCl. Uranium, thorium and plutonium are electrodeposited separately onto platinum discs and counted alpha-spectrometrically using surface barrier silicon diodes and a multichannel analyzer. The method was developed using bovine liver and applied to dog and human tissues. The mean radiochemical recoveries of these actinides in different organs were better than 70%. 6 references, 2 tables

  18. Comparative behavior of americium and plutonium in wastewater

    International Nuclear Information System (INIS)

    Tsvetaeva, N.E.; Filin, V.M.; Ragimov, T.K.; Rudaya, L.Y.; Shapiro, K.Y.; Shcherbakov, B.Y.

    1986-01-01

    This paper studies the behavior of trace americium and plutoniumin wastewater fed into purification systems. Activities of the elements were determined on a semiconductive alpha-ray spectrometer. the distributio nonuniformity, or heterogeneity, of americium and plutonium per unit volume of wastewater was determined quantitatively before and after passage through filter papers. The two elements were found to be in a colloidal or pseudocolloidal state in the original wastewater sample at pH 6. On acidifying the wastewater from pH 4 to 1 M nitric acid the americium passed quantitatively into the water phase but the most plutonium remained in the colloidal or pseudocolloidal state. the plutonium also passed quantitatively into the water phase in wastewater at a 1 M nitric acid acidity but only after a prolonged (12-day) hold. A knowledge of the heterogeneity of plutonium and americium in wastewaters made it possible to quickly distinguish their state, i.e., colloidal, pseudocolloidal, or in true solution

  19. Improvement of sample preparation for input plutonium accountability measurement by isotope dilution gammy-ray spectroscopy

    International Nuclear Information System (INIS)

    Nishida, K.; Kuno, Y.; Sato, S.; Masui, J.; Li, T.K.; Parker, J.L.; Hakkila, E.A.

    1992-01-01

    The sample preparation method for the isotope dilution gamma-ray spectrometry (IDGS) technique has been further improved for simultaneously determining the plutonium concentration and isotopic composition of highly irradiated spent-fuel dissolver solutions. The improvement includes using ion-exchange filter papers (instead of resin beads, as in two previous experiments) for better separation and recovery of plutonium from fission products. The results of IDGS measurements for five dissolver solutions are in good agreement with those by mass spectrometry with ∼0.4% for plutonium concentration and ∼0.1% for 239 Pu isotopic composition. The precision of the plutonium concentration is ∼1% with a 1-h count time. The technique could be implemented as an alternative method for input accountability and verification measurements in reprocessing plants

  20. 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)

  1. Detection of {alpha} particles using semiconductors. Application to the control of plutonium extraction; Detection des particules {alpha} par semiconducteurs application au controle de l'extraction du plutonium

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-03-01

    A study is made of a particles produced by thick sources, using either diffused junction or surface barrier semiconductor detectors for controlling continuously the plutonium extraction process. For this, a presenting apparatus is described in which the solutions to be analyzed flow in contact with the detector protected by a thin mica membrane. A method is described which gives a precise recording of the spectra and which thus allows the separation of two or more {alpha} emitters present in the same solution. This method has been applied to the measurement of {sup 239}Pu in the the presence of {sup 241}Am with an accuracy of {+-}5 per cent. In the second part of the report is considered the detection of plutonium in solutions of {beta} - {gamma} emitting fission products. Pile-up is reduced by using a fast amplification chain associated to totally depleted thin detectors. Under these conditions a few mg of {sup 239}Pu can be detected in solutions of fission products having an activity of 100 curies/liter. A method is given for discriminating {alpha} and {beta} particles, it is based on the difference in the collection times for the charges liberated by these particles in the detector. (author) [French] On etudie la detection de particules {alpha} issues de sources epaisses par detecteurs semiconducteurs a jonction diffusee ou a barriere de surface pour le controle continu du procede d'extraction du plutonium. A cet effet on decrit un appareil presentateur dans lequel les solutions a analyser circulent au contact du detecteur protege par une membrane mince de mica. On decrit une methode qui permet par le trace precis des spectres de separer deux ou plusieurs emetteurs {alpha} presents dans une meme solution. Cette methode a ete appliquee a la mesure du {sup 239}Pu en presence de {sup 241}Am avec une precision de {+-} 5 pour cent. Dans la deuxieme partie on traite de la detection du plutonium dans des solutions de produits de fission emetteurs {beta} and {gamma

  2. Tracing discharges of plutonium and technetium from nuclear processing plants by ultra-sensitive accelerator mass spectrometry

    International Nuclear Information System (INIS)

    Fifield, L.K.; Hausladen, P.A.; Cresswell, R.G.; Di Tada, M.L.; Day, J.P.; Carling, R.S.; Oughton, D.H.

    1999-01-01

    Historical discharges of plutonium from the Russian nuclear processing plant at Mayak in the Urals have been traced in sediments, soils and river water using ultra-sensitive detection of plutonium isotopes by accelerator mass spectrometry (AMS). Significant advantages of AMS over other techniques are its very high sensitivity. which is presently ∼10 6 atoms (1 μBq), and its ability to determine the 240 Pu/ 239 Pu ratio. The latter is a sensitive indicator of the source of the plutonium, being very low (1-2%) for weapons grade plutonium, and higher (∼ 20%) for plutonium from civil reactors or fallout from nuclear weapons testing. Since this ratio has changed significantly over the years of discharges from Mayak, a measurement can provide important information about the source of plutonium at a particular location. Similar measurements have been performed on samples from the Kara Sea which contains a graveyard of nuclear submarines from the former Soviet Union. AMS techniques have also been developed for detection of 99 Tc down to levels of a few femtograms. This isotope is one of the most prolific fission products and has a very long half-life of 220 ka. Hundreds of kg have been discharged from the nuclear reprocessing plant at Sellafield in the UK. While there may be public health issues associated with these discharges which can be addressed with AMS, these discharges may also constitute a valuable oceanographic tracer experiment in this climatically-important region of the world's oceans. Applications to date have included a human uptake study to assess long-term retention of 99 Tc in the body, and a survey of seaweeds from northern Europe to establish a baseline for a future oceanographic study

  3. Measurements of plutonium in environmental samples

    International Nuclear Information System (INIS)

    D'Alberti, F.; Risposi, L.

    1996-01-01

    Within the activities connected with the start up of the PETRA Laboratory (Processo per l'Estrazione di Terre Rare ed Attinidi, i.e. process for extraction of rare earths and actinides), the Radiation Protection Unit of the J.R.C.-Ispra has carried out a well planned set of experimental measurements aimed at evaluating the zero point of the isotopes of plutonium in environmental samples by alfa spectrometry. After the International Moratorium in 1963, no release of plutonium has occurred in the environment apart from the burn up of SNAP 9A satellite in April 1964. Since then the plutonium concentration in air and in fallout samples has been continuously decreasing requiring, therefore, optimization of both instrumentation and experimental measurement procedures in order to obtain better sensibilities. In this work, the experimental methodology followed at the J.R.C.-Ispra for measurements of plutonium concentration in air, deposition and soil is described and the plutonium behaviour in these samples is reported and discussed starting from 1961

  4. Measurements of plutonium in environmental samples

    Energy Technology Data Exchange (ETDEWEB)

    D' Alberti, F; Risposi, L [Instituto di Fisica Applicata, University of Milan, Milan (Italy)

    1996-01-01

    Within the activities connected with the start up of the PETRA Laboratory (Processo per l'Estrazione di Terre Rare ed Attinidi, i.e. process for extraction of rare earths and actinides), the Radiation Protection Unit of the J.R.C.-Ispra has carried out a well planned set of experimental measurements aimed at evaluating the zero point of the isotopes of plutonium in environmental samples by alfa spectrometry. After the International Moratorium in 1963, no release of plutonium has occurred in the environment apart from the burn up of SNAP 9A satellite in April 1964. Since then the plutonium concentration in air and in fallout samples has been continuously decreasing requiring, therefore, optimization of both instrumentation and experimental measurement procedures in order to obtain better sensibilities. In this work, the experimental methodology followed at the J.R.C.-Ispra for measurements of plutonium concentration in air, deposition and soil is described and the plutonium behaviour in these samples is reported and discussed starting from 1961.

