Science.gov

Sample records for plutonium purification cycle

  1. Plutonium purification cycle in centrifugal extractors: from flowsheet design to industrial operation

    SciTech Connect

    Baron, P.; Dinh, B.; Duhamet, J.; Drain, F.; Meze, F.; Lavenu, A.

    2008-07-01

    The extension of the UP2 plant at La Hague includes a new plutonium purification cycle using multistage centrifugal extractors to replace the previous cycle that used mixer/settler banks. This type of extractor is suitable for the treatment of fuel containing a high proportion of plutonium-238, as its short residence time limits solvent degradation. This paper deals with the research done to devise its flowsheet, the centrifugal extractors in which it is operated, as well as the feedback of six years of industrial operation.

  2. PLUTONIUM PURIFICATION PROCESS EMPLOYING THORIUM PYROPHOSPHATE CARRIER

    DOEpatents

    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.

  3. Concentration and purification of plutonium or thorium

    DOEpatents

    Hayden, John A.; Plock, Carl E.

    1976-01-01

    In this invention a first solution obtained from such as a plutonium/thorium purification process or the like, containing plutonium (Pu) and/or thorium (Th) in such as a low nitric acid (HNO.sub.3) concentration may have the Pu and/or Th separated and concentrated by passing an electrical current from a first solution having disposed therein an anode to a second solution having disposed therein a cathode and separated from the first solution by a cation permeable membrane, the Pu or Th cation permeating the cation membrane and forming an anionic complex within the second solution, and electrical current passage affecting the complex formed to permeate an anion membrane separating the second solution from an adjoining third solution containing disposed therein an anode, thereby effecting separation and concentration of the Pu and/or Th in the third solution.

  4. Purification of aqueous plutonium chloride solutions via precipitation and washing.

    SciTech Connect

    Stroud, M. A.; Salazar, R. R.; Abney, Kent David; Bluhm, E. A.; Danis, J. A.

    2003-01-01

    Pyrochemical operations at Los Alamos Plutonium Facility (TA-55) use high temperature melt s of calcium chloride for the reduction of plutonium oxide to plutonium metal and hi gh temperature combined melts of sodium chloride and potassium chloride mixtures for the electrorefining purification of plutonium metal . The remaining plutonium and americium are recovered from thes e salts by dissolution in concentrated hydrochloric acid followed by either solvent extraction or io n exchange for isolation and ultimately converted to oxide after precipitation with oxalic acid . Figur e 1 illustrates the current aqueous chloride flow sheet used for plutonium processing at TA-55 .

  5. Plutonium transmutation in thorium fuel cycle

    SciTech Connect

    Necas, Vladimir; Breza, Juraj |; Darilek, Petr

    2007-07-01

    The HELIOS spectral code was used to study the application of the thorium fuel cycle with plutonium as a supporting fissile material in a once-through scenario of the light water reactors PWR and VVER-440 (Russian design). Our analysis was focused on the plutonium transmutation potential and the plutonium radiotoxicity course of hypothetical thorium-based cycles for current nuclear power reactors. The paper shows a possibility to transmute about 50% of plutonium in analysed reactors. Positive influence on radiotoxicity after 300 years and later was pointed out. (authors)

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

    SciTech Connect

    Gardner, Kyle Shelton; Kimball, David Bryan; Skidmore, Bradley Evan

    2016-09-28

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

  7. Tests of alternative reductants in the second uranium purification cycle

    SciTech Connect

    Thompson, M.C.

    1980-05-01

    Miniature mixer-settler tests of the second uranium purification cycle show that plutonium cannot be removed by hydroxylamine-hydrazine (NH/sub 2/OH-N/sub 2/H/sub 4/) because the acidity is too high, or by 2,5-di-t-pentylhydroquinone because HNO/sub 3/ oxidizes the hydroquinone. Plutonium can be removed satisfactorily when U(IV)-hydrazine is used as the reductant.

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

    SciTech Connect

    Kimball, David Bryan; Skidmore, Bradley Evan

    2016-06-22

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

  9. Thermal Cycling on Fatigue Failure of the Plutonium Vitrification Melter

    SciTech Connect

    Jordan, Jeffrey; Gorczyca, Jennifer

    2009-02-11

    One method for disposition of excess plutonium is vitrification into cylindrical wasteforms. Due to the hazards of working with plutonium, the vitrification process must be carried out remotely in a shielded environment. Thus, the equipment must be easily maintained. With their simple design, induction melters satisfy this criterion, making them ideal candidates for plutonium vitrification. However, due to repeated heating and cooling cycles and differences in coefficients of thermal expansion of contacting materials fatigue failure of the induction melter is of concern. Due to the cost of the melter, the number of cycles to failure is critical. This paper presents a method for determining the cycles to failure for an induction melter by using the results from thermal and structural analyses as input to a fatigue failure model.

  10. METHOD AND MEANS FOR ELECTROLYTIC PURIFICATION OF PLUTONIUM

    DOEpatents

    Bjorklund, C.W.; Benz, R.; Maraman, W.J.; Leary, J.A.; Walsh, K.A.

    1960-02-01

    The technique of electrodepositing pure plutonium from a fused salt electrolyte of PuCl/sub 3/ and aixati metal halides is described. When an iron cathode is used, the plutonium deposit alloys therewith in the liquid state at the 400 to 600 deg C operating temperature, such liquid being allowed to drip through holes in the cathode and collect in a massive state in a tantallum cup. The process is adaptable to continuous processing by the use of depleted plutonium fuel as the anode: good to excellent separation from fission products is obtained with a Pu--Fe "fission" anode containing representative fractions of Ce, Ru, Zr, La, Mo, and Nb.

  11. Improved recovery and purification of plutonium at Los Alamos using macroporous anion exchange resin

    SciTech Connect

    Marsh, S.F.; Mann, M.J.

    1987-05-01

    For almost 30 years, Los Alamos National Laboratory has used anion exchange in nitric acid as the major aqueous process or the recovery and purification of plutonium. One of the few disadvantages of this system is the particularly slow rate at which the anionic nitrato complex of Pu(IV) equilibrates with the resin. The Nuclear Materials Process Technology Group at Los Alamos recently completed an ion exchange development program that focused on improving the slow sorption kinetics that limits this process. A comprehensive investigation of modern anion exchange resins identified porosity and bead size as the properties that most influence plutonium sorption kinetics. Our study found that small beads of macroporous resin produced a dramatic increase in plutonium process efficiency. The Rocky Flats Plant has already adopted this improved ion exchange technology, and it currently is being evaluated for use in other DOE plutonium-processing facilities.

  12. In situ purification, alloying and casting methodology for metallic plutonium

    NASA Astrophysics Data System (ADS)

    Lashley, Jason C.; Blau, Michael S.; Staudhammer, Karl P.; Pereyra, Ramiro A.

    Plutonium metal that has been double ER (electrorefined/electrorefining) was further purified via zone refining, using a floating molten zone to minimize the introduction of impurities. The temperature of the molten zone was 750°C, and the atmosphere was 10 -5 Pa. A total of ten zone refining passes were made at a travel rate of 1.5 cm/h. There were 19 elements reduced to quantities below the minimum detectable limits (MDL) by zone refining, while P, K, and W were significantly reduced. The zone-refined metal was then used in an in situ distillation, alloying, and casting step to prepare tapered specimens for single-crystal growth experiments. Specifically, 241Am was distilled from Pu metal by levitating Pu metal with 1 wt% Ga in the melt in a Crystallox vertical electromagnetic levitation crucible at 10 -5 Pa. The Pu is alloyed with Ga to stabilize the δ phase (fcc symmetry) upon solidification. The Pu was chill-cast directly from the electromagnetic levitation field into 1- cm tapered specimens. A water-cooled ceramic mold was used, and the Pu metal was cooled at a rate of 100°C/min. A microstructure examination of the specimen showed 10 × 25 μm acicular grains with a density of 15.938 g/cm 3 (±0.002 g/cm 3).

  13. Plutonium

    NASA Astrophysics Data System (ADS)

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

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

  14. Materials measurement and accounting in an operating plutonium conversion and purification process. Phase I. Process modeling and simulation. [PUCSF code

    SciTech Connect

    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.

  15. Implications of Plutonium isotopic separation on closed fuel cycles and repository design

    SciTech Connect

    Forsberg, C.

    2013-07-01

    Advances in laser enrichment may enable relatively low-cost plutonium isotopic separation. This would have large impacts on LWR closed fuel cycles and waste management. If Pu-240 is removed before recycling plutonium as mixed oxide (MOX) fuel, it would dramatically reduce the buildup of higher plutonium isotopes, Americium, and Curium. Pu-240 is a fertile material and thus can be replaced by U-238. Eliminating the higher plutonium isotopes in MOX fuel increases the Doppler feedback, simplifies reactor control, and allows infinite recycle of MOX plutonium in LWRs. Eliminating fertile Pu-240 and Pu-242 reduces the plutonium content in MOX fuel and simplifies fabrication. Reducing production of Pu-241 reduces production of Am-241 - the primary heat generator in spent nuclear fuels after several decades. Reducing heat generating Am-241 would reduce repository cost and waste toxicity. Avoiding Am- 241 avoids its decay product Np-237, a nuclide that partly controls long-term oxidizing repository performance. Most of these benefits also apply to LWR plutonium recycled into fast reactors. There are benefits for plutonium isotopic separation in fast reactor fuel cycles (particularly removal of Pu-242) but the benefits are less. (author)

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

    NASA Astrophysics Data System (ADS)

    Ewing, R. C.

    2009-12-01

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

  17. LAB-SCALE DEMONSTRATION OF PLUTONIUM PURIFICATION BY ANION EXCHANGE, PLUTONIUM (IV) OXALATE PRECIPITATION, AND CALCINATION TO PLUTONIUM OXIDE TO SUPPORT THE MOX FEED MISSION

    SciTech Connect

    Crowder, M.; Pierce, R.

    2012-08-22

    H-Canyon and HB-Line are tasked with the production of PuO{sub 2} from a feed of plutonium metal. The PuO{sub 2} will provide feed material for the MOX Fuel Fabrication Facility. After dissolution of the Pu metal in H-Canyon, the solution will be transferred to HB-Line for purification by anion exchange. Subsequent unit operations include Pu(IV) oxalate precipitation, filtration and calcination to form PuO{sub 2}. This report details the results from SRNL anion exchange, precipitation, filtration, calcination, and characterization tests, as requested by HB-Line1 and described in the task plan. This study involved an 80-g batch of Pu and employed test conditions prototypical of HB-Line conditions, wherever feasible. In addition, this study integrated lessons learned from earlier anion exchange and precipitation and calcination studies. H-Area Engineering selected direct strike Pu(IV) oxalate precipitation to produce a more dense PuO{sub 2} product than expected from Pu(III) oxalate precipitation. One benefit of the Pu(IV) approach is that it eliminates the need for reduction by ascorbic acid. The proposed HB-Line precipitation process involves a digestion time of 5 minutes after the time (44 min) required for oxalic acid addition. These were the conditions during HB-line production of neptunium oxide (NpO{sub 2}). In addition, a series of small Pu(IV) oxalate precipitation tests with different digestion times were conducted to better understand the effect of digestion time on particle size, filtration efficiency and other factors. To test the recommended process conditions, researchers performed two nearly-identical larger-scale precipitation and calcination tests. The calcined batches of PuO{sub 2} were characterized for density, specific surface area (SSA), particle size, moisture content, and impurities. Because the 3013 Standard requires that the calcination (or stabilization) process eliminate organics, characterization of PuO{sub 2} batches monitored the

  18. Plutonium in Concentrated Solutions

    SciTech Connect

    Clark, Sue B.; Delegard, Calvin H.