  5. Uptake of Plutonium-238 into Solanum tuberosum L. (potato plants) in presence of complexing agent EDTA.

    Science.gov (United States)

    Tawussi, Frank; Gupta, Dharmendra K; Mühr-Ebert, Elena L; Schneider, Stephanie; Bister, Stefan; Walther, Clemens

    2017-11-01

    Bioavailability and plant uptake of radionuclides depend on various factors. Transfer into different plant parts depends on chemical and physical processes, which need to be known for realistic ingestion dose modelling when these plants are used for food. Within the scope of the present work, the plutonium uptake by potato plants (Solanum tuberosum L.) was investigated in hydroponic solution of low concentration [Pu] = 10 -9  mol L -1 . Particular attention was paid to the speciation of radionuclides in the solution which was modelled by the speciation code PHREEQC. The speciation, the solubility and therefore the plant availability of radionuclides mainly depend on the pH value and the redox potential of the solution. During the contamination period, the redox potential did not change significantly. In contrast, the pH value showed characteristic changes depending on exudates excreted by the plants. Plant roots took up high amounts of plutonium (37%-50% of the added total amount). In addition to the uptake into the roots, the radionuclides can also adsorb to the exterior root surface. The solution-to-plant transfer factor showed values between 0.03 and 0.80 (Bq kg -1 / Bq L -1 ) for the potato tubers. By addition of the complexing agent EDTA (10 -4  mol L-1), the plutonium uptake from solution increased by 58% in tubers and by 155% in shoots/leaves. The results showed that excreted substances by plants affect bioavailability of radionuclides at low concentration, on the one hand. On the other hand, the uptake of plutonium by roots and the accumulation in different plant parts can lead to non-negligible ingestion doses, even at low concentration. We are aware of the limited transferability of data obtained in hydroponic solutions to plants growing in soil. However, the aim of this study is twofold: First we want to investigate the influence of Pu speciation on plant uptake in a rather well defined system which can be modelled using available thermodynamic data

  6. Plutonium Finishing Plant (PFP) Final Safety Analysis Report (FSAR) [SEC 1 THRU 11

    Energy Technology Data Exchange (ETDEWEB)

    ULLAH, M K

    2001-02-26

    The Plutonium Finishing Plant (PFP) is located on the US Department of Energy (DOE) Hanford Site in south central Washington State. The DOE Richland Operations (DOE-RL) Project Hanford Management Contract (PHMC) is with Fluor Hanford Inc. (FH). Westinghouse Safety Management Systems (WSMS) provides management support to the PFP facility. Since 1991, the mission of the PFP has changed from plutonium material processing to preparation for decontamination and decommissioning (D and D). The PFP is in transition between its previous mission and the proposed D and D mission. The objective of the transition is to place the facility into a stable state for long-term storage of plutonium materials before final disposition of the facility. Accordingly, this update of the Final Safety Analysis Report (FSAR) reflects the current status of the buildings, equipment, and operations during this transition. The primary product of the PFP was plutonium metal in the form of 2.2-kg, cylindrical ingots called buttoms. Plutonium nitrate was one of several chemical compounds containing plutonium that were produced as an intermediate processing product. Plutonium recovery was performed at the Plutonium Reclamation Facility (PRF) and plutonium conversion (from a nitrate form to a metal form) was performed at the Remote Mechanical C (RMC) Line as the primary processes. Plutonium oxide was also produced at the Remote Mechanical A (RMA) Line. Plutonium processed at the PFP contained both weapons-grade and fuels-grade plutonium materials. The capability existed to process both weapons-grade and fuels-grade material through the PRF and only weapons-grade material through the RMC Line although fuels-grade material was processed through the line before 1984. Amounts of these materials exist in storage throughout the facility in various residual forms left from previous years of operations.

  7. Spectroscopy of plutonium-organic complexes

    International Nuclear Information System (INIS)

    Richmann, M.K.; Reed, D.T.

    1995-01-01

    Information on the spectroscopy of plutonium-organic complexes is needed to help establish the speciation of these complexes under environmentally relevant conditions. Laser photoacoustic spectroscopy (LPAS) and absorption spectrometry were used to characterize the Pu(IV)-citrate and Pu(IV)-nitrilotriacetic acid (NTA) complexes at concentrations of 10 -3 --10 -7 M in aqueous solution. Good agreement was observed between the band shape of the LPAS and absorption spectra for the Pu(IV)-NTA complex. Agreement for the Pu(IV)-citrate complex was not quite as good. In both cases, a linear dependence of the LPAS signal on laser power and total concentration of the complexes was noted. This work is part of an ongoing research effort to study key subsurface interactions of plutonium-organic complexes

  8. Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Plutonium Metals, Oxides, and Solutions on the High Performance Computing Platform Moonlight

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, Bryan Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-12-05

    This report documents a validation of the MCNP6 Version 1.0 computer code on the high performance computing platform Moonlight, for operations at Los Alamos National Laboratory (LANL) that involve plutonium metals, oxides, and solutions. The validation is conducted using the ENDF/B-VII.1 continuous energy group cross section library at room temperature. The results are for use by nuclear criticality safety personnel in performing analysis and evaluation of various facility activities involving plutonium materials.

  9. Dosage of plutonium by isotopic dilution in irradiated fuels; Dosage du plutonium par dilution isotopique dans les combustibles irradies

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    Plutonium determination in irradiated fuels has been carried out for several years by isotopic dilution by Sebaci and SSM in collaboration. SECACI has made available to the SSM the necessary space and equipment in its Fontenay laboratories. This work has shown the importance of the valency cycle which should make it possible to obtain a uniform isotopic distribution in sample tracer mixtures, and also a satisfactory U/Pu separation. Now it has been noticed that the presence of an excess of uranium considerably modifies the oxidation-reduction reaction kinetics of the plutonium. We have therefore been led to change certain parts of the operational technique so as to have an efficient cycle and to thereby improve the U/Pu separation; the stability of the thermionic emission of the plutonium, connected to the quantity of residual uranium, has at the same time been improved and we can now carry out more precise isotopic analyses. We have also tried to eliminate as far as possible the isotopic contaminations by:using a more rational operational method; the equipment used has been the object of a special study. The evaporations are carried out so as to prevent the formation of saturated vapours inside the glove box. The material which cannot be changed after each operation is carefully cleaned every time a new sample is treated. With this technique, a second calibration of the tracer T{sub 2} has been undertaken using a new standard solution. This solution has been prepared very carefully by weighing uranium and plutonium of known chemical purity, and we believe that it can be guaranteed to be a good reference solution. The value of the {sup 233}U/{sup 242}Pu ratio of the tracer has been obtained with a relative accuracy of 0,5 per cent. This modified method is now being applied to the analysis of rods irradiated in G-3. (author) [French] La determination du plutonium par dilution isotopique dans les combustibles irradies est pratiquee depuis plusieurs annees en

  10. Managing plutonium in Britain. Current options

    International Nuclear Information System (INIS)

    1998-01-01

    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 were 'currently involved in a formal

  11. 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

  12. Erosional losses of fallout plutonium

    International Nuclear Information System (INIS)

    Foster, G.R.; Hakonson, T.E.

    1987-01-01

    Plutonium from fallout after atmospheric explosion of nuclear weapons in the 1950's and 1960s is being redistributed over the landscape by soil erosion and carried on sediment by streams to oceans. Erosion rates computed with the Universal Soil Loss Equation for more than 200,000 sample points on nonfederal land across the US were used to estimate plutonium removal rates by soil erosion. On the average, only about 4% of the eroded sediment reaches the outlet of a major river. The remaining sediment is deposited en route, and because deposition is a selective process, the sediment is enriched in fine particles having the highest concentration of plutonium because of the element's strong association with clay and silt-sized sediment. Estimated enrichment ratios, sediment delivery ratios, and erosion rates were used to estimate annual delivery of fallout plutonium. These estimates ranged from 0.002% of the initial fallout plutonium inventory for the Savannah River basin to 0.01% for the Columbia River basin, to 0.02% for the Hudson and Rio Grande River basins, to 0.08% for the Mississippi River basin. If the deposition of plutonium had been uniformly 1 mCi/km 2 , the estimated plutonium activity on suspended sediment would range from about 7 fCi/g of sediment of the Savannah River basin, to 9 fCi/g for the Mississippi River basin, to 12 fCi/g for the Hudson River basin, to 14 fCi/g for the Columbia and Rio Grande River basins. 45 references, 2 figures, 17 tables

  13. Los Alamos DP West Plutonium Facility decontamination project

    International Nuclear Information System (INIS)

    Garde, R.; Cox, E.J.; Valentine, A.M.

    1982-01-01

    The DP West Plutonium Facility operated by the Los Alamos National Laboratory, Los Alamos, New Mexico, was decontaminated between April 1978 and April 1981. The facility was constructed in 1944 to 1945 to produce plutonium metal and fabricate parts for nuclear weapons. It was continually used as a plutonium processing and research facility until mid-1978. Decontamination operations included dismantling and removing gloveboxes and conveyor tunnels; removing process systems, utilities, and exhaust ducts; and decontaminating all remaining surfaces. This report describes glovebox and conveyor tunnel separations, decontamination techniques, health and safety considerations, waste management procedures, and costs of the operation

  14. Nondestructive analysis of plutonium contaminated soil

    International Nuclear Information System (INIS)

    Smith, H.E.; Taylor, L.H.