    2002-08-01

    Complex, high ionic strength media are used throughout the plutonium cycle, from its processing and purification in nitric acid, to waste storage and processing in alkaline solutions of concentrated electrolytes, to geologic disposal in brines. Plutonium oxidation/reduction, stability, radiolysis, solution and solid phase chemistry have been studied in such systems. In some cases, predictive models for describing Pu chemistry under such non-ideal conditions have been developed, which are usually based on empirical databases describing specific ion interactions. In Chapter 11, Non-Ideal Systems, studies on the behavior of Pu in various complex media and available model descriptions are reviewed.

  19. Effects on the long term storage container by thermal cycling alpha plutonium

    SciTech Connect

    Flamm, B.F.; Prenger, F.C.; Veirs, D.K.; Hill, D.D.; Isom, G.M.

    1998-03-01

    Experiments were conducted to determine the validity of the steady state temperature limit of 100 C established by the DOE-STD-3013-96 for storing alpha plutonium metal. Studies with an alpha plutonium ingot combined with strain gauge measurements indicate that the stainless steel storage container, yields very little (0.005 in.) to the expanding plutonium metal as it undergoes alpha beta phase transformation at temperatures above 112 C. Another experiment using an alpha plutonium rod for point loading of the container wall showed no measured deformation of the container. The results of strain measurements for alpha beta and beta alpha transformations for twenty five thermal cycles are reported. Finite element modeling using the measured data predicts that the compressive yield strength is 3,500 psi versus the literature value of 13,000 psi.

  20. Rapid separation and purification of uranium and plutonium from dilute-matrix samples

    SciTech Connect

    Armstrong, Christopher R.; Ticknor, Brian W.; Hall, Gregory; Cadieux, James R.

    2014-03-11

    This work presents a streamlined separation and purification approach for trace uranium and plutonium from dilute (carrier-free) matrices. The method, effective for nanogram quantities of U and femtogram to picogram quantities of Pu, is ideally suited for environmental swipe samples that contain a small amount of collected bulk material. As such, it may be applicable for processing swipe samples such as those collected in IAEA inspection activities as well as swipes that are loaded with unknown analytes, such as those implemented in interlaboratory round-robin or proficiency tests. Additionally, the simplified actinide separation could find use in internal laboratory monitoring of clean room conditions prior to or following more extensive chemical processing. We describe key modifications to conventional techniques that result in a relatively rapid, cost-effective, and efficient U and Pu separation process. We demonstrate the efficacy of implementing anion exchange chromatography in a single column approach. We also show that hydrobromic acid is an effective substitute in lieu of hydroiodoic acid for eluting Pu. Lastly, we show that nitric acid is an effective digestion agent in lieu of perchloric acid and/or hydrofluoric acid. A step by step procedure of this process is detailed.

  1. Rapid separation and purification of uranium and plutonium from dilute-matrix samples

    DOE PAGES

    Armstrong, Christopher R.; Ticknor, Brian W.; Hall, Gregory; ...

    2014-03-11

    This work presents a streamlined separation and purification approach for trace uranium and plutonium from dilute (carrier-free) matrices. The method, effective for nanogram quantities of U and femtogram to picogram quantities of Pu, is ideally suited for environmental swipe samples that contain a small amount of collected bulk material. As such, it may be applicable for processing swipe samples such as those collected in IAEA inspection activities as well as swipes that are loaded with unknown analytes, such as those implemented in interlaboratory round-robin or proficiency tests. Additionally, the simplified actinide separation could find use in internal laboratory monitoring ofmore » clean room conditions prior to or following more extensive chemical processing. We describe key modifications to conventional techniques that result in a relatively rapid, cost-effective, and efficient U and Pu separation process. We demonstrate the efficacy of implementing anion exchange chromatography in a single column approach. We also show that hydrobromic acid is an effective substitute in lieu of hydroiodoic acid for eluting Pu. Lastly, we show that nitric acid is an effective digestion agent in lieu of perchloric acid and/or hydrofluoric acid. A step by step procedure of this process is detailed.« less

  2. The benefits of an advanced fast reactor fuel cycle for plutonium management

    SciTech Connect

    Hannum, W.H.; McFarlane, H.F.; Wade, D.C.; Hill, R.N.

    1996-12-31

    The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium and long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a `focus area` for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed.

  3. Expected behavior of plutonium in the IFR fuel cycle

    NASA Astrophysics Data System (ADS)

    Steunenberg, R. K.; Johnson, I.

    The Integral Fast Reactor (IFR) is a metal-fueled, sodium-cooled reactor that will consist initially of a U-Zr alloy core in which the enriched uranium will be replaced gradually by plutonium bred in a uranium blanket. The plutonium is concentrated to the required level by extraction from the molten blanket material with a CaCl2-BaCl2 salt containing MgCl2 as an oxidant (halide slagging). The CaCl2-BaCl2 salt containing dissolved PuCl3 and UCl3 is added to the core process where fission products are removed by electrorefining, using a liquid cadmium anode, a metal cathode, and a LiCl-NaCl-CaCl2-BaCl2 molten salt electrolyte. The product is recovered as a metallic deposit on the cathode. The Halide slagging step is operated at about 1250 deg and the electrorefining step at about 450 C. These processes are expected to give low fission-product decontamination factors of the order of 100.

  4. PURIFICATION OF PLUTONIUM USING A CERIUM PRECIPITATE AS A CARRIER FOR FISSION PRODUCTS

    DOEpatents

    Faris, B.F.; Olson, C.M.

    1961-07-01

    Bismuth phosphate carrier precipitation processes are described for the separation of plutonium from fission products wherein in at least one step bismuth phosphate is precipitated in the presence of hexavalent plutonium thereby carrying a portion of the fission products from soluble plu tonium values. In this step, a cerium phosphate precipitate is formed in conjunction with the bismuth phosphate precipitate, thereby increasing the amount of fission products removed from solution.

  5. A comparative assessment of the economics of plutonium disposition including comparison with other nuclear fuel cycles

    SciTech Connect

    Williams, K.A.; Miller, J.W.; Reid, R.L.

    1997-05-01

    DOE has been evaluating three technologies for the disposition of approximately 50 metric tons of surplus plutonium from defense-related programs: reactors, immobilization, and deep boreholes. As part of the process supporting an early CY 1997 Record of Decision (ROD), a comprehensive assessment of technical viability, cost, and schedule has been conducted. Oak Ridge National Laboratory has managed and coordinated the life-cycle cost (LCC) assessment effort for this program. This paper discusses the economic analysis methodology and the results prior to ROD. Other objectives of the paper are to discuss major technical and economic issues that impact plutonium disposition cost and schedule. Also to compare the economics of a once-through weapons-derived MOX nuclear fuel cycle to other fuel cycles, such as those utilizing spent fuel reprocessing. To evaluate the economics of these technologies on an equitable basis, a set of cost estimating guidelines and a common cost-estimating format were utilized by all three technology teams. This paper also includes the major economic analysis assumptions and the comparative constant-dollar and discounted-dollar LCCs.

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

    SciTech Connect

    Persiani, P.J.

    1995-06-01

    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. Reference annual mass flows and inventories for a representative 1,400 Mwe Pressurized Water Reactor (PWR) fuel cycle have been investigated for three cases: the 100 percent uranium oxide UO{sub 2} fuel loading once through cycle, and the 33 percent mixed oxide MOX loading configuration for a first and second plutonium recycle. The analysis addresses fuel cycle developments; plutonium and uranium inventory and flow balances; nuclear fuel processing operations; UO{sub 2} once-through and MOX first and second recycles; and the economic incentives to draw-down the excess separated plutonium stores. The preliminary analysis explores several options in reducing the excess separated plutonium arisings and HEU, and the consequences of the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials on nonproliferation and safeguards policy assessments.

  7. HB-LINE ANION EXCHANGE PURIFICATION OF AFS-2 PLUTONIUM FOR MOX

    SciTech Connect

    Kyser, E. A.; King, W. D.

    2012-07-31

    Non-radioactive cerium (Ce) and radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the feasibility of using either gadolinium nitrate (Gd) or boric acid (B as H{sub 3}BO{sub 3}) as a neutron poison in the H-Canyon dissolution process. Expected typical concentrations of probable impurities were tested and the removal of these impurities by a decontamination wash was measured. Impurity concentrations are compared to two specifications - designated as Column A or Column B (most restrictive) - proposed for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). Use of Gd as a neutron poison requires a larger volume of wash for the proposed Column A specification. Since boron (B) has a higher proposed specification and is more easily removed by washing, it appears to be the better candidate for use in the H-Canyon dissolution process. Some difficulty was observed in achieving the Column A specification due to the limited effectiveness that the wash step has in removing the residual B after ~4 BV's wash. However a combination of the experimental 10 BV's wash results and a calculated DF from the oxalate precipitation process yields an overall DF sufficient to meet the Column A specification. For those impurities (other than B) not removed by 10 BV's of wash, the impurity is either not expected to be present in the feedstock or process, or recommendations have been provided for improvement in the analytical detection/method or validation of calculated results. In summary, boron is recommended as the appropriate neutron poison for H-Canyon dissolution and impurities are expected to meet the Column A specification limits for oxide production in HB-Line.

  8. HB-LINE ANION EXCHANGE PURIFICATION OF AFS-2 PLUTONIUM FOR MOX

    SciTech Connect

    Kyser, E.; King, W.

    2012-04-25

    Non-radioactive cerium (Ce) and radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the feasibility of using either gadolinium nitrate (Gd) or boric acid (B as H{sub 3}BO{sub 3}) as a neutron poison in the H-Canyon dissolution process. Expected typical concentrations of probable impurities were tested and the removal of these impurities by a decontamination wash was measured. Impurity concentrations are compared to two specifications - designated as Column A or Column B (most restrictive) - proposed for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). Use of Gd as a neutron poison requires a larger volume of wash for the proposed Column A specification. Since boron (B) has a higher proposed specification and is more easily removed by washing, it appears to be the better candidate for use in the H-Canyon dissolution process. Some difficulty was observed in achieving the Column A specification due to the limited effectiveness that the wash step has in removing the residual B after {approx}4 BV's wash. However a combination of the experimental 10 BV's wash results and a calculated DF from the oxalate precipitation process yields an overall DF sufficient to meet the Column A specification. For those impurities (other than B) not removed by 10 BV's of wash, the impurity is either not expected to be present in the feedstock or process, or recommendations have been provided for improvement in the analytical detection/method or validation of calculated results. In summary, boron is recommended as the appropriate neutron poison for H-Canyon dissolution and impurities are expected to meet the Column A specification limits for oxide production in HB-Line.

  9. A novel concept of QUADRISO particles Part III : applications to the plutonium-thorium fuel cycle.

    SciTech Connect

    Talamo, A.

    2009-03-01

    In the present study, a plutonium-thorium fuel cycle is investigated including the {sup 233}U production and utilization. A prismatic thermal High Temperature Gas Reactor (HTGR) and the novel concept of quadruple isotropic (QUADRISO) coated particles, designed at the Argonne National Laboratory, have been used for the study. In absorbing QUADRISO particles, a burnable poison layer surrounds the central fuel kernel to flatten the reactivity curve as a function of time. At the beginning of life, the fuel in the QUADRISO particles is hidden from neutrons, since they get absorbed in the burnable poison before they reach the fuel kernel. Only when the burnable poison depletes, neutrons start streaming into the fuel kernel inducing fission reactions and compensating the fuel depletion of ordinary TRISO particles. In fertile QUADRISO particles, the absorber layer is replaced by natural thorium with the purpose of flattening the excess of reactivity by the thorium resonances and producing {sup 233}U. The above configuration has been compared with a configuration where fissile (neptunium-plutonium oxide from Light Water Reactors irradiated fuel) and fertile (natural thorium oxide) fuels are homogeneously mixed in the kernel of ordinary TRISO particles. For the {sup 233}U utilization, the core has been equipped with europium oxide absorbing QUADRISO particles.

  10. On the equilibrium isotopic composition of the thorium-uranium-plutonium fuel cycle

    NASA Astrophysics Data System (ADS)

    Marshalkin, V. Ye.; Povyshev, V. M.

    2016-12-01

    The equilibrium isotopic compositions and the times to equilibrium in the process of thorium-uranium-plutonium oxide fuel recycling in VVER-type reactors using heavy water mixed with light water are estimated. It is demonstrated thEhfat such reactors have a capacity to operate with self-reproduction of active isotopes in the equilibrium mode.

  11. On the equilibrium isotopic composition of the thorium–uranium–plutonium fuel cycle

    SciTech Connect

    Marshalkin, V. Ye. Povyshev, V. M.

    2016-12-15

    The equilibrium isotopic compositions and the times to equilibrium in the process of thorium–uranium–plutonium oxide fuel recycling in VVER-type reactors using heavy water mixed with light water are estimated. It is demonstrated thEhfat such reactors have a capacity to operate with self-reproduction of active isotopes in the equilibrium mode.

  12. Nuclear fuel cycle risk assessment: survey and computer compilation of risk-related literature. [Once-through Cycle and Plutonium Recycle

    SciTech Connect

    Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

    1982-10-01

    The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searched and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included.

  13. PLUTONIUM COMPOUNDS AND PROCESS FOR THEIR PREPARATION

    DOEpatents

    Wolter, F.J.; Diehl, H.C. Jr.

    1958-01-01

    This patent relates to certain new compounds of plutonium, and to the utilization of these compounds to effect purification or separation of the plutonium. The compounds are organic chelate compounds consisting of tetravalent plutonium together with a di(salicylal) alkylenediimine. These chelates are soluble in various organic solvents, but not in water. Use is made of this property in extracting the plutonium by contacting an aqueous solution thereof with an organic solution of the diimine. The plutonium is chelated, extracted and effectively separated from any impurities accompaying it in the aqueous phase.

  14. Plutonium controversy

    SciTech Connect

    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)

  15. Recovery of americium-241 from aged plutonium metal

    SciTech Connect

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

    1980-12-01

    About 5 kg of ingrown /sup 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 /sup 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% /sup 241/AmO/sub 2/; residual impurities were primarily lead and nickel.

  16. Pyrochemical processing of plutonium. Technology review report

    SciTech Connect

    Coops, M.S.; Knighton, J.B.; Mullins, L.J.

    1982-09-08

    Non-aqueous processes are now in routine use for direct conversion of plutonium oxide to metal, molten salt extraction of americium, and purification of impure metals by electrorefining. These processes are carried out at elevated temperatures in either refractory metal crucibles or magnesium-oxide ceramics in batch-mode operation. Direct oxide reduction is performed in units up to 700 gram PuO/sub 2/ batch size with molten calcium metal as the reductant and calcium chloride as the reaction flux. Americium metal is removed from plutonium metal by salt extraction with molten magnesium chloride. Electrorefining is used to isolate impurities from molten plutonium by molten salt ion transport in a controlled potential oxidation-reduction cell. Such cells can purify five or more kilograms of impure metal per 5-day electrorefining cycle. The product metal obtained is typically > 99.9% pure, starting from impure feeds. Metal scrap and crucible skulls are recovered by hydriding of the metallic residues and recovered either as impure metal or oxide feeds.

  17. BIOGEOCHEMICAL CYCLING AND ENVIRONMENTAL STABILITY OF PLUTONIUM RELEVANT TO LONG-TERM STEWARDSHIP OF DOE SITES

    SciTech Connect

    Francis, A.J.; Gillow, J.B.; Dodge, C.J.

    2006-06-01

    Pu is generally considered to be relatively immobile in the terrestrial environment, with the exception of transport via airborne and erosion mechanisms. More recently the transport of colloidal forms of Pu is being studied as a mobilization pathway from subsurface contaminated soils and sediments. The overall objective of this research is to understand the biogeochemical cycling of Pu in environments of interest to long-term DOE stewardship issues. Microbial processes are central to the immobilization of Pu species, through the metabolism of organically complexed Pu species and Pu associated with extracellular carrier phases and the creation of environments favorable for retardation of Pu transport.

  18. BIOGEOCHEMICAL CYCLING AND ENVIRONMENTAL STABILITY OF PLUTONIUM RELEVANT TO LONG-TERM STEWARDSHIP OF DOE SITES.

    SciTech Connect

    FRANCIS, A.J.; GILLOW, J.P.; DODGE, C.J.

    2006-11-16

    Pu is generally considered to be relatively immobile in the terrestrial environment, with the exception of transport via airborne and erosion mechanisms. More recently the transport of colloidal forms of Pu is being studied as a mobilization pathway from subsurface contaminated soils and sediments. The overall objective of this research is to understand the biogeochemical cycling of Pu in environments of interest to long-term DOE stewardship issues. Microbial processes are central to the immobilization of Pu species, through the metabolism of organically complexed Pu species and Pu associated with extracellular carrier phases and the creation of environments favorable for retardation of Pu transport.

  19. Recent plutonium science and technology at ORNL

    SciTech Connect

    Bell, J.T.

    1985-01-01

    Plutonium research and development (R and D) at ORNL has generally followed development of the nuclear fuel cycle. Basic plutonium chemistry studies have diminished since the mid-1970s; however, significant efforts have been made recently to determine fundamental characteristics of the aqueous plutonium polymer and to develop thermodynamic representations of plutonium oxides. Some studies have also been made on plutonium phosphates related to waste isolation and on definition of the oxidation states of environmental plutonium. The remaining work has been supported by the Consolidated Fuel Reprocessing Program (CFRP) and includes: (1) establishment of boundary limits for polymer formation in Purex systems; (2) preparation of mixed uranium-plutonium oxide microspheres by internal gelation sol-gel techniques; (3) direct thermal denitration of aqueous systems; and (4) plutonium/uranium extraction from spent fast reactor fuels.

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

    SciTech Connect

    Fife, K.W.; West, M.H.

    1987-05-01

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

  1. CONVERSION OF PLUTONIUM TRIFLUORIDE TO PLUTONIUM TETRAFLUORIDE

    DOEpatents

    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.

  2. Filières nucléaires et gestion du plutonium et des actinides mineurs la recherche de la flexibilité du cycle

    NASA Astrophysics Data System (ADS)

    Thomas, Jean-Baptiste

    2002-10-01

    Transuranics management concerns all NPP types, because of the specifications for a sustainable development. Multiple recycling is mandatory. Neutronic abundance can be obtained in fast spectrum, or by adding external neutrons or (temporarily) with additional 235U. The LWRs can control the plutonium inventory and significantly reduce the amount of transuranics transferred to the geological repository, thanks to the use of innovative nuclear fuel in a limited part of the NPP fleet. HTR adapted to transuranics burning can help. In the future, in addition to the liquid metal FBR, a strategy based on a gas cooled technological line and advanced fuel opens a second path towards fast spectra. Strategies for defining the optimal mix of reactor types in the nuclear fleet at a given time and demonstrating the fuel cycle flexibility are under study. To cite this article: J.-B. Thomas, C. R. Physique 3 (2002) 783-796.

  3. Electrochemical investigation into the mechanism of plutonium reduction in electrorefining

    SciTech Connect

    McCurry, L.E.; Moy, G.M.M.

    1987-01-01

    Currently impure plutonium metal is purified at Los Alamos National Laboratory by a molten salt electrorefining process. Electrorefining is an effective method for producing high-purity plutonium metal (> 99.95%). In general this process involves the oxidation of impure plutonium metal from a molten plutonium anode or a solvent metal/plutonium anode, transport of plutonium ions through a molten salt electrolyte, and reduction of the plutonium ions at a tungsten cathode to pure plutonium metal. Purification of the plutonium metal from impurities is based on the difference in free energies of formation between the various metallic impurities associated with plutonium. To obtain a better understanding of the overall electrorefining process and its inefficiencies, an electrochemical investigation into the mechanism for plutonium reduction in a typical electrorefining environment was undertaken. Cyclic voltammetry was selected as the method for determining the electrode mechanism for plutonium reduction at tungsten electrodes. In addition to the standard electrorefining melt (equimolar NaCl-KCl), additional melts that were being investigated in our solvent anode work were also investigated. With insight gained from this investigation, it was hoped that a better selection of electrorefining operating parameters could be obtained.

  4. Radionuclide Basics: Plutonium

    EPA Pesticide Factsheets

    Plutonium (chemical symbol Pu) is a radioactive metal. Plutonium is considered a man-made element. Plutonium-239 is used to make nuclear weapons. Pu-239 and Pu-240 are byproducts of nuclear reactor operations and nuclear bomb explosions.

  5. Plutonium story

    SciTech Connect

    Seaborg, G T

    1981-09-01

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

  6. Plutonium Story

    DOE R&D Accomplishments Database

    Seaborg, G. T.

    1981-09-01

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

  7. PLUTONIUM ALLOYS

    DOEpatents

    Chynoweth, W.

    1959-06-16

    The preparation of low-melting-point plutonium alloys is described. In a MgO crucible Pu is placed on top of the lighter alloying metal (Fe, Co, or Ni) and the temperature raised to 1000 or 1200 deg C. Upon cooling, the alloy slug is broke out of the crucible. With 14 at. % Ni the m.p. is 465 deg C; with 9.5 at. % Fe the m.p. is 410 deg C; and with 12.0 at. % Co the m.p. is 405 deg C. (T.R.H.) l6262 l6263 ((((((((Abstract unscannable))))))))

  8. PRODUCTION OF PLUTONIUM METAL

    DOEpatents

    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.

  9. SEPARATION OF PLUTONIUM

    DOEpatents

    Maddock, A.G.; Smith, F.

    1959-08-25

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

  10. STRIPPING PROCESS FOR PLUTONIUM

    DOEpatents

    Kolodney, M.

    1959-10-01

    A method for removing silver, nickel, cadmium, zinc, and indium coatings from plutonium objects while simultaneously rendering the plutonium object passive is described. The coated plutonium object is immersed as the anode in an electrolyte in which the plutonium is passive and the coating metal is not passive, using as a cathode a metal which does not dissolve rapidly in the electrolyte. and passing an electrical current through the electrolyte until the coating metal is removed from the plutonium body.

  11. Plutonium Finishing Plant safety evaluation report

    SciTech Connect

    Not Available

    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.

  12. Plutonium waste incineration using pyrohydrolysis

    SciTech Connect

    Meyer, M.L.

    1991-12-31

    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.

  13. Plutonium waste incineration using pyrohydrolysis

    SciTech Connect

    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.

  14. METHOD OF SEPARATING PLUTONIUM

    DOEpatents

    Brown, H.S.; Hill, O.F.