    1977-01-01

    Plutonium contaminated soil is currently being removed from a covered liquid waste disposal trench near the Pu Processing facility on the Hanford Project. This soil with the plutonium is being mined using remote techniques and equipment. The mined soil is being packaged for placement into retrievable storage, pending possible recovery. To meet the requirements of criticality safety and materials accountability, a nondestructive analysis program has been developed to determine the quantity of plutonium in each packing-storage container. This paper describes the total measurement program: equipment systems, calibration techniques, matrix assumption, instrument control program and a review of laboratory operating experience

  15. 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)

  16. Plutonium

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    Plutonium, which was obtained and identified for the first time in 1941 by chemist Glenn Seaborg - through neutron irradiation of uranium 238 - is closely related to the history of nuclear energy. From the very beginning, because of the high radiotoxicity of plutonium, a tremendous amount of research work has been devoted to the study of the biological effects and the consequences on the environment. It can be said that plutonium is presently one of the elements, whose nuclear and physico-chemical characteristics are the best known. The first part of this issue is a survey of the knowledge acquired on the subject, which emphasizes the sanitary effects and transfer into the environment. Then the properties of plutonium related to energy generation are dealt with. Fissionable, like uranium 235, plutonium has proved a high-performance nuclear fuel. Originally used in breeder reactors, it is now being more and more widely recycled in light water reactors, in MOX fuel. Reprocessing, recycling and manufacturing of these new types of fuel, bound of become more and more widespread, are now part of a self-consistent series of operations, whose technical, economical, industrial and strategical aspects are reviewed. (author)

  17. Learning more about plutonium

    International Nuclear Information System (INIS)

    2005-01-01

    This document offers chemical, metallurgical and economical information on the plutonium, a hard white radioelement. It deals also on the plutonium formation in the earth, the plutonium use in the nuclear industry, the plutonium in the environment and the plutonium toxicity. (A.L.B.)

  18. 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.)

  19. Processing plutonium-contaminated soild for volume reduction using the segmented gate system

    International Nuclear Information System (INIS)

    Moroney, K.S.; Moroney, J.D.; Turney, J.M.; Doane, R.W.

    1994-01-01

    TMA/Eberline has developed and demonstrated an effective method for removing mixed plutonium and americium contamination from a coral soil matrix at the Defense Nuclear Agency's Johnston Atoll site. TMA's onsite soil processing for volume reduction is ongoing at a rate of over 2000 metric tons per week. The system uses arrays of sensitive radiation detectors coupled with sophisticated computer software developed by Eberline Instrument Corporation. The proprietary software controls four soil sorting units operating in parallel that utilize TMA's unique Segmented Gate System technology to remove radiologically contaminated soil from a moving supply on conveyor belts. Clean soil is released for use elsewhere on the island. Contaminated soil is diverted to either a metal drum for collecting higher activity open-quotes hotclose quotes particles (>5000 Becquerels), or to a supplementary soil washing process designed to remove finely divided particles of dispersed low level contamination. Site contamination limits specify maximum dispersed radioactivity of no more than 500 Becquerels per kilogram of soil averaged over no more than 0.1 cubic meter. Results of soil processing at this site have been excellent. After processing over 50,000 metric tons, the volume of contaminated material that would have required expensive special handling, packaging, and disposal as radioactive waste has been successfully reduced by over 98 percent. By mid-January 1994, nearly three million kiloBecquerels of plutonium/americium contamination had been physically separated from the contaminated feed by TMA's Segmented Gate System, and quality control sampling showed no radioactivity above release criteria in the open-quotes cleanclose quotes soil pile

  20. Appraisal of BWR plutonium burners for energy centers

    International Nuclear Information System (INIS)

    Williamson, H.E.

    1976-01-01

    The design of BWR cores with plutonium loadings beyond the self-generation recycle (SGR) level is investigated with regard to their possible role as plutonium burners in a nuclear energy center. Alternative plutonium burner approaches are also examined including the substitution of thorium for uranium as fertile material in the BWR and the use of a high-temperature gas reactor (HTGR) as a plutonium burner. Effects on core design, fuel cycle facility requirements, economics, and actinide residues are considered. Differences in net fissile material consumption among the various plutonium-burning systems examined were small in comparison to uncertainties in HTGR, thorium cycle, and high plutonium-loaded LWR technology. Variation in the actinide content of high-level wastes is not likely to be a significant factor in determining the feasibility of alternate systems of plutonium utilization. It was found that after 10,000 years the toxicity of actinide high-level wastes from the plutonium-burning fuel cycles was less than would have existed if the processed natural ores had not been used for nuclear fuel. The implications of plutonium burning and possible future fuel cycle options on uranium resource conservation are examined in the framework of current ERDA estimates of minable uranium resources

  1. Containers for short-term storage of nuclear materials at the Los Alamos plutonium facility

    International Nuclear Information System (INIS)

    Hagan, R.; Gladson, J.

    1997-01-01

    The Los Alamos Plutonium Facility for the past 18 yr has stored nuclear samples for archiving and in support of nuclear materials research and processing programs. In the past several years, a small number of storage containers have been found in a deteriorated condition. A failed plutonium container can cause personnel contamination exposure and expensive physical area decontamination. Containers are stored in a physically secure radiation area vault, making close inspection costly in the form of personnel radiation exposure and work time. A moderate number of these containers are used in support of plutonium processing and must withstand daily handling abuse. A 2-yr evaluation of failed containers and those that have shown no deterioration has been conducted. Based on that study, a program was established to formalize our packing methods and materials and standardize the size and shape of containers that are used for short-term use. A standardized set of containers was designed, evaluated, tested, and procured for use in the facility. This paper reviews our vault storage problems, shows some failed containers, and presents our planned solutions to provide safe and secure containment of nuclear materials

  2. Acid-digestion treatment of plutonium-containing waste

    International Nuclear Information System (INIS)

    Wieczorek, H.; Kemmler, G.; Krause, H.

    1981-01-01

    The Radioactive Acid-Digestion Test Unit (RADTU) has been constructed at Hanford to demonstrate the application of the acid-digestion process for treating combustible transuranic wastes and scrap materials. The RADTU, with its original tray digestion vessel, has recently completed a six-month campaign processing potentially contaminated non-glovebox wastes from a Hanford plutonium facility. During this campaign, it processed 2100 kg largely cellulosic wastes at an average sustained processing rate of 3 kg/h as limited by the acid-waste contact and the water boil-off rate from the acid feeds. The on-line operating efficiency was nearly 50% on a twelve-hour day, five-day week basis. Following this campaign, a new annular high-rate digester has been installed for testing. In preliminary tests with simulated wastes, the new digester demonstrated a sustained capacity of 10 kg/h with greatly improved intimacy of contact between the digestion acid and the waste. The new design also doubles the heat-transfer surface, which is expected to provide at least twice the water boil-off rate of the previous tray digester design. Following shakedown testing with simulated and low-level wastes, the new unit will be used to process combustible plutonium scrap and waste from Hanford plutonium facilities for the purposes of volume reduction, plutonium recovery, and stabilization of the final waste form. (author)

  3. 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.)

  4. Contribution to the study of the process of purification of plutonium by extraction with trilaurylamine

    International Nuclear Information System (INIS)

    Saey, Jean-Claude

    1966-01-01

    This work addresses the process of plutonium purification which uses trilaurylamine nitrate. In order to use this nitrate in its solid state and at ordinary temperature, a secondary solvent must be added which must have some properties: low volume mass and viscosity, high boiling and ignition temperatures, rather low miscibility with water, high stability in front of joint actions of nitric acid and radiations, and no reaction with the alkylammonium nitrate and the complex. Thus, the author addresses phenomena of immiscibility and identifies some important molecular characteristics which could lead to the selection of another secondary solvent than dodecane. The decalin seem interesting and its behaviour is studied. A mixing of dodecane and decalin is used as extraction mixing. The obtained results are discussed. Finally, the author notices that using this mixing in the plutonium purification process results in a large increase of metal concentrations and a decrease of risks of crystallisation, without any major drawback, in a continuously operating micro-industrial installation

  5. Design of an integrated non-destructive plutonium assay facility

    International Nuclear Information System (INIS)

    Moore, C.B.

    1984-01-01

    The Department of Energy requires improved technology for nuclear materials accounting as an essential part of new plutonium processing facilities. New facilities are being constructed at the Savannah River Plant by the Du Pont Company, Operating Contractor, to recover plutonium from scrap and waste material generated at SRP and other DOE contract processing facilities. This paper covers design concepts and planning required to incorporate state-of-the-art plutonium assay instruments developed at several national laboratories into an integrated, at-line nuclear material accounting facility operating in the production area. 3 figures

  6. Alecto, criticality experiment on a plutonium solution. Experimental results. Vessel number 1 (φ = 324 mm)

    International Nuclear Information System (INIS)

    Bruna, J.; Brunet, J.F.; Caizergues, R.; Clouet D'orval, C.; Kremser, J.; Leclerc, J.; Verriere, P.