    1958-02-01

    Plutonium hexafluoride is a satisfactory fluorinating agent and may be reacted with various materials capable of forming fluorides, such as copper, iron, zinc, etc., with consequent formation of the metal fluoride and reduction of the plutonium to the form of a lower fluoride. In accordance with the present invention, it has been found that the reactivity of plutonium hexafluoride with other fluoridizable materials is so great that the process may be used as a method of separating plutonium from mixures containing plutonium hexafluoride and other vaporized fluorides even though the plutonium is present in but minute quantities. This process may be carried out by treating a mixture of fluoride vapors comprising plutonium hexafluoride and fluoride of uranium to selectively reduce the plutonium hexafluoride and convert it to a less volatile fluoride, and then recovering said less volatile fluoride from the vapor by condensation.

  15. Plutonium Training Opportunities

    SciTech Connect

    Balatsky, Galya Ivanovna; Wolkov, Benjamin

    2015-03-26

    This report was created to examine the current state of plutonium training in the United States and to discover ways in which to ensure that the next generation of plutonium workers are fully qualified.

  16. PREPARATION OF PLUTONIUM TRIFLUORIDE

    DOEpatents

    Burger, L.L.; Roake, W.E.

    1961-07-11

    A process of producing plutonium trifluoride by reacting dry plutonium(IV) oxalate with chlorofluorinated methane or ethane at 400 to 450 deg C and cooling the product in the absence of oxygen is described.

  17. PROCESS FOR PURIFYING PLUTONIUM

    DOEpatents

    Mastick, D.F.; Wigner, E.P.

    1958-05-01

    A method is described of separating plutonium from small amounts of uranium and other contaminants. An acidic aqueous solution of higher valent plutonium and hexavalent uranium is treated with a soluble iodide to obtain the plutonium in the plus three oxidation state while leaving the uranium in the hexavalent state, adding a soluble oxalate such as oxalic acid, and then separating the insoluble plus the plutonium trioxalate from the solution.

  18. Dose estimates of alternative plutonium pyrochemical processes.

    SciTech Connect

    Kornreich, D. E.; Jackson, J. W.; Boerigter, S. T.; Averill, W. A.; Fasel, J. H.

    2002-01-01

    We have coupled our dose calculation tool Pandemonium with a discrete-event, object-oriented, process-modeling system ProMosO to analyze a set of alternatives for plutonium purification operations. The results follow expected trends and indicate, from a dose perspective, that an experimental flowsheet may warrant further research to see if it can be scaled to industrial levels. Flowsheets that include fluoride processes resulted in the largest doses.

  19. PLUTONIUM CLEANING PROCESS

    DOEpatents

    Kolodney, M.

    1959-12-01

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

  20. PLUTONIUM-THORIUM ALLOYS

    DOEpatents

    Schonfeld, F.W.

    1959-09-15

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

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

    SciTech Connect

    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.

  2. METHOD OF SEPARATING PLUTONIUM

    DOEpatents

    Heal, H.G.

    1960-02-16

    BS>A method of separating plutonium from aqueous nitrate solutions of plutonium, uranium. and high beta activity fission products is given. The pH of the aqueous solution is adjusted between 3.0 to 6.0 with ammonium acetate, ferric nitrate is added, and the solution is heated to 80 to 100 deg C to selectively form a basic ferric plutonium-carrying precipitate.

  3. PREPARATION OF PLUTONIUM HALIDES

    DOEpatents

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

    1958-11-01

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

  4. PLUTONIUM-ZIRCONIUM ALLOYS

    DOEpatents

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

    1960-08-30

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

  5. PREPARATION OF PLUTONIUM

    DOEpatents

    Kolodney, M.

    1959-07-01

    Methods are presented for the electro-deposition of plutonium from fused mixtures of plutonium halides and halides of the alkali metals and alkaline earth metals. Th salts, preferably chlorides and with the plutonium prefer ably in the trivalent state, are placed in a refractory crucible such as tantalum or molybdenam and heated in a non-oxidizing atmosphere to 600 to 850 deg C, the higher temperatatures being used to obtain massive plutonium and the lower for the powder form. Electrodes of graphite or non reactive refractory metals are used, the crucible serving the cathode in one apparatus described in the patent.

  6. Continuous plutonium dissolution apparatus

    DOEpatents

    Meyer, F.G.; Tesitor, C.N.

    1974-02-26

    This invention is concerned with continuous dissolution of metals such as plutonium. A high normality acid mixture is fed into a boiler vessel, vaporized, and subsequently condensed as a low normality acid mixture. The mixture is then conveyed to a dissolution vessel and contacted with the plutonium metal to dissolve the plutonium in the dissolution vessel, reacting therewith forming plutonium nitrate. The reaction products are then conveyed to the mixing vessel and maintained soluble by the high normality acid, with separation and removal of the desired constituent. (Official Gazette)

  7. North Korean plutonium production

    SciTech Connect

    Albright, D.

    1994-12-01

    In 1992, as part of its obligations under the Nuclear Non-Proliferation Treaty, North Korea declared that it had earlier separated about 100 grams of plutonium from damaged fuel rods removed from a 25 megawatt-thermal (MW{sub t}) gas-graphite reactor at Yongbyon. The plutonium was separated at the nearby {open_quotes}Radiochemical Laboratory.{close_quotes} Separated plutonium is the raw ingredient for making nuclear weapons, but 100 grams is too little to make a crude bomb. Based on intelligence reports and IAEA inspections, North Korea may have separated enough plutonium for a nuclear weapon. Regardless of whether this is true, there is no doubt that North Korea has enough weapons-grade plutonium in spent fuel to make four or five nuclear weapons. But it cannot turn this plutonium into nuclear weapons unless it separates the plutonium from the spent fuel. Preventing the North from separating any more plutonium must remain a global priority. The IAEA must also be able to verify North Korea`s past nuclear activities and determine the amount of plutonium North Korea may have diverted in the past.

  8. Anthropogenic plutonium-244 in the environment: Insights into plutonium's longest-lived isotope.

    PubMed

    Armstrong, Christopher R; Brant, Heather A; Nuessle, Patterson R; Hall, Gregory; Cadieux, James R

    2016-02-22

    Owing to the rich history of heavy element production in the unique high flux reactors that operated at the Savannah River Site, USA (SRS) decades ago, trace quantities of plutonium with highly unique isotopic characteristics still persist today in the SRS terrestrial environment. Development of an effective sampling, processing, and analysis strategy enables detailed monitoring of the SRS environment, revealing plutonium isotopic compositions, e.g., (244)Pu, that reflect the unique legacy of plutonium production at SRS. This work describes the first long-term investigation of anthropogenic (244)Pu occurrence in the environment. Environmental samples, consisting of collected foot borne debris, were taken at SRS over an eleven year period, from 2003 to 2014. Separation and purification of trace plutonium was carried out followed by three stage thermal ionization mass spectrometry (3STIMS) measurements for plutonium isotopic content and isotopic ratios. Significant (244)Pu was measured in all of the years sampled with the highest amount observed in 2003. The (244)Pu content, in femtograms (fg = 10(-15) g) per gram, ranged from 0.31 fg/g to 44 fg/g in years 2006 and 2003 respectively. In all years, the (244)Pu/(239)Pu atom ratios were significantly higher than global fallout, ranging from 0.003 to 0.698 in years 2014 and 2003 respectively.

  9. Methods for separation/purification utilizing rapidly cycled thermal swing sorption

    DOEpatents

    Tonkovich, Anna Lee Y.; Monzyk, Bruce F.; Wang, Yong; VanderWiel, David P.; Perry, Steven T.; Fitzgerald, Sean P.; Simmons, Wayne W.; McDaniel, Jeffrey S.; Weller, Jr., Albert E.

    2004-11-09

    The present invention provides apparatus and methods for separating fluid components. In preferred embodiments, the apparatus and methods utilize microchannel devices with small distances for heat and mass transfer to achieve rapid cycle times and surprisingly large volumes of fluid components separated in short times using relatively compact hardware.

  10. DELTA PHASE PLUTONIUM ALLOYS

    DOEpatents

    Cramer, E.M.; Ellinger, F.H.; Land. C.C.

    1960-03-22

    Delta-phase plutonium alloys were developed suitable for use as reactor fuels. The alloys consist of from 1 to 4 at.% zinc and the balance plutonium. The alloys have good neutronic, corrosion, and fabrication characteristics snd possess good dimensional characteristics throughout an operating temperature range from 300 to 490 deg C.

  11. PLUTONIUM-CERIUM ALLOY

    DOEpatents

    Coffinberry, A.S.

    1959-01-01

    An alloy is presented for use as a reactor fuel. The binary alloy consists essentially of from about 5 to 90 atomic per cent cerium and the balance being plutonium. A complete phase diagram for the cerium--plutonium system is given.

  12. ELECTRODEPOSITION OF PLUTONIUM

    DOEpatents

    Wolter, F.J.

    1957-09-10

    A process of electrolytically recovering plutonium from dilute aqueous solutions containing plutonium ions comprises electrolyzing the solution at a current density of about 0.44 ampere per square centimeter in the presence of an acetate-sulfate buffer while maintaining the pH of the solution at substantially 5 and using a stirred mercury cathode.

  13. 31. VIEW OF A WORKER HOLDING A PLUTONIUM 'BUTTON.' PLUTONIUM, ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    31. VIEW OF A WORKER HOLDING A PLUTONIUM 'BUTTON.' PLUTONIUM, A MAN-MADE SUBSTANCE, WAS RARE. SCRAPS RESULTING FROM PRODUCTION AND PLUTONIUM RECOVERED FROM RETIRED NUCLEAR WEAPONS WERE REPROCESSED INTO VALUABLE PURE-PLUTONIUM METAL (9/19/73). - Rocky Flats Plant, Bounded by Indiana Street & Routes 93, 128 & 72, Golden, Jefferson County, CO

  14. PLUTONIUM-CERIUM-COBALT AND PLUTONIUM-CERIUM-NICKEL ALLOYS

    DOEpatents

    Coffinberry, A.S.

    1959-08-25

    >New plutonium-base teroary alloys useful as liquid reactor fuels are described. The alloys consist of 10 to 20 atomic percent cobalt with the remainder plutonium and cerium in any desired proportion, with the plutonium not in excess of 88 atomic percent; or, of from 10 to 25 atomic percent nickel (or mixture of nickel and cobalt) with the remainder plutonium and cerium in any desired proportion, with the plutonium not in excess of 86 atomic percent. The stated advantages of these alloys over unalloyed plutonium for reactor fuel use are a lower melting point and a wide range of permissible plutonium dilution.

  15. Plutonium storage criteria

    SciTech Connect

    Chung, D.; Ascanio, X.

    1996-05-01

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

  16. Plutonium bioaccumulation in seabirds.

    PubMed

    Strumińska-Parulska, Dagmara I; Skwarzec, Bogdan; Fabisiak, Jacek

    2011-12-01

    The aim of the paper was plutonium (²³⁸Pu and ²³⁹⁺²⁴⁰Pu) determination in seabirds, permanently or temporarily living in northern Poland at the Baltic Sea coast. Together 11 marine birds species were examined: 3 species permanently residing in the southern Baltic, 4 species of wintering birds and 3 species of migrating birds. The obtained results indicated plutonium is non-uniformly distributed in organs and tissues of analyzed seabirds. The highest plutonium content was found in the digestion organs and feathers, the smallest in skin and muscles. The plutonium concentration was lower in analyzed species which feed on fish and much higher in herbivorous species. The main source of plutonium in analyzed marine birds was global atmospheric fallout.