    1963-01-01

    ALECTO is a critical experiment intended for the neutronic study of homogeneous aqueous multiplying media. It essentially consists of a cylindrical tank, reflected or not, where can be made critical a solution of fissionable material fed into the tank from a geometrically subcritical storage. The studies effected on this assembly concern on one hand the determination of critical masses, on the other hand the nuclear parameters used in neutron calculations. The container tested in the first series of experiments hereby described is a cylindrical tank, 324 mm diameter with a convex bottom, water reflected on the sides and on the inferior part. The minimum critical mass of this tank was determined and was found to be: M cmin = 845 ± 7 g. The decay constant of prompt neutrons as a function of reactivity was determined by the pulsed neutron technique. At the critical state, it was found to be: α c = 73 ± 6 s -1 . Furthermore, from the study of this tank, were derived a number of safety regulations for plutonium solutions. (authors) [fr

  7. 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)

  8. 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)

  9. Thermal expansion and transformation behavior of cerium and plutonium alloys: an application of the Aptekar-Ponyatovsky regular solution model.

    Science.gov (United States)

    Lawson, A C; Lashley, J C

    2011-09-14

    In this paper we apply the Aptekar-Ponyatovsky (AP) regular solution thermodynamic model to the analysis of experimental data for the coefficient of thermal expansion (CTE) and determine the AP model parameters for unalloyed cerium metal, Ce-Th-La alloys, and Pu-Ga alloys. We find that the high temperature CTE of cerium metal follows the predictions of the AP model based on low temperature, high pressure data. For Ce-Th-La alloys we use the AP parameters to track the suppression of the first-order γ-α cerium transition. We show the AP model accounts for the negative CTE observed for Pu-Ga alloys and is equivalent to an earlier invar model. Finally, we apply the AP parameters obtained for Pu-Ga alloys to rationalize the observed δ-α transformation pressures of these alloys. We show that the anomalous values of the Grüneisen and Grüneisen-Anderson parameters are important features of the thermal properties of plutonium. A strong analogy between the properties of plutonium and cerium is confirmed.

  10. Photochemical technique for reduction of uranium and subsequently plutonium in the Purex process

    International Nuclear Information System (INIS)

    Goldstein, M.; Barker, J.J.; Gangwer, T.

    1976-09-01

    A photochemical modification of the Purex process is described in which a purified side stream of UO 2 ++ ion is reduced to U +4 outside the radioactive area of the reprocessing plant. The U +4 is then cycled back to step 2 of the Purex process to reduce the plutonium and effect separation within the partitioning column. This process is shown to be very energy efficient and compatible with existing conventional lamp technology. Preliminary cost estimates of the energy requirements for photon production are essentially negligible. Conceptual systems and photochemical reactor designs are presented. Potential benefits of this system are discussed

  11. Natural hazards that may trigger a radiological release from a plutonium processing facility

    Energy Technology Data Exchange (ETDEWEB)

    Selvidge, J. E.

    1977-04-28

    Calculations show the probability of a tornado striking a plutonium area at Rocky Flats is 2.2 x 10/sup -4/ per year. The source term (expected value of plutonium release) should such an event occur is calculated at 3.3 x 10/sup -7/ grams. The source term for high-velocity, downslope winds is higher--2.2 x 10/sup -3/ grams. The probability of a meteorite that weighs one or more pounds (453 grams) striking a plutonium area is estimated at 8.88 x 10/sup -7/ per year. Because of this small probability and the remote chance that a plutonium release would occur even if a meteorite hit occurred, the hazard from meteorite impact is considered negligible. Conservative assumptions result in all calculated frequencies being almost certainly too high. Empirical observations have indicated lower frequencies than those calculated.

  12. Natural hazards that may trigger a radiological release from a plutonium processing facility

    International Nuclear Information System (INIS)

    Selvidge, J.E.

    1977-01-01

    Calculations show the probability of a tornado striking a plutonium area at Rocky Flats is 2.2 x 10 -4 per year. The source term (expected value of plutonium release) should such an event occur is calculated at 3.3 x 10 -7 grams. The source term for high-velocity, downslope winds is higher--2.2 x 10 -3 grams. The probability of a meteorite that weighs one or more pounds (453 grams) striking a plutonium area is estimated at 8.88 x 10 -7 per year. Because of this small probability and the remote chance that a plutonium release would occur even if a meteorite hit occurred, the hazard from meteorite impact is considered negligible. Conservative assumptions result in all calculated frequencies being almost certainly too high. Empirical observations have indicated lower frequencies than those calculated

  13. Characteristics of airborne plutonium resuspended from near-background aged surface-sources

    International Nuclear Information System (INIS)

    Sehmel, G.A.

    1982-11-01

    Plutonium content in samples of airborne solids collected at five Hanford sites was determined in several experiments directed toward investigating resuspension processes for aged surface sources. Though airborne plutonium concentrations are extremely low, radiochemical technique sensitivities allow plutonium characterization to be considered as a function of host-particle diameter in samples of airborne solids. Plutonium concentrations and activity densities are a function of aerodynamic particle diameter, sampling height, wind speed, wind direction and plutonium isotopic ratios

  14. The plutonium society

    International Nuclear Information System (INIS)

    Mez, L.; Richter, M.

    1981-01-01

    The lectures of an institute are reported on, which took place between 25th and 27th January 1980 in Berlin. The subsequent public panel discussion with representations from the political parties is then documentated in a few press-reports. The themes of the 8 lectures are: views and facts on plutonium, plutonium as an energy resource, military aspects of the production of plutonium, economic aspects of the plutonium economy, the position of the trade unions on the industrial reconversion, the alleged inevitability of a plutonium society and the socio-political alternatives and perspectives of nuclear waste disposal. (UA) [de

  15. Purification process of uranium hexafluoride containing traces of plutonium fluoride and/or neptunium fluoride

    International Nuclear Information System (INIS)

    Aubert, J.; Bethuel, L.; Carles, M.

    1983-01-01

    In this process impure uranium hexafluoride is contacted with a metallic fluoride chosen in the group containing lead fluoride PbF 2 , uranium fluorides UFsub(4+x) (0 3 at a temperature such as plutonium and/or neptunium are reduced and pure uranium hexafluoride is recovered. Application is made to uranium hexafluoride purification in spent fuel reprocessing [fr

  16. Measurement and regulation of the level of a homogeneous plutonium reactor; Mesure et regulation du niveau d'un reacteur homogene au plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Berger, F; Bertrand, J

    1958-12-01

    Reactivity depends strongly on disturbances of the level of the plutonium solution In the homogeneous reactor. Proserpine has a small cylindrical core, 250 mm diameter, and 10 liters volume. With a view to reducing the dangers due to corrosion and contamination, the solution level in the core is raised by pneumatic pressure. The level is stabilized by means of a regulating system. During critical experiments the variations of the level are less than one hundredth part of a millimeter. (author) [French] Les variations du niveau de la solution de plutonium dans le reacteur homogene Proserpine ont une grosse influence sur la reactivite, car le coeur est petit (10 litres de solution dans un cylindre de diametre 250 mm). En vue de reduire les dangers dus a la corrosion et a la contamination, la commande du volume liquide est pneumatique. Nous avons realise la stabilite du niveau par une regulation qui, dans les essais en regime critique, limite les variations du plan liquide a une fraction de centieme de millimetre. (auteur)

  17. What is plutonium stabilization, and what is safe storage of plutonium?

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1995-01-01

    The end of the cold war has resulted in the shutdown of nuclear weapons production and the start of dismantlement of significant numbers of nuclear weapons. This, in turn, is creating an inventory of plutonium requiring interim and long-term storage. A key question is, ''What is required for safe, multidecade, plutonium storage?'' The requirements for storage, in turn, define what is needed to stabilize the plutonium from its current condition into a form acceptable for interim and long-term storage. Storage requirements determine if research is required to (1) define required technical conditions for interim and long-term storage and (2) develop or improve current stabilization technologies. Storage requirements depend upon technical, policy, and economic factors. The technical issues are complicated by several factors. Plutonium in aerosol form is highly hazardous. Plutonium in water is hazardous. The plutonium inventory is in multiple chemical forms--some of which are chemically reactive. Also, some of the existing storage forms are clearly unsuitable for storage periods over a few years. Gas generation by plutonium compounds complicates storage: (1) all plutonium slowly decays creating gaseous helium and (2) the radiation from plutonium decay can initiate many chemical reactions-some of which generate significant quantities of gases. Gas generation can pressurize sealed storage packages. Last nuclear criticality must be avoided

  18. Isotope dilution alpha spectrometry for the determination of plutonium concentration in irradiated fuel dissolver solution : IDAS and R-IDAS

    International Nuclear Information System (INIS)

    Ramaniah, M.V.; Jain, H.C.; Aggarwal, S.K.; Chitambar, S.A.; Kavimandan, V.D.; Almaula, A.I.; Shah, P.M.; Parab, A.R.; Sant, V.L.