  17. Plutonium Immobilization Canister Loading

    SciTech Connect

    Hamilton, E.L.

    1999-01-26

    This disposition of excess plutonium is determined by the Surplus Plutonium Disposition Environmental Impact Statement (SPD-EIS) being prepared by the Department of Energy. The disposition method (Known as ''can in canister'') combines cans of immobilized plutonium-ceramic disks (pucks) with vitrified high-level waste produced at the SRS Defense Waste Processing Facility (DWPF). This is intended to deter proliferation by making the plutonium unattractive for recovery or theft. The envisioned process remotely installs cans containing plutonium-ceramic pucks into storage magazines. Magazines are then remotely loaded into the DWPF canister through the canister neck with a robotic arm and locked into a storage rack inside the canister, which holds seven magazines. Finally, the canister is processed through DWPF and filled with high-level waste glass, thereby surrounding the product cans. This paper covers magazine and rack development and canister loading concepts.

  18. Method for dissolving plutonium dioxide

    DOEpatents

    Tallent, Othar K.

    1978-01-01

    The fluoride-catalyzed, non-oxidative dissolution of plutonium dioxide in HNO.sub.3 is significantly enhanced in rate by oxidizing dissolved plutonium ions. It is believed that the oxidation of dissolved plutonium releases fluoride ions from a soluble plutonium-fluoride complex for further catalytic action.

  19. The blending strategy for the plutonium immobilization program

    SciTech Connect

    Ebbinghaus, B B; Edmunds, T A; Gentry, S; Gray, L W; Riley, D C; Spingarn, J; VanKonynenburg, R A

    1999-02-12

    The Department of Energy (DOE) has declared approximately 38.2 tonnes of weapons-grade plutonium to be excess to the needs of national security, 14.3 tonnes of fuel- and reactor-grade plutonium excess to DOE needs, and anticipates an additional 7 tonnes to be declared excess to national security needs. Of this 59.5 tonnes, DOE anticipates that {approximately} 7.5 tonnes will be dispositioned as spent fuel at the Geologic Repository and {approximately} 2 tonnes will be declared below the safeguards termination limit and be discard3ed as TRU waste at WIPP. The remaining 50 tonnes of excess plutonium exists in many forms and locations around the country, and is under the control of several DOE offices. In addition to the plutonium, the feed stock also contains about 17 tonnes of depleted uranium, about 600 kg of highly enriched uranium, and many kilograms of neptunium and thorium and about 8 to 10 tonnes of tramp impurities. The Materials Disposition Program (MD) will be received materials packaged by these other Programs to disposition in a manor that meets the spent fuel standard. To minimize the cost of characterization of the feedstock and to minimize purification processes, a blending strategy will be followed. The levelization of the impurities, the plutonium isotopics, and the actinide impurities will also provide some benefits in the area of proliferation resistance. The overall strategy will be outlined and the benefits of following a blending instead of a purification program will be discussed.

  20. Design approaches for a cycling adsorbent/photocatalyst system for indoor air purification: formaldehyde example.

    PubMed

    Chin, Paul; Ollis, David F

    2008-04-01

    A kinetic model for a cycling adsorbent/photocatalyst combination for formaldehyde removal in indoor air (Chin et al. J. Catalysis 2006, 237, 29-37) was previously developed in our lab, demonstrating agreement with lab-scale batch operation data of other researchers (Shiraishi et al. Chem. Engineer. Sci. 2003, 58, 929-934). Model parameters evaluated included adsorption equilibrium and rate constants for the adsorbent (activated carbon) honeycomb rotor, and catalytic rate constant for pseudo-first-order formaldehyde destruction in the titanium dioxide photoreactor. This paper explores design consequences for this novel system. In particular, the batch parameter values are used to model both adsorbent and photocatalyst behavior for continuous operation in typical residential home challenges. Design variables, including realistic make-up air fraction, adsorbent honeycomb rotation speed, and formaldehyde source emission rate, are considered to evaluate the ability of the system to achieve World Health Organization pollutant guidelines. In all circumstances, the size of the required rotating adsorbent bed and photoreactor for single-stage operation and the resultant formaldehyde concentration in the home are calculated. The ability of how well such a system might be accommodated within the typical dimensions of commercial ventilation ducts is also considered.

  1. Excess plutonium disposition: The deep borehole option

    SciTech Connect

    Ferguson, K.L.

    1994-08-09

    This report reviews the current status of technologies required for the disposition of plutonium in Very Deep Holes (VDH). It is in response to a recent National Academy of Sciences (NAS) report which addressed the management of excess weapons plutonium and recommended three approaches to the ultimate disposition of excess plutonium: (1) fabrication and use as a fuel in existing or modified reactors in a once-through cycle, (2) vitrification with high-level radioactive waste for repository disposition, (3) burial in deep boreholes. As indicated in the NAS report, substantial effort would be required to address the broad range of issues related to deep bore-hole emplacement. Subjects reviewed in this report include geology and hydrology, design and engineering, safety and licensing, policy decisions that can impact the viability of the concept, and applicable international programs. Key technical areas that would require attention should decisions be made to further develop the borehole emplacement option are identified.

  2. Plutonium Vulnerability Management Plan

    SciTech Connect

    1995-03-01

    This Plutonium Vulnerability Management Plan describes the Department of Energy`s response to the vulnerabilities identified in the Plutonium Working Group Report which are a result of the cessation of nuclear weapons production. The responses contained in this document are only part of an overall, coordinated approach designed to enable the Department to accelerate conversion of all nuclear materials, including plutonium, to forms suitable for safe, interim storage. The overall actions being taken are discussed in detail in the Department`s Implementation Plan in response to the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 94-1. This is included as Attachment B.

  3. Progress on plutonium stabilization

    SciTech Connect

    Hurt, D.

    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.

  4. PLUTONIUM SEPARATION METHOD

    DOEpatents

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

    1958-11-18

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

  5. PLUTONIUM ELECTROREFINING CELLS

    DOEpatents

    Mullins, L.J. Jr.; Leary, J.A.; Bjorklund, C.W.; Maraman, W.J.

    1963-07-16

    Electrorefining cells for obtaining 99.98% plutonium are described. The cells consist of an impure liquid plutonium anode, a molten PuCl/sub 3/-- alkali or alkaline earth metal chloanode, a molten PuCl/sub 3/-alkali or alkaline earth metal chloride electrolyte, and a nonreactive cathode, all being contained in nonreactive ceramic containers which separate anode from cathode by a short distance and define a gap for the collection of the purified liquid plutonium deposited on the cathode. Important features of these cells are the addition of stirrer blades on the anode lead and a large cathode surface to insure a low current density. (AEC)

  6. Plutonium radiation surrogate

    DOEpatents

    Frank, Michael I [Dublin, CA

    2010-02-02

    A self-contained source of gamma-ray and neutron radiation suitable for use as a radiation surrogate for weapons-grade plutonium is described. The source generates a radiation spectrum similar to that of weapons-grade plutonium at 5% energy resolution between 59 and 2614 keV, but contains no special nuclear material and emits little .alpha.-particle radiation. The weapons-grade plutonium radiation surrogate also emits neutrons having fluxes commensurate with the gamma-radiation intensities employed.

  7. Plutonium dissolution process

    DOEpatents

    Vest, Michael A.; Fink, Samuel D.; Karraker, David G.; Moore, Edwin N.; Holcomb, H. Perry

    1996-01-01

    A two-step process for dissolving plutonium metal, which two steps can be carried out sequentially or simultaneously. Plutonium metal is exposed to a first mixture containing approximately 1.0M-1.67M sulfamic acid and 0.0025M-0.1M fluoride, the mixture having been heated to a temperature between 45.degree. C. and 70.degree. C. The mixture will dissolve a first portion of the plutonium metal but leave a portion of the plutonium in an oxide residue. Then, a mineral acid and additional fluoride are added to dissolve the residue. Alteratively, nitric acid in a concentration between approximately 0.05M and 0.067M is added to the first mixture to dissolve the residue as it is produced. Hydrogen released during the dissolution process is diluted with nitrogen.

  8. Plutonium: Requiem or reprieve

    SciTech Connect

    Pillay, K.K.S.

    1996-01-01

    Many scientific discoveries have had profound effects on humanity and its future. However, the discovery of fissionable characteristics of a man-made element, plutonium, discovered in 1941 by Glenn Seaborg and associates, has probably had the greatest impact on world affairs. Although about 20 new elements have been synthesized since 1940, element 94 unarguably had the most dramatic impact when it was introduced to the world as the core of the nuclear bomb dropped on Nagasaki. Ever since, large quantities of this element have been produced, and it has had a major role in maintaining peace during the past 50 years. in addition, the rapid spread of nuclear power technology worldwide contributed to major growth in the production of plutonium as a by-product. This article discusses the following issues related to plutonium: plutonium from Nuclear Power Generation; environmental safety and health issues; health effects; safeguards issues; extended storage; disposal options.

  9. Preserving Plutonium-244 as a National Asset

    SciTech Connect

    Patton, Bradley D; Alexander, Charles W; Benker, Dennis; Collins, Emory D; Romano, Catherine E; Wham, Robert M

    2011-01-01

    Plutonium-244 (244 Pu) is an extremely rare and long-lived isotope of plutonium with a half-life of 80 million years. Measureable amounts of 244 Pu are found in neither reactor-grade nor weapons-grade plutonium. Production of this isotope requires a very high thermal flux to permit the two successive neutron captures that convert 242 Pu to 243 Pu to 244 Pu, particularly given the short (about 5 hour) half-life of 243 Pu. Such conditions simply do not exist in plutonium production processes. Therefore, 244 Pu is ideal for precise radiochemical analyses measuring plutonium material properties and isotopic concentrations in items containing plutonium. Isotope dilution mass spectrometry is about ten times more sensitive when using 244 Pu rather than 242 Pu for determining plutonium isotopic content. The isotope can also be irradiated in small quantities to produce superheavy elements. The majority of the existing global inventory of 244 Pu is contained in the outer housing of Mark-18A targets at the Savannah River Site (SRS). The total inventory is about 20 grams of 244 Pu in about 400 grams of plutonium distributed among the 65 targets. Currently, there are no specific plans to preserve these targets. Although the cost of separating and preserving this material would be considerable, it is trivial in comparison to new production costs. For all practical purposes, the material is irreplaceable, because new production would cost billions of dollars and require a series of irradiation and chemical separation cycles spanning up to 50 years. This paper will discuss a set of options for overcoming the significant challenges to preserve the 244 Pu as a National Asset: (1) the need to relocate the material from SRS in a timely manner, (2) the need to reduce the volume of material to the extent possible for storage, and (3) the need to establish an operational capability to enrich the 244 Pu in significant quantities. This paper suggests that if all the Mark-18A plutonium is

  10. The carbonate complexation of plutonium(IV)

    SciTech Connect

    Hobart, D E; Palmer, P D; Newton, T W

    1985-01-01

    Plutonium(IV) carbonate complexes are expected to be of particular importance in typical groundwaters at the Yucca Mountain site of the candidate nuclear waste repository being studied by the Nevada Nuclear Waste Storage Investigations Project. The chemistry of these complexes is also important in the areas of nuclear fuel reprocessing and purification, actinide separations, and environmental studies. This report describes initial experiments performed to determine the identity and equilibrium quotients of plutonium(IV) carbonate complexes. These experiments were performed at pH values between 7.2 and 9.6 using a spectrophotometric method. In addition, a brief review of the published literature on Pu(IV) carbonate complexes is presented. Since Pu(IV) exhibits low solubility in the near-neutral pH range, a complex-competition reaction where citrate ligands compete with carbonate ions for the plutonium will be employed. This will permit us to study the pure carbonate system; study the mixed carbonate/citrate system, and confirm and extend the literature work on the pure citrate system. The current experiments have demonstrated the existence of at least three distinct species in the pH region studied. This work will continue in the extended study of the pure citrate system, followed by the investigation of the citrate/carbonate complex/competition reaction. 9 refs., 4 figs., 2 tabs.