    1980-01-01

    The report presents a new technique, Isotope Dilution Alpha Spectrometry (IDAS) and Reverse Isotope Dilution Alpha Spectrometry (R-IDAS) for determining the concentration of plutonium in the irradiated fuel dissolver solution. The method exploits sup(238)Pu in IDAS and sup(239)Pu in R-IDAS as a spike and provides an alternative method to Isotope Dilution Mass Spectrometry (IDMS) which requires enriched sup(242)Pu as a spike. Depending upon the burn-up of the fuel, sup(238)Pu or sup(239)Pu is used as a spike to change the sup(238)Pu/(sup(239)Pu+sup(240)Pu)α activity ratio in the sample by a factor of 10. This change is determined by α-spectrometry on electrodeposited sources using a solid state silicon surface barrier detector coupled to a multichannel analyser. The validity of a simple method based on the geometric progression (G.P.) decrease for the far tail of the spectrum to correct for the tail contribution of sup(238)Pu peak (5.50 MeV) to the low energy sup(239)Pu + sup(240)Pu peak (5.17 MeV) is established. Results for the plutonium concentration on different irradiated fuel dissolver solutions with burn-uo ranging from J,000 to 100,000 MWD/TU are presented and compared with those obtained by IDMS. The values obtained by IDAS or R-IDAS and IDMS agree within 0.5%. (auth.)

  19. Status of plutonium ceramic immobilization processes and immobilization forms

    International Nuclear Information System (INIS)

    Ebbinghaus, B.B.; Van Konynenburg, R.A.; Vance, E.R.; Jostsons, A.

    1996-01-01

    Immobilization in a ceramic followed by permanent emplacement in a repository or borehole is one of the alternatives currently being considered by the Fissile Materials Disposition Program for the ultimate disposal of excess weapons-grade plutonium. To make Pu recovery more difficult, radioactive cesium may also be incorporated into the immobilization form. Valuable data are already available for ceramics form R ampersand D efforts to immobilize high-level and mixed wastes. Ceramics have a high capacity for actinides, cesium, and some neutron absorbers. A unique characteristic of ceramics is the existence of mineral analogues found in nature that have demonstrated actinide immobilization over geologic time periods. The ceramic form currently being considered for plutonium disposition is a synthetic rock (SYNROC) material composed primarily of zirconolite (CaZrTi 2 O 7 ), the desired actinide host phase, with lesser amounts of hollandite (BaAl 2 Ti 6 O 16 ) and rutile (TiO 2 ). Alternative actinide host phases are also being considered. These include pyrochlore (Gd 2 Ti 2 O 7 ), zircon (ZrSiO 4 ), and monazite (CePO 4 ), to name a few of the most promising. R ampersand D activities to address important technical issues are discussed. Primarily these include moderate scale hot press fabrications with plutonium, direct loading of PuO 2 powder, cold press and sinter fabrication methods, and immobilization form formulation issues

  20. In-plant measurements of gamma-ray transmissions for precise K-edge and passive assay of plutonium concentration and isotopic abundance in product solutions at the Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    Asakura, Y.; Kondo, I.; Masui, J.; Shoji, K.; Russo, P.A.; Hsue, S.T.; Sprinkle, J.K. Jr.; Johnson, S.S.

    1982-01-01

    A field test has been carried out for more than 2 years for determination of plutonium concentration by K-edge absorption densitometry and for determination of plutonium isotopic abundance by transmission-corrected passive gamma-ray spectrometry. This system was designed and built at Los Alamos National Laboratory and installed at the Tokai reprocessing plant of the Power Reactor and Nuclear Fuel Development Corporation as a part of the Tokai Advanced Safeguards Technology Exercise (TASTEX). For K-edge measurement of plutonium concentration, the transmissions at two discrete gamma-ray energies are measured using the 121.1- and 122.1-keV gamma rays from 75 Se and 57 Co. Intensities of the plutonium passive gamma rays in the energy regions between 38 and 51 keV and between 129 and 153 keV are used for determination of the isotopic abundances. More than 200 product solution samples have been measured in a timely fashion during these 2 years. The relative precisions and accuracies of the plutonium concentration measurement are shown to be within 0.6% (1 sigma) in these applications, and those for plutonium isotopic abundances are within 3% for 238 Pu, 0.4% for 239 Pu, 1.2% for 240 Pu, 1.3% for 241 Pu, and 7% for 242 Pu. The time required is 10 min for the concentration assay, 10 min for the isotopics assay, and about 15 min for handling procedures in the laboratory

  1. Fast Thorium Molten Salt Reactors Started with Plutonium

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Brissot, R.; Liatard, E.; Meplan, O.; Nuttin, A.; Mathieu, L.

    2006-01-01

    One of the pending questions concerning Molten Salt Reactors based on the 232 Th/ 233 U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since 233 U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing 233 U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce 233 U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/ 233 U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into 233 U. A particular reactor configuration is used, called 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum, allowing Plutonium to be used as fissile matter. The conversion capacities of such Molten Salt Reactors are excellent. For Molten Salt Reactors only started with Plutonium, the assets of the Thorium fuel cycle turn out to be quickly recovered and the reactor's characteristics turn out to be equivalent to Molten Salt Reactors operated with 233 U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with 233 U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)

  2. 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.)

  3. Comparative behavior of plutonium and americium in the equatorial Pacific

    International Nuclear Information System (INIS)

    Noshkin, V.E.; Wong, K.M.; Jokela, T.A.; Brunk, J.L.; Eagle, R.J.

    1983-01-01

    Inventories of 239 + 240 Pu and 241 Am greatly in excess of global fallout levels persist in the benthic environments of Bikini and Enewetak Atolls. The amount of 239 + 240 Pu mobilized to solution at the atolls can be predicted from a distribution coefficient K/sub d/ of 2.3 x 10 5 and the mean sediment concentrations. The mobilized 239 + 240 Pu has solute-like characteristics and different valence states coexist in solution - the largest fraction of the soluble plutonium is in an oxidized form (+V,VI). The adsorption of plutonium to sediments is not completely reversible because of changes that occur in the relative amounts of the mixed oxidation states in solution with time. Characteristics of 239 + 240 Pu described at one location may not necessarily describe its behavior elsewhere. The relative amounts of 241 Am to 239 + 240 Pu may be altered in future years because of mobilization and radiological decay

  4. 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

  5. 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)

  6. Plutonium, proliferation, and the price of reprocessing

    International Nuclear Information System (INIS)

    Gilinsky, V.

    1978-01-01

    France and Britain disagree with the US on whether deferring fuel reprocessing that provides plutonium for export can help contain proliferation. The US has veto power over reprocessing of US-supplied fuels for non-EURATOM countries, but exceptions will be made for movement within the EURATOM community. Political issues will be influenced by the magnitude of the financial investments, however, and commercial considerations have until recently dominated and complicated international safeguards. The author notes that US policy was reversed by the gradual acknowledgment that the same international inspection of plutonium stockpiles would not work as it had for low-enriched fuel and that economic interests must have a lower priority to avoiding proliferation. He cites the combination of sudden policy shifts, failure to prove that present reactors are best, and long-term distrust of US economic motives as failing to persuade either the French or British, who feel the best safeguard is provided by their high-security reprocessing facilities. Still to be resolved are the conditions under which plutonium must be returned to its owners, a problem that must determine safe international transport and storage and international management. Technical fixes, such as the CIVEX process, cannot contribute to the solution for several decades, while reprocessing is no longer considered a first step in waste disposal and would be more expensive and complicated than present waste disposal procedures. The author concedes merit in President Carter's requirement of separating ''the legitimate and necessary use of uranium'' and nuclear fuels that are also explosives

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

    International Nuclear Information System (INIS)

    Patterson, M.W.; Thompson, R.J.

    1994-01-01

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

  8. Plutonium in uranium deposits

    International Nuclear Information System (INIS)

    Curtis, D.; Fabryka-Martin, J.; Aguilar, R.; Attrep, M. Jr.; Roensch, F.

    1992-01-01

    Plutonium-239 (t 1/2 , 24,100 yr) is one of the most persistent radioactive constituents of high-level wastes from nuclear fission power reactors. Effective containment of such a long-lived constituent will rely heavily upon its containment by the geologic environment of a repository. Uranium ore deposits offer a means to evaluate the geochemical properties of plutonium under natural conditions. In this paper, analyses of natural plutonium in several ores are compared to calculated plutonium production rates in order to evaluate the degree of retention of plutonium by the ore. The authors find that current methods for estimating production rates are neither sufficiently accurate nor precise to provide unambiguous measures of plutonium retention. However, alternative methods for evaluating plutonium mobility are being investigated, including its measurement in natural ground waters. Preliminary results are reported and establish the foundation for a comprehensive characterization of plutonium geochemistry in other natural environments

  9. Disposal criticality analysis for immobilized plutonium: Internal configurations

    International Nuclear Information System (INIS)

    Gottlieb, P.; Massari, J.R.; Cloke, P.L.