  11. Advanced fuels for plutonium management in pressurized water reactors

    NASA Astrophysics Data System (ADS)

    Vasile, A.; Dufour, Ph; Golfier, H.; Grouiller, J. P.; Guillet, J. L.; Poinot, Ch; Youinou, G.; Zaetta, A.

    2003-06-01

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

  12. Innovative concepts for the plutonium facilities at La Hague

    NASA Astrophysics Data System (ADS)

    Gillet, B.; Gresle, A.; Drain, F.

    2000-07-01

    The commercial strategy of COGEMA is now based on a combined reprocessing-conditioning-recycling proposal: the reprocessing plants at La Hague ensure plutonium recovery, purification, and conditioning, and the mixed-oxide (MOX) fuel fabrication plant MELOX at Marcoule ensures its recycling into MOX nuclear fuel. This strategy is enabled thanks to technological and process innovations resulting from an extensive R&D program over the past twenty years. First, the UP3 plant (the T4 plutonium facility in particular) have benefited from these innovations. Second, experience gained from the UP3 plant and new developments have been integrated into the UP2-800 plutonium facilities (R4, URP, UCD) to continue cost reduction and performance optimizing.

  13. Plutonium Speciation, Solubilization, and Migration in soils

    SciTech Connect

    Neu, Mary; Haire, Richard G.

    1999-06-01

    The DOE is currently conducting cleanup activities at its nuclear weapons development sites, many of which have accumulated plutonium in soils for 50 years. To properly control Pu migration in soils within Federal sites and onto public lands, better evaluate the public risk, and design effective remediation strategies, a fundamental understanding of Pu speciation and environmental transport, and release mechanisms is needed. The key scientific goals of this project are: to determine Pu concentrations and speciation at a contaminated DOE site; to study the formation, stability, and structural and spectroscopic features of environmentally relevant Pu species; to determine the mechanism(s) of interaction between Pu and Mn/Fe minerals and the potential release of Pu via redox cycling; and to model the environmental behavior of plutonium. Our long-term goal is to use characterization, thermodynamic, mineral interaction, and mobility data to develop better models of radionuclide transport and risk assessment, and to enable the development of science-based decontamination strategies. This research will fill important gaps between basic actinide science and the problems impeding site clean-up, plutonium disposition, and accurate risk assessment. Information gained will allow for the development of technologies and clean-up approaches targeting particular plutonium contaminants and improved assessment of risks associated with actinide migration, site remediation, and decontamination. By combining very specific study of plutonium at the Rocky Flats Environmental Test Site (RFETS), a well characterized contaminated site, with laboratory studies on the most important plutonium and mineral component systems, we will provide essential knowledge of contaminant characteristics and distinguish critical geochemical processes and mechanisms.

  14. METHOD OF MAKING PLUTONIUM DIOXIDE

    DOEpatents

    Garner, C.S.

    1959-01-13

    A process is presented For converting both trivalent and tetravalent plutonium oxalate to substantially pure plutonium dioxide. The plutonium oxalate is carefully dried in the temperature range of 130 to300DEC by raising the temperature gnadually throughout this range. The temperature is then raised to 600 C in the period of about 0.3 of an hour and held at this level for about the same length of time to obtain the plutonium dioxide.

  15. METHOD OF PRODUCING PLUTONIUM TETRAFLUORIDE

    DOEpatents

    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.

  16. Plutonium Disposition Now!

    SciTech Connect

    Buckner, M.R.

    1995-05-24

    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.

  17. Aqueous Solution Chemistry of Plutonium

    SciTech Connect

    Clark, David L.

    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.

  18. Lithium metal reduction of plutonium oxide to produce plutonium metal

    DOEpatents

    Coops, Melvin S.

    1992-01-01

    A method is described for the chemical reduction of plutonium oxides to plutonium metal by the use of pure lithium metal. Lithium metal is used to reduce plutonium oxide to alpha plutonium metal (alpha-Pu). The lithium oxide by-product is reclaimed by sublimation and converted to the chloride salt, and after electrolysis, is removed as lithium metal. Zinc may be used as a solvent metal to improve thermodynamics of the reduction reaction at lower temperatures. Lithium metal reduction enables plutonium oxide reduction without the production of huge quantities of CaO--CaCl.sub.2 residues normally produced in conventional direct oxide reduction processes.

  19. Purification of U and Pu from Bulk Environmental Samples for Analysis by MC-ICPMS

    SciTech Connect

    Williams, R W; Genetti, V; Ramon, E

    2005-02-23

    This procedure gives the methods used at LLNL for the purification of uranium and plutonium from bulk environmental samples provided by the IAEA through the DOE Network of Analytical Laboratories (NWAL).

  20. Plutonium 239 Equivalency Calculations

    SciTech Connect

    Wen, J

    2011-05-31

    This document provides the basis for converting actual weapons grade plutonium mass to a plutonium equivalency (PuE) mass of Plutonium 239. The conversion can be accomplished by performing calculations utilizing either: (1) Isotopic conversions factors (CF{sub isotope}), or (2) 30-year-old weapons grade conversion factor (CF{sub 30 yr}) Both of these methods are provided in this document. Material mass and isotopic data are needed to calculate PuE using the isotopic conversion factors, which will provide the actual PuE value at the time of calculation. PuE is the summation of the isotopic masses times their associated isotopic conversion factors for plutonium 239. Isotopic conversion factors are calculated by a normalized equation, relative to Plutonium 239, of specific activity (SA) and cumulated dose inhalation affects based on 50-yr committed effective dose equivalent (CEDE). The isotopic conversion factors for converting weapons grade plutonium to PuE are provided in Table-1. The unit for specific activity (SA) is curies per gram (Ci/g) and the isotopic SA values come from reference [1]. The cumulated dose inhalation effect values in units of rem/Ci are based on 50-yr committed effective dose equivalent (CEDE). A person irradiated by gamma radiation outside the body will receive a dose only during the period of irradiation. However, following an intake by inhalation, some radionuclides persist in the body and irradiate the various tissues for many years. There are three groups CEDE data representing lengths of time of 0.5 (D), 50 (W) and 500 (Y) days, which are in reference [2]. The CEDE values in the (W) group demonstrates the highest dose equivalent value; therefore they are used for the calculation.

  1. SULFIDE METHOD PLUTONIUM SEPARATION

    DOEpatents

    Duffield, R.B.

    1958-08-12

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

  2. Superconductivity in plutonium compounds

    NASA Astrophysics Data System (ADS)

    Sarrao, J. L.; Bauer, E. D.; Mitchell, J. N.; Tobash, P. H.; Thompson, J. D.

    2015-07-01

    Although the family of plutonium-based superconductors is relatively small, consisting of four compounds all of which crystallize in the tetragonal HoCoGa5 structure, these materials serve as an important bridge between the known Ce- and U-based heavy fermion superconductors and the high-temperature cuprate superconductors. Further, the partial localization of 5f electrons that characterizes the novel electronic properties of elemental plutonium appears to be central to the relatively high superconducting transition temperatures that are observed in PuCoGa5, PuRhGa5, PuCoIn5, and PuRhIn5.

  3. Management of plutonium contaminated waste

    SciTech Connect

    Grover, J.R.

    1982-01-01

    This study surveys the current management schemes for plutonium contaminated wastes arising from a reference mixed oxide fuel fabrication plant, and identifies possible areas of future research. It also outlines strategies for the future management of plutonium contaminated wastes. Topics of discussion include: the quantities and characteristics of various plutonium contaminated wastes produced by a plant; the current waste management practices for both solid and liquid plutonium contaminated wastes, considering measurement methods, transportation, storage and disposal; current practice for the problems of decommissioning and decontamination, and possible methods for the recovery of plutonium contaminated wastes.

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

    SciTech Connect

    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. Automated controlled-potential coulometer for plutonium determination

    SciTech Connect

    Hollen, R.M.; Jackson, D.D.

    1981-05-01

    The automated controlled-potential coulometer for the determination of plutonium described in this report is the second in a series of automated instruments designed to determine plutonium and uranium contents in nuclear fuel cycle materials. The measurement precision of the instrument is 0.1% relative standard deviation at the 5-mg plutonium level. A highly selective method of analysis was developed, involving reduction of plutonium to Pu(III) in a 5.5 M hydrochloric acid-0.015 M sulfamic acid electrolyte; oxidation of diverse ions, but not Pu(III); addition of phosphate complexant to reduce the Pu(III)-Pu(IV) potential; and oxidation of Pu(III) to Pu(IV) as the measurement step. Construction details of the mechanical and electrical systems of the instrument and control-system software are described, along with instrument preoperational adjustments and tests and sample analysis operations.

  6. Plutonium Disposition by Immobilization

    SciTech Connect

    Gould, T.; DiSabatino, A.; Mitchell, M.

    2000-03-07

    The ultimate goal of the Department of Energy (DOE) Immobilization Project is to develop, construct, and operate facilities that will immobilize between 17 to 50 tonnes (MT) of U.S. surplus weapons-usable plutonium materials in waste forms that meet the ''spent fuel'' standard and are acceptable for disposal in a geologic repository. Using the ceramic can-in-canister technology selected for immobilization, surplus plutonium materials will be chemically combined into ceramic forms which will be encapsulated within large canisters of high level waste (HLW) glass. Deployment of the immobilization capability should occur by 2008 and be completed within 10 years. In support of this goal, the DOE Office of Fissile Materials Disposition (MD) is conducting development and testing (D&T) activities at four DOE laboratories under the technical leadership of Lawrence Livermore National Laboratory (LLNL). The Savannah River Site has been selected as the site for the planned Plutonium Immobilization Plant (PIP). The D&T effort, now in its third year, will establish the technical bases for the design, construction, and operation of the U. S. capability to immobilize surplus plutonium in a suitable and cost-effective manner. Based on the D&T effort and on the development of a conceptual design of the PIP, automation is expected to play a key role in the design and operation of the Immobilization Plant. Automation and remote handling are needed to achieve required dose reduction and to enhance operational efficiency.

  7. Plutonium: An introduction

    SciTech Connect

    Condit, R.H.

    1993-10-01

    This report is a summary of the history and properties of plutonium. It presents information on the atoms, comparing chemical and nuclear properties. It looks at the history of the atom, including its discovery and production methods. It summarizes the metallurgy and chemistry of the element. It also describes means of detecting and measuring the presence and quantity of the element.

  8. Atomic spectrum of plutonium

    SciTech Connect

    Blaise, J.; Fred, M.; Gutmacher, R.G.

    1984-08-01

    This report contains plutonium wavelengths, energy level classifications, and other spectroscopic data accumulated over the past twenty years at Laboratoire Aime Cotton (LAC) Argonne National Laboratory (ANL), and Lawrence Livermore National Laboratory (LLNL). The primary purpose was term analysis: deriving the energy levels in terms of quantum numbers and electron configurations, and evaluating the Slater-Condon and other parameters from the levels.

  9. Single-step purification of recombinant proteins using elastin-like peptide-mediated inverse transition cycling and self-processing module from Neisseria meningitides FrpC.

    PubMed

    Liu, Wen-Jun; Wu, Qian; Xu, Bi; Zhang, Xin-Yu; Xia, Xiao-Li; Sun, Huai-Chang

    2014-06-01

    Purification of recombinant proteins is a major task and challenge in biotechnology and medicine. In this paper we report a novel single-step recombinant protein purification system which was based on elastin-like peptide (ELP)-mediated reversible phase transition and FrpC self-processing module (SPM)-mediated cleavage. After construction of a SPM-ELP fusion expression vector, we cloned the coding sequence for green fluorescent protein (GFP), the Fc portion of porcine IgG (pFc) or human β defensin 3 (HBD3) into the vector, transformed the construct into Escherichia coli, and induced the fusion protein expression with IPTG. The target-SPM-ELP fusion proteins GFP-SPM-ELP, Fc-SPM-ELP and HBD3-SPM-ELP were expressed in a soluble form and efficiently purified from the clarified cell extracts by two rounds of inverse transition cycling (ITC). Under the optimized conditions, the SPM-mediated cleavage efficiencies for the three fusion proteins ranged from 92% to 93%. After an additional round of ITC, the target proteins GFP, pFc and HBD3 were recovered with purities ranging from 90% to 100% and yields ranging from 1.1 to 36mg/L in shake flasks. The endotoxin levels in all of the three target proteins were <0.03EU/mg. The three target proteins were functionally active with the expected molecular weights. These experimental results confirmed the high specificity and efficiency of SPM-mediated cleavage, and suggested the applicability of SPM-ELP fusion system for purification of recombinant proteins.

  10. Plutonium Finishing Plant. Interim plutonium stabilization engineering study

    SciTech Connect

    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.

  11. Stabilization and immobilization of military plutonium: A non-proliferation perspective

    SciTech Connect

    Leventhal, P.

    1996-05-01

    The Nuclear Control Institute welcomes this DOE-sponsored technical workshop on stabilization and immobilization of weapons plutonium (W Pu) because of the significant contribution it can make toward the ultimate non-proliferation objective of eliminating weapons-usable nuclear material, plutonium and highly enriched uranium (HEU), from world commerce. The risk of theft or diversion of these materials warrants concern, as only a few kilograms in the hands of terrorists or threshold states would give them the capability to build nuclear weapons. Military plutonium disposition questions cannot be addressed in isolation from civilian plutonium issues. The National Academy of Sciences has urged that {open_quotes}further steps should be taken to reduce the proliferation risks posed by all of the world`s plutonium stocks, military and civilian, separated and unseparated...{close_quotes}. This report discusses vitrification and a mixed oxide fuels option, and the effects of disposition choices on civilian plutonium fuel cycles.

  12. 4. VIEW OF PLUTONIUM CANISTER ON CHAINVEYOR. SCRAP PLUTONIUM WAS ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    4. VIEW OF PLUTONIUM CANISTER ON CHAINVEYOR. SCRAP PLUTONIUM WAS COLLECTED INTO CANS AT INDIVIDUAL WORKSTATIONS. THE CANS WERE TRANSFERRED VIA THE CHAIN CONVEYOR TO A WORKSTATION IN MODULE C WHERE THE MATERIAL WAS COMPRESSED INTO BRIQUETTES FOR LATER USE. (6/20/93) - Rocky Flats Plant, Plutonium Manufacturing Facility, North-central section of Plant, just south of Building 776/777, Golden, Jefferson County, CO

  13. Plutonium age dating reloaded

    NASA Astrophysics Data System (ADS)

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

    2014-05-01

    Although the age determination of plutonium is and has been a pillar of nuclear forensic investigations for many years, additional research in the field of plutonium age dating is still needed and leads to new insights as the present work shows: Plutonium is commonly dated with the help of the 241Pu/241Am chronometer using gamma spectrometry; in fewer cases the 240Pu/236U chronometer has been used. The age dating results of the 239Pu/235U chronometer and the 238Pu/234U chronometer are scarcely applied in addition to the 240Pu/236U chronometer, although their results can be obtained simultaneously from the same mass spectrometric experiments as the age dating result of latter. The reliability of the result can be tested when the results of different chronometers are compared. The 242Pu/238U chronometer is normally not evaluated at all due to its sensitivity to contamination with natural uranium. This apparent 'weakness' that renders the age dating results of the 242Pu/238U chronometer almost useless for nuclear forensic investigations, however turns out to be an advantage looked at from another perspective: the 242Pu/238U chronometer can be utilized as an indicator for uranium contamination of plutonium samples and even help to identify the nature of this contamination. To illustrate this the age dating results of all four Pu/U clocks mentioned above are discussed for one plutonium sample (NBS 946) that shows no signs of uranium contamination and for three additional plutonium samples. In case the 242Pu/238U chronometer results in an older 'age' than the other Pu/U chronometers, contamination with either a small amount of enriched or with natural or depleted uranium is for example possible. If the age dating result of the 239Pu/235U chronometer is also influenced the nature of the contamination can be identified; enriched uranium is in this latter case a likely cause for the missmatch of the age dating results of the Pu/U chronometers.

  14. Surprising Coordination for Plutonium in the First Plutonium (III) Borate

    SciTech Connect

    Wang, Shuao; Alekseev, Evgeny V.; Depmeier, Wulf; Albrecht-Schmitt, Thomas E.

    2011-02-22

    The first plutonium(III) borate, Pu2[B12O18(OH)4Br2(H2O)3]·0.5H2O, has been prepared by reacting plutonium(III) with molten boric acid under strictly anaerobic conditions. This compound contains a three-dimensional polyborate network with triangular holes that house the plutonium(III) sites. The plutonium sites in this compound are 9- and 10-coordinate and display atypical geometries.

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

    SciTech Connect

    Navratil, J D

    1985-05-13

    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.

  16. Fast Thorium Molten Salt Reactors Started with Plutonium

    SciTech Connect

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

    2006-07-01

    One of the pending questions concerning Molten Salt Reactors based on the {sup 232}Th/{sup 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 {sup 233}U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing {sup 233}U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce {sup 233}U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/{sup 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 {sup 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 {sup 233}U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with {sup 233}U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)

  17. Plutonium Speciation, Solubilization and Migration in Soils

    SciTech Connect

    Neu, Mary P.; Haire, Richard G.

    2002-06-01

    The DOE is currently conducting cleanup activities at its nuclear weapons development sites, many of which have accumulated plutonium in soils for 50 years. To properly control Pu migration in soils within Federal sites and onto public lands, better evaluate the public risk, and design effective remediation strategies, a fundamental understanding of Pu speciation and environmental transport is needed. The key scientific goals of this project are: to determine Pu concentrations and speciation at contaminated DOE sites; to study the formation, stability, and structural and spectroscopic features of environmentally relevant Pu species; to determine the mechanism(s) of interaction between Pu and Mn/Fe minerals and the potential release of Pu via redox cycling; and to model the environmental behavior of plutonium. Our goal is to use characterization, thermodynamic, mineral interaction, and mobility data to develop better models of radionuclide transport and risk assessment, and to enable the development of science-based decontamination strategies.

  18. Gamma radiation characteristics of plutonium dioxide fuel

    NASA Technical Reports Server (NTRS)

    Gingo, P. J.

    1969-01-01

    Investigation of plutonium dioxide as an isotopic fuel for Radioisotope Thermoelectric Generators yielded the isotopic composition of production-grade plutonium dioxide fuel, sources of gamma radiation produced by plutonium isotopes, and the gamma flux at the surface.

  19. Plutonium process monitoring (PPM) system

    NASA Astrophysics Data System (ADS)

    Wong, A. S.; Ricketts, T. E.; Pansoy-Hejlvik, M. E.; Ramsey, K. B.; Hansel, K. M.; Romero, M. K.

    2000-07-01

    In mid-1980, Marsh and Pope developed an online gamma system to monitor americium, uranium and plutonium gamma rays during anion-exchange process for plutonium aqueous recovery operations. It has been shown that the real-time elution profiles of actinide impurities are important for plutonium loss via break-through, waste minimization, and process monitoring. However, the current monitoring equipment and data acquisition software are obsolete and are frequently problematic. In 1999, we redesigned the on-line gamma monitoring system in collaboration with Perkin-Elmer ORTEC (Oak Ridge, TN) to enhance and upgrade the current system. This paper describes the new integrated plutonium process monitoring (PPM) system for the aqueous plutonium recovery and anion-exchange processes at the Los Alamos Plutonium Facility.

  20. Plutonium recovery from organic materials

    DOEpatents

    Deaton, R.L.; Silver, G.L.

    1973-12-11

    A method is described for removing plutonium or the like from organic material wherein the organic material is leached with a solution containing a strong reducing agent such as titanium (III) (Ti/sup +3None)/, chromium (II) (Cr/ sup +2/), vanadium (II) (V/sup +2/) ions, or ferrous ethylenediaminetetraacetate (EDTA), the leaching yielding a plutonium-containing solution that is further processed to recover plutonium. The leach solution may also contain citrate or tartrate ion. (Official Gazette)

  1. PROCESS OF PRODUCING SHAPED PLUTONIUM

    DOEpatents

    Anicetti, R.J.

    1959-08-11

    A process is presented for producing and casting high purity plutonium metal in one step from plutonium tetrafluoride. The process comprises heating a mixture of the plutonium tetrafluoride with calcium while the mixture is in contact with and defined as to shape by a material obtained by firing a mixture consisting of calcium oxide and from 2 to 10% by its weight of calcium fluoride at from 1260 to 1370 deg C.

  2. Manufacturing of Plutonium Tensile Specimens

    SciTech Connect

    Knapp, Cameron M

    2012-08-01

    Details workflow conducted to manufacture high density alpha Plutonium tensile specimens to support Los Alamos National Laboratory's science campaigns. Introduces topics including the metallurgical challenge of Plutonium and the use of high performance super-computing to drive design. Addresses the utilization of Abaqus finite element analysis, programmable computer numerical controlled (CNC) machining, as well as glove box ergonomics and safety in order to design a process that will yield high quality Plutonium tensile specimens.

  3. PLUTONIUM-URANIUM ALLOY

    DOEpatents

    Coffinberry, A.S.; Schonfeld, F.W.

    1959-09-01

    Pu-U-Fe and Pu-U-Co alloys suitable for use as fuel elements tn fast breeder reactors are described. The advantages of these alloys are ease of fabrication without microcracks, good corrosion restatance, and good resistance to radiation damage. These advantages are secured by limitation of the zeta phase of plutonium in favor of a tetragonal crystal structure of the U/sub 6/Mn type.

  4. MOLDS FOR CASTING PLUTONIUM

    DOEpatents

    Anderson, J.W.; Miley, F.; Pritchard, W.C.