    1998-03-01

    The analysis for immobilized Pu follows the disposal criticality analysis methodology. In this study the focus is on determining the range of chemical compositions of the configurations which can occur following the aqueous degradation processes, particularly with respect to the concentrations of uranium, plutonium, and the principal neutron absorber, gadolinium. The principal analysis tool is a mass balance program that computes the amounts of plutonium, uranium, gadolinium, and chromium in solution as a function of time with inputs from a range of possible waste form dissolution rates, stainless steel corrosion rates, and compound solubilities for the neutronically significant elements. For the waste forms and degradation modes considered here, it is possible to preclude the possibility of criticality by maintaining a plutonium loading limit. Since the presence of hafnium is shown to increase this loading limit, the defense-in-depth policy would suggest the maximization of the amount of Hf as a backup criticality control material. At the end of 1997, after this study was completed, the ceramic waste form was downselected and a new formulation was developed, with the amount of Hf increased to the point where internal criticality may no longer be possible. In addition, recent calculations indicate that GdPO 4 is insoluble over a much broader range of pH than is Gd 2 O 3 , so that its use as the Gd carrier in the waste form would provide an extra margin of defense-in-depth

  10. Fabrication of inert matrix fuel for the incineration of plutonium - a feasibility study

    International Nuclear Information System (INIS)

    Burghartz, M.; Ledergerber, G.; Ingold, F.; Xie, T.; Botta, F.; Idemitsu, K.

    1998-01-01

    The internal gelation process has been applied to fabricate classical fuel based on uranium like UO 2 and MOX. For recent aims to destroy plutonium in the most effective way, a uranium free fuel was evaluated. The fuel development at PSI has been redirected to a fuel based on zirconium oxide or a mixture of zirconia and a conducting material leading to ceramic/metal (CERMET) or ceramic/ceramic (CERCER) combinations. A feasibility study was carried out to demonstrate that microspheres based on zirconia and spinel can be fabricated. The gelation parameters were investigated leading to optimised compositions for the starting solutions. Studies to fabricate a composite material (from zirconia and spinel) are ongoing. If the zirconia/spinel ratio is chosen appropriately, the low thermal conductivity of pure zirconia could be compensated by the higher thermal conductivity of spinel. Another solution to improve the low thermal conductivity of zirconia is the development of a CERMET, which consists of fine particles bearing plutonium in a cubic zirconia dispersed in a metallic matrix. The fabrication of such a CERMET is also being studied. (author)

  11. 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)

  12. The use of plutonium

    International Nuclear Information System (INIS)

    Marshall, W.

    1980-01-01

    The use of plutonium as a vital energy source producing maximum economic benefit with minimum proliferation risks is discussed. Having considered the production of plutonium, several possible plutonium fuel cycle options are identified and the economic value to be attached to plutonium for each examined. It is shown how the use of plutonium in fast reactors gives an opportunity for a non-proliferation policy not available when plutonium is used only in thermal reactors. From the technical considerations reviewed concerning plutonium and fast reactors it is shown that an economic regime involving international trade in spent thermal reactor fuel is possible which benefits equally those countries with fast reactors and those without and also assists in avoiding the proliferation of nuclear weapons. (U.K.)

  13. 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

  14. High-temperature enthalpies of plutonium monocarbide and plutonium sesquicarbide

    International Nuclear Information System (INIS)

    Oetting, F.L.

    1979-01-01

    The high-temperature enthalpies of plutonium monocarbide and plutonium sesquicarbide have been determined with a copper-block calorimeter of the isoperibol type. The experimental enthalpy data, which was measured relative to 298 K, covered the temperature range from 400 to 1500 K. The calculation of the temperature rise of the calorimeter takes into account the added heat evolution from the radioactive decay of the plutonium samples. These enthalpy results, combined with the heat capacity and entropy of the respective carbide at 298 K available from the literature, has made it possible to generate tables of thermodynamic functions for the plutonium carbides. The behavior of the heat capacity of both of the plutonium carbides, i.e., a relatively steep increase in the heat capacity as the temperature increases, may be attributed to a premelting effect with the formation of vacancies within the crystal lattice although a theoretical treatment of this phenomenon is not given

  15. Inert atmosphere system for plutonium processing gloveboxes

    International Nuclear Information System (INIS)

    Bogard, C.F.; Calkins, K.W.; Rogers, R.F.

    1975-01-01

    Recent efforts to reduce fire hazards in plutonium processing operations are described. In such operations, the major environmental controls are developed through various kinds of glovebox systems. In evaluating the air-atmosphere glovebox systems, formerly in use at Rocky Flats and many other plants, a decision was made to convert to a recirculating ''inert'' atmosphere. The inert atmosphere consists of nitrogen, supplied from an on-site generating plant, diluting oxygen content to one to 5 percent by volume. Problems encountered during the change over included: determination of all factors influencing air leakage into the system, and reducing leakage to the practical minimum; meeting all fire and safety standards on the filter plenum and exhaust systems; provision for converting portions of the system to an air atmosphere to conduct maintenance work; inclusion of oxygen analyzers throughout the system to check gas quality and monitor for leaks; and the use of automatic controls to protect against a variety of potential malfunctions. The current objectives to reduce fire hazards have been met and additional safeguards were added. The systems are operating satisfactorily. (U.S.)

  16. Study of plutonium cycle in marine ecosystems

    International Nuclear Information System (INIS)

    Merino Pareja, J.; Sanchez Cabeza, J. A.; Molero Savall, J.; Masque Barri, P.

    1998-01-01

    The distribution, transport and accumulation mechanisms of transuranics (and other radionuclides) in the marine environment depend on the source term, biogeochemical cycles, transport with the water masses, sedimentation processes and transfer mechanisms in the trophic chain. The biogeochemical behaviour of plutonium, which has been the focus of our work, was studied using the following approaches: determination of the physico-chemical speciation of plutonium in marine waters, vertical flux in the water column, uptake by marine organisms (phytoplankton and zooplankton) and distribution in dements cores. A preliminary model of the accumulation and distribution of plutonium in the first levels of the marine food chain in the Irish Sea has also been formulated. All this information allowed us to obtain an integrated view of the behaviour of plutonium in the marine environment. (Author) 14 refs

  17. Extraction of plutonium from lean residues by room-temperature fluoride volatility

    International Nuclear Information System (INIS)

    Campbell, G.M.; Foropoulos, J.; Kennedy, R.C.; Dye, B.A.; Behrens, R.G.

    1990-01-01

    The use of dioxygen difluoride and krypton difluoride for the recovery of plutonium from lean residues by conversion to gaseous plutonium hexa-fluoride is being investigated. The synthesis of dioxygen difluoride in practical quantity has been demonstrated. Fluorination of plutonium compounds under ideal conditions supports the contention that a viable process can be developed. Application of the method to lean plutonium residues is in the early stage of development

  18. Mechanism of plutonium metal dissolution in HNO3-HF-N2H4 solution

    International Nuclear Information System (INIS)

    Karraker, D.G.

    1985-01-01

    An oxidation-reduction balance of the products of the dissolution of plutonium metal and alloys in HNO 3 -HF-N 2 H 4 solution shows that the major reactions during dissolution are the reduction of nitrate to NH 3 , N 2 and N 2 O by the metal, and the oxidation of H free radicals to NH 3 by N 2 H 4 . Reactions between HNO 3 and N 2 H 4 produce varying amounts of HN 3 . The reaction rate is greater for delta-Pu than alpha-Pu, and is increased by higher concentrations of HF and HNO 3 . The low yield of reduced nitrogen species indicates that nitrate is reduced on the metal surface without producing a significant concentration of species that react with N 2 H 4 . It is conjectured that intermediate Pu valences and electron transfer within the metal are involved. 7 refs., 3 tabs

  19. Western Option - Disarmament of Russian Weapon Plutonium

    International Nuclear Information System (INIS)

    Tveiten, B.; Petroll, M.R.

    2002-01-01

    The Western Option concept describes an approach to the conversion of weapon-grade plutonium from Russian nuclear warheads under the special aspects of meeting the criteria of irreversible utilization. Putting this concept of plutonium conversion into non-weapon-grade material into effect would make a major contribution to improving security worldwide. This study is based on an agreement between the Russian Federation and the United States of America concluded in September 2000. It provides for the conversion of 34 t of weapon-grade plutonium in each of the two states. This goal is also supported by other G8 countries. While the United States performs its part of the agreement under its sole national responsibility, the Russian program needs financial support by Western states. Expert groups have pointed out several options as a so-called basic scenario. The funds of approx. US Dollar 2 billion required to put them into effect have not so far been raised. The Western Option approach described in this contribution combines results of the basic scenario with other existing experience and with technical solutions available for plutonium conversion. One of the attractions of the Western Option lies in its financial advantages, which are estimated to amount to approx. US Dollar 1 billion. (orig.) [de

  20. Extrinsic and intrinsic complexities of the Los Alamos plutonium processing facility

    International Nuclear Information System (INIS)

    Bearse, R.C.; Roberts, N.J.; Longmire, V.L.

    1985-01-01

    Analysis of the data obtained in one year of plutonium accounting at Los Alamos reveals significant complexity. Much of this complexity arises from the complexity of the processes themselves. Additional complexity is induced by errors in the data entry process. It is important to note that there is no evidence that this complexity is adversely affecting the accounting in the plant. The authors have been analyzing transaction data from fiscal year 1983 processing. This study involved 62,595 transactions. The data have been analyzed using the relational database program INGRES on a VAX 11/780 computer. This software allows easy manipulation of the original data and subsets drawn from it. The authors have been attempting for several years to understand the global features of the TA-55 accounting data. This project has underscored several of the system's complexities

  1. Study of the potential use of carburized niobium in plutonium processing

    International Nuclear Information System (INIS)

    Johnson, M.J.

    1998-01-01

    Carburized refractory metals, especially tantalum, have been shown to possess properties useful for application as hardware in the plutonium-processing environment. These applications are driven in part by a desire to minimize the production of radioactively contaminated waste. The current use of ceramics as containment materials for Pu processing are not ideal due to the short service life of the hardware, placing an additional burden on the contaminated waste stream. Carburized niobium has been examined for use as an improved hardware material. The Nb-C system is analogous to the previously studied Ta-C system. The low density of niobium relative to tantalum will improve the ergonomics of the glovebox environment. The choice of the Nb-C system will be supported by a thermodynamic and kinetic analysis. Preliminary results of the processing investigation also will be presented

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

    International Nuclear Information System (INIS)

    Araujo, Izilda da Cruz de

    1995-01-01

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

  3. Lixiviation of plutonium contaminated solid wastes by aqueous solution of electro-generated reducing agents

    International Nuclear Information System (INIS)

    Agarande, Michelle

    1991-01-01

    This study concerns the development of the new concept for the decontamination of plutonium bearing solid wastes, based on the lixiviation of the wastes using electro-generated reducing agents. First, a comparative study of the kinetics of the dissolution of pure PuO 2 (prepared by calcination of Pu (IV) oxalate at 450 C) in sulfuric acid media, with different reducing agents, was realized. Qualitatively these reagents can be sorted in three groups: 1 / fast kinetics for Cr(II), V(II) and U(III); 2 / slow kinetics for Ti(III); 3 / very slow kinetics for V(III) and U(VI). In order to contribute to the design of an electrochemical reactor for the generation of the reducing agents usable for the lixiviation of plutonium bearing solid wastes, the study of the diffusion coefficients of both oxidized and reduced forms of different redox couples, at different temperatures, was undertaken. The results of this study also permits, from the knowledge of the diffusional activation energy of the ions, to conclude that the dissolution of pure plutonium dioxide under the action of these reducing agents is not diffusion limited. The feasibility of the plutonium decontamination treatment of synthetic or real solid wastes was then studied at laboratory scale using electro-generated V(II), which is with Cr(II) among the best reagents. The efficiency of the treatment was good, (80 pc Pu solubilisation yield), especially in the case of cellulosic or miscellaneous organic wastes. (author) [fr

  4. Determination of plutonium and uranium in the same aliquot by potentiometric titration

    International Nuclear Information System (INIS)

    Karekar, C.V.; Chander, Keshav; Nair, G.M.; Natarajan, P.R.

    1986-01-01

    A potentiometric titration method was developed for the determination of plutonium and uranium in the same aliquot in nitric acid medium. Plutonium was first determined by oxidation to Pu(VI) by fuming with HClO 4 . Pu(VI) was reduced to Pu(IV) with known excess of Fe(II). Uranium in the same solution was determined by reduction to U(IV) with Fe(II) in H 3 PO 4 medium. For the quantity of plutonium and uranium in the range of 3-5 mg per aliquot a precision of +-0.2% and +-0.4%, respectively, was obtained. (author)

  5. Plutonium

    International Nuclear Information System (INIS)

    Koelzer, W.

    1989-03-01

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

  6. International management and storage of plutonium and spent fuel

    International Nuclear Information System (INIS)

    1978-09-01

    The first part of this study discusses certain questions that may arise from the disseminated production and storage of plutonium and, in the light of the relevant provisions of the Agency's Statute, examines possible arrangements for the storage of separated plutonium under international auspices and its release to meet energy or research requirements. The second part of the study deals similarly with certain problems presented by growing accumulations of spent fuel from light-water reactors in various countries and examines possible solutions, including the establishment of regional or multinational spent fuel storage facilities

  7. Plutonium

    International Nuclear Information System (INIS)

    Mueller-Christiansen, K.; Wollesen, M.

    1979-01-01

    As emotions and fear of plutonium are neither useful for the non-professionals nor for the political decision makers and the advantages and disadvantages of plutonium can only put against each other under difficulties, the paper wants to present the most essential scientific data of plutonium in a generally understandable way. Each of the individual sections is concluded and they try to give an answer to the most discussed questions. In order to make understanding easier, the scientific facts are only brought at points where it cannot be done without for the correctness of the presentation. Many details were left out knowingly. On the other hand, important details are dealt with several times if it seems necessary for making the presentation correct. The graphical presentations and the figures in many cases contain more than said in the text. They give the interested reader hints to scientific-technical coherences. The total material is to enable the reader to form his own opinion on plutonium problems which are being discussed in public. (orig./HP) [de

  8. Sorption/Desorption Interactions of Plutonium with Montmorillonite

    Science.gov (United States)

    Begg, J.; Zavarin, M.; Zhao, P.; Kersting, A. B.

    2012-12-01

    Plutonium (Pu) release to the environment through nuclear weapon development and the nuclear fuel cycle is an unfortunate legacy of the nuclear age. In part due to public health concerns over the risk of Pu contamination of drinking water, predicting the behavior of Pu in both surface and sub-surface water is a topic of continued interest. Typically it was assumed that Pu mobility in groundwater would be severely restricted, as laboratory adsorption studies commonly show that naturally occurring minerals can effectively remove plutonium from solution. However, evidence for the transport of Pu over significant distances at field sites highlights a relative lack of understanding of the fundamental processes controlling plutonium behavior in natural systems. At several field locations, enhanced mobility is due to Pu association with colloidal particles that serve to increase the transport of sorbed contaminants (Kersting et al., 1999; Santschi et al., 2002, Novikov et al., 2006). The ability for mineral colloids to transport Pu is in part controlled by its oxidation state and the rate of plutonium adsorption to, and desorption from, the mineral surface. Previously we have investigated the adsorption affinity of Pu for montmorillonite colloids, finding affinities to be similar over a wide range of Pu concentrations. In the present study we examine the stability of adsorbed Pu on the mineral surface. Pu(IV) at an initial concentration of 10-10 M was pre-equilibrated with montmorillonite in a background electrolyte at pH values of 4, 6 and 8. Following equilibration, aliquots of the suspensions were placed in a flow cell and Pu-free background electrolyte at the relevant pH was passed through the system. Flow rates were varied in order to investigate the kinetics of desorption and hence gain a mechanistic understanding of the desorption process. The flow cell experiments demonstrate that desorption of Pu from the montmorillonite surface cannot be modeled as a simple

  9. Experience with civil plutonium management: Technology and economics

    International Nuclear Information System (INIS)

    Zarimpas, N.; Stevens, G.H.

    1997-01-01

    Recent NEA work on plutonium has been, essentially, targeted at economic and scientific aspects and the need to identify suitable technical solutions, despite the existing political uncertainties associated with their implementation. Such studies provide the facts and current views concerning plutonium and its civil use; address questions influencing the choice of fuel cycle options and illustrate how economic and logistic assessments of the alternatives could be undertaken. An ad-hoc expert group, with a membership drawn from fifteen countries and three international organisations, which was formed in early 1994 under the auspices of the NEA, with the task of identifying, examining and evaluating the broad technical questions related to plutonium management, has just published its work. This paper discusses the work and main conclusions of the expert group and focuses on the following two topics: Technologies, already implemented, which provide for short and medium-term storage of plutonium or for recycling the plutonium through reactors. A brief review is provided of experience gained with them and technical commentaries are made on their potential future deployment; Such technologies may, in the longer term, be joined by a further range which are, in some cases, already under development. Attention is drawn to those additional options that may become available. Another NEA expert group studied in detail the economics of the open and closed fuel cycles and reported, in 1994, its main findings: Some economic considerations of importance to various aspects of plutonium recycling are also presented in this paper. (author)

  10. Status of plutonium ceramic immobilization processes and immobilization forms

    Energy Technology Data Exchange (ETDEWEB)

    Ebbinghaus, B.B.; Van Konynenburg, R.A. [Lawrence Livermore National Lab., CA (United States); Vance, E.R.; Jostsons, A. [Australian Nuclear Science and Technology Organization, Menai (Australia)] [and others

    1996-05-01

    Immobilization in a ceramic followed by permanent emplacement in a repository or borehole is one of the alternatives currently being considered by the Fissile Materials Disposition Program for the ultimate disposal of excess weapons-grade plutonium. To make Pu recovery more difficult, radioactive cesium may also be incorporated into the immobilization form. Valuable data are already available for ceramics form R&D efforts to immobilize high-level and mixed wastes. Ceramics have a high capacity for actinides, cesium, and some neutron absorbers. A unique characteristic of ceramics is the existence of mineral analogues found in nature that have demonstrated actinide immobilization over geologic time periods. The ceramic form currently being considered for plutonium disposition is a synthetic rock (SYNROC) material composed primarily of zirconolite (CaZrTi{sub 2}O{sub 7}), the desired actinide host phase, with lesser amounts of hollandite (BaAl{sub 2}Ti{sub 6}O{sub 16}) and rutile (TiO{sub 2}). Alternative actinide host phases are also being considered. These include pyrochlore (Gd{sub 2}Ti{sub 2}O{sub 7}), zircon (ZrSiO{sub 4}), and monazite (CePO{sub 4}), to name a few of the most promising. R&D activities to address important technical issues are discussed. Primarily these include moderate scale hot press fabrications with plutonium, direct loading of PuO{sub 2} powder, cold press and sinter fabrication methods, and immobilization form formulation issues.

  11. Decontamination of Battelle-Columbus' Plutonium Facility. Final report

    International Nuclear Information System (INIS)

    Rudolph, A.; Kirsch, G.; Toy, H.L.

    1984-01-01

    The Plutonium Laboratory, owned and operated by Battelle Memorial Institute's Columbus Division, was located in Battelle's Nuclear Sciences area near West Jefferson, Ohio, approximately 17 miles west of Columbus, Ohio. Originally built in 1960 for plutonium research and processing, the Plutonium Laboratory was enlarged in 1964 and again in 1967. With the termination of the Advanced Fuel Program in March, 1977, the decision was made to decommission the Plutonium Laboratory and to decontaminate the building for unrestricted use. Decontamination procedures began in January, 1978. All items which had come into contact with radioactivity from the plutonium operations were cleaned or disposed of through prescribed channels, maintaining procedures to ensure that D and D operations would pose no risk to the public, the environment, or the workers. The entire program was conducted under the cognizance of DOE's Chicago Operations Office. The building which housed the Plutonium Laboratory has now been decontaminated to levels allowing it to house ordinary laboratory and office operations. A ''Finding of No Significant Impact'' (FNSI) was issued in May, 1980

  12. Exploiting the plutonium stockpiles in PWRs by using inert matrix fuel

    International Nuclear Information System (INIS)

    Lombardi, C.; Mazzola, A.

    1996-01-01

    The plutonium coming from dismantled warheads and that already stockpiled coming from spent fuel reprocessing have raised many concerns related to proliferation resistance, environmental safety and economy. The option of disposing of plutonium by fission is one of the most widely discussed and many proposals for plutonium burning in a safe and economical manner have been put forward. Due to their diffusion, PWRs appear to be the main candidates for the reduction of the plutonium stockpiles. In order to achieve a high plutonium consumption rate, a uranium-free fuel may be conceived, based on the dilution of PuO 2 within a carrier matrix made of inert oxide. In this paper, a partial loading of inert matrix fuel in a current technology PWR was investigated with 3-D calculations. The results indicated that this solution has good plutonium elimination capabilities: commercial PWRs operating in a once-through cycle scheme can transmute more than 98% of the loaded Pu-239 and 73 or 81% of the overall initially loaded reactor grade or weapons grade plutonium, respectively. The plutonium still let in the spent fuel was of poor quality and then offered a better proliferation resistance. Power peaking problems could be faced with the adoption of burnable absorbers: IFBA seemed to be particularly suitable. In spite of a reduction of the overall plutonium loaded mass by a factor 3.7 or 5.4 depending on its quality, there was no evidence of an increase of the minor actinides radiotoxicity after a time period of about 25 years. (author)

  13. Method of purification of solutions containing plutonium and/or neptunium by plutonium and/or neptunium separation

    Energy Technology Data Exchange (ETDEWEB)

    Herz, D; Kankura, R; Wenzel, U

    1975-03-27

    Aqueous solutions containing, amongst other radiation sources, 10/sup -2/ to 10/sup -7/ mol per litre Pu and/or Np - especially aqueous solutions produced when reprocessing fuel elements, in particular of HTR reactors, e.g. according to the Thorex process - can be cleaned from the two metals without much expenditure by using a separating column: determination of the H/sup +/ ion concentration (0.5-6 mol/litre solution), addition of a stabilizing agent (NaNO/sub 2/ or Fe (NH/sub 2/SO/sub 3/)/sub 2/) and passage across a separating column in which the granulate charge is wetted with a long-chain alkylamine (e.g. trioctylamine) which retains the metal. The production of a charge (of Voltalef UF 300) is described. The process may also be carried out continuously.

  14. Plutonium controversy

    International Nuclear Information System (INIS)

    Richmond, C.R.

    1980-01-01

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

  15. Perspective on plutonium

    International Nuclear Information System (INIS)

    Sun, L.S.

    1993-01-01

    This paper is intended as a brief overview on the element plutonium. Plutonium is the first primarily man-made element to play a significant role not only in technological development, but also in the economic growth of many countries. The importance of plutonium centers around its enormous energy making it ideal for wide-scale use in reactors, while the nuclear industry continues to work toward improving safety and efficiency of plutonium as a reactor fuel politicians and the public still debate over the safety and benefits of nuclear power. (30 refs.)

  16. Plutonium controversy

    Energy Technology Data Exchange (ETDEWEB)

    Richmond, C.R.

    1980-01-01

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

  17. Hydrometallurgical treatment of plutonium. Bearing salt baths waste

    International Nuclear Information System (INIS)

    Bros, P.; Gozlan, J.P.; Lecomte, M.; Bourges, J.

    1993-01-01

    The salt flux issuing from the electrorefining of plutonium metal alloy in salt baths (KCI + NaCI) poses a difficult problem of the back-end alpha waste management. An alternative to the salt process promoted by Los Alamos Laboratory is to develop a hydrometallurgical treatment. A new process based on the electrochemistry technique in aqueous solution has been defined and tested successfully in the CEA. The diagram of the process exhibits two principal steps: in the head-end, a dissolution in HNO 3 medium accompanied with an electrolytic dechlorination leading to a quantitative elimination of chloride as CI 2 gas followed by its trapping one soda lime cartridge, a complete oxidative dissolution of the refractory Pu residues by electrogenerated Ag(II), in the back-end: the Pu and Am recoveries by chromatographic extractions. (authors). 10 figs., 9 refs

  18. 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.

  19. Plutonium, nuclear fuel; Le plutonium, combustible nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Grison, E [Commissariat a l' Energie Atomique, Fontenay aux Roses (France). Centre d' Etudes Nucleaires, Saclay

    1960-07-01

    A review of the physical properties of metallic plutonium, its preparation, and the alloys which it forms with the main nuclear metals. Appreciation of its future as a nuclear fuel. (author) [French] Apercu sur les proprietes physiques du plutonium metallique, sa preparation, ses alliages avec les principaux metaux nucleaires. Consideration sur son avenir en tant que combustible nucleaire. (auteur)

  20. Some studies on the extraction of plutonium from phosphate containing nitric acid solutions using DBDECMP as extractant (Preprint No. CT-24)

    International Nuclear Information System (INIS)

    Sagar, V.B.; Pawar, S.M.; Oak, M.S.; Sivaramakrishnan, C.K.

    1988-02-01

    Extraction studies have been carried out to explore the feasibility of separation of Pu(IV) from phosphate containing analytical wastes generated in the laboratory. Distribution data on the extraction of Pu(IV) from dibutyl-N,N diethylcarbamoylmethylenephosphonate (DBDECMP) in xylene from an aqu eous nitric acid and its mixture with sulphuric as well as with sulphuric and phosphoric acids were obtained. Based on the data obtained the conditions for the recovery of plutonium from such waste solutions are suggested. (author)