    1962-02-27

    A coated mold for casting plutonium comprises a mold base portion of a material which remains solid and stable at temperatures as high as the pouring temperature of the metal to be cast and having a thin coating of the order of 0.005 inch thick on the interior thereof. The coating is composed of finely divided calcium fluoride having a particle size of about 149 microns. (AEC)

  5. Welding Plutonium Storage Containers

    SciTech Connect

    HUDLOW, SL

    2004-04-20

    The outer can welder (OCW) in the FB-Line Facility at the Savannah River Site (SRS) is a Gas Tungsten Arc Weld (GTAW) system used to create outer canisters compliant with the Department of Energy 3013 Standard, DOE-STD-3013-2000, Stabilization, Packaging, and Storage of Plutonium-Bearing Materials. The key welding parameters controlled and monitored on the outer can welder Data Acquisition System (DAS) are weld amperage, weld voltage, and weld rotational speed. Inner 3013 canisters from the Bagless Transfer System that contain plutonium metal or plutonium oxide are placed inside an outer 3013 canister. The canister is back-filled with helium and welded using the outer can welder. The completed weld is screened to determine if it is satisfactory by reviewing the OCW DAS key welding parameters, performing a helium leak check, performing a visual examination by a qualified weld inspector, and performing digital radiography of the completed weld. Canisters with unsatisfactory welds are cut open and repackaged. Canisters with satisfactory welds are deemed compliant with the 3013 standard for long-term storage.

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

    SciTech Connect

    Bronson, M.C.

    1997-10-01

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

  7. From separations to reconstitution - a short history of Plutonium in the U.S. and Russia

    SciTech Connect

    Gray, L W

    1999-04-15

    During the cold war plutonium was produced in reactors in both the US and Russia. It was then separated from the residual uranium and fission products by a variety of precipitation processes, such as Bismuth Phosphate, Redox, Butex, Purex, etc. in the US and uranium acetate and Purex in Russia. After a period of time in the field, plutonium weapons were recycled and the plutonium re-purified and returned to weapons. purification was accomplished by a variety of aqueous and molten salt processes, such as nitric-hydrofluoric acid dissolution followed by anion exchange, Purex modifications, molten salt extraction, electrorefining, etc. in the US and nitric acid dissolution or sodium hydroxide fusion followed by anion exchange in Russia. At the end of the Cold War, plutonium production of weapons-grade plutonium was cut off in the US and is expected to be cut off in Russia shortly after the turn of the century. Now both countries are looking at methods to reconstitute plutonium with fission products to render it no longer useful for nuclear weapons. These methods include immobilization in a ceramic matrix and then encasement in fission product laden glass, irradiation of MOX fuel, and disposal as waste in WIPP in the US and irradiation of MOX fuel in Russia. This paper details the contrast between the treatment of plutonium during the cold war and after the cold war was over.

  8. AUTOSEP, an automated system for the quantitative ion exchange separation of plutonium from impurities

    SciTech Connect

    Freeman, B.P.; Weiss, J.R.; Pietri, C.E.

    1981-08-01

    The anion exchange separation of interfering impurities from plutonium in the Dowex-1 8 N HNO/sub 3/ system is a reliable purification method. An automated system, AUTOSEP, based on this manual procedure has been developed for greater productivity. Samples, in groups of ten, each containing 5--10 mg of plutonium are processed automatically in this system. AUTOSEP incorporates a means of programming reagent delivery, adjustment of the sample to the plutonium (IV) oxidation state via Fe(II) reduction/HNO/sub 3/ oxidation, plutonium sorption on the resin in 8 N HNO/sub 3/, washing of the resin bed with 8 N HNO/sub 3/ to remove impurities, elution of the purified plutonium with 0.36 N HCl--0.01 N HF, and waste effluent disposal. The reagents are delivered by gravity from a module whose only moving parts are rotary valves. The eluted plutonium solutions are collected for subsequent controlled-potential coulometric analysis. The average recovery of plutonium determined by controlled-potential coulometry using this apparatus was 100.00% with a relative standard deviation of 0.10%.

  9. Plutonium worker dosimetry.

    PubMed

    Birchall, Alan; Puncher, M; Harrison, J; Riddell, A; Bailey, M R; Khokryakov, V; Romanov, S

    2010-05-01

    Epidemiological studies of the relationship between risk and internal exposure to plutonium are clearly reliant on the dose estimates used. The International Commission on Radiological Protection (ICRP) is currently reviewing the latest scientific information available on biokinetic models and dosimetry, and it is likely that a number of changes to the existing models will be recommended. The effect of certain changes, particularly to the ICRP model of the respiratory tract, has been investigated for inhaled forms of (239)Pu and uncertainties have also been assessed. Notable effects of possible changes to respiratory tract model assumptions are (1) a reduction in the absorbed dose to target cells in the airways, if changes under consideration are made to the slow clearing fraction and (2) a doubling of absorbed dose to the alveolar region for insoluble forms, if evidence of longer retention times is taken into account. An important factor influencing doses for moderately soluble forms of (239)Pu is the extent of binding of dissolved plutonium to lung tissues and assumptions regarding the extent of binding in the airways. Uncertainty analyses have been performed with prior distributions chosen for application in epidemiological studies. The resulting distributions for dose per unit intake were lognormal with geometric standard deviations of 2.3 and 2.6 for nitrates and oxides, respectively. The wide ranges were due largely to consideration of results for a range of experimental data for the solubility of different forms of nitrate and oxides. The medians of these distributions were a factor of three times higher than calculated using current default ICRP parameter values. For nitrates, this was due to the assumption of a bound fraction, and for oxides due mainly to the assumption of slower alveolar clearance. This study highlights areas where more research is needed to reduce biokinetic uncertainties, including more accurate determination of particle transport rates

  10. Ultratrace Analysis of Uranium and Plutonium By Mass Spectrometry

    SciTech Connect

    Wacker, John F.; Wogman, Ned A.; Olsen, Khris B.; Petersen, Steven L.; Farmer, O T.; Kelley, James M.; Eiden, Greg C.; Maiti, Tapas C.

    2003-01-01

    At the Pacific Northwest National Laboratory (PNNL), we have developed highly sensitive methods to analyze uranium and plutonium in environmental samples. The development of an ultratrace analysis capability for measuring uranium and plutonium has arisen from a need to detect and characterize environmental samples for signatures associated with nuclear industry processes. Our most sensitive well-developed methodologies employ thermal ionization mass spectrometry (TIMS), however, recent advances in inductively coupled plasma mass spectrometry (ICP-MS) have shown considerable promise for use in detecting uranium and plutonium at ultratrace levels. The work at PNNL has included the development of both chemical separation and purification techniques, as well as the development of mass spectrometric instrumentation and techniques. At the heart of our methodology for TIMS analysis is a procedure that utilizes 100-microliter-volumes of analyte for chemical processing to purify, separate, and load actinide elements into resin beads for subsequent mass spectrometric analysis. The resin bead technique has been combined with a thorough knowledge of the physicochemistry of thermal ion emission to achieve femtogram detection limits for the TIMS analysis of plutonium in environmental samples.

  11. METHOD FOR OBTAINING PLUTONIUM METAL AND ALLOYS OF PLUTONIUM FROM PLUTONIUM TRICHLORIDE

    DOEpatents

    Reavis, J.G.; Leary, J.A.; Maraman, W.J.

    1962-11-13

    A process is given for both reducing plutonium trichloride to plutonium metal using cerium as the reductant and simultaneously alloying such plutonium metal with an excess of cerium or cerium and cobalt sufficient to yield the desired nuclear reactor fuel composition. The process is conducted at a temperature from about 550 to 775 deg C, at atmospheric pressure, without the use of booster reactants, and a substantial decontamination is effected in the product alloy of any rare earths which may be associated with the source of the plutonium. (AEC)

  12. PREPARATION OF HALIDES OF PLUTONIUM

    DOEpatents

    Garner, C.S.; Johns, I.B.

    1958-09-01

    A dry chemical method is described for preparing plutonium halides, which consists in contacting plutonyl nitrate with dry gaseous HCl or HF at an elevated temperature. The addition to the reaction gas of a small quantity of an oxidizing gas or a reducing gas will cause formation of the tetra- or tri-halide of plutonium as desired.

  13. Photochemical preparation of plutonium pentafluoride

    DOEpatents

    Rabideau, Sherman W.; Campbell, George M.

    1987-01-01

    The novel compound plutonium pentafluoride may be prepared by the photodissociation of gaseous plutonium hexafluoride. It is a white solid of low vapor pressure, which consists predominantly of a face-centered cubic structure with a.sub.o =4.2709.+-.0.0005 .ANG..

  14. PLUTONIUM-URANIUM-TITANIUM ALLOYS

    DOEpatents

    Coffinberry, A.S.

    1959-07-28

    A plutonium-uranium alloy suitable for use as the fuel element in a fast breeder reactor is described. The alloy contains from 15 to 60 at.% titanium with the remainder uranium and plutonium in a specific ratio, thereby limiting the undesirable zeta phase and rendering the alloy relatively resistant to corrosion and giving it the essential characteristic of good mechanical workability.

  15. METHOD OF REDUCING PLUTONIUM COMPOUNDS

    DOEpatents

    Johns, I.B.

    1958-06-01

    A method is described for reducing plutonium compounds in aqueous solution from a higher to a lower valence state. This reduction of valence is achieved by treating the aqueous solution of higher valence plutonium compounds with hydrogen in contact with an activated platinum catalyst.

  16. SEPARATION OF PLUTONIUM FROM URANIUM

    DOEpatents

    Feder, H.M.; Nuttall, R.L.

    1959-12-15

    A process is described for extracting plutonium from powdered neutron- irradiated urarium metal by contacting the latter, while maintaining it in the solid form, with molten magnesium which takes up the plutonium and separating the molten magnesium from the solid uranium.

  17. SOLVENT EXTRACTION PROCESS FOR PLUTONIUM

    DOEpatents

    Anderson, H.H.; Asprey, L.B.

    1960-02-01

    A process of separating plutonium in at least the tetravalent state from fission products contained in an aqueous acidic solution by extraction with alkyl phosphate is reported. The plutonium can then be back-extracted from the organic phase by contact with an aqueous solution of sulfuric, phosphoric, or oxalic acid as a complexing agent.

  18. PLUTONIUM-CERIUM-COPPER ALLOYS

    DOEpatents

    Coffinberry, A.S.

    1959-05-12

    A low melting point plutonium alloy useful as fuel is a homogeneous liquid metal fueled nuclear reactor is described. Vessels of tungsten or tantalum are useful to contain the alloy which consists essentially of from 10 to 30 atomic per cent copper and the balance plutonium and cerium. with the plutontum not in excess of 50 atomic per cent.

  19. Nature of foulants and fouling mechanism in the Protein A MabSelect resin cycled in a monoclonal antibody purification process.

    PubMed

    Zhang, Shaojie; Daniels, William; Salm, Jeffrey; Glynn, Judy; Martin, Joseph; Gallo, Christopher; Godavarti, Ranga; Carta, Giorgio

    2016-01-01

    The composition and origin of foulants and their spatial distribution within the particles of the Protein A MabSelect resin cycled in a mAb purification process are determined using electron and confocal microscopy techniques with gold and fluorescently labeled protein probes that associate with the foulants. The results show that the foulants are primarily related to the mAb product, are heterogeneously dispersed both on the outer surface and in the interior of the resin beads, and accumulate only when loading the conditioned CHO cell culture supernatant. Insignificant accumulation is seen if the process is run with purified mAb or with the null cell culture supernatant. When bound to the Protein A ligand, the mAb responsible for the observed fouling behavior is shown to associate with BSA and α-lactalbumin. This property is exploited using labeled versions of these lipophilic proteins to assess the effectiveness of improved resin cleaning processes and to elucidate the fouling mechanism. Resin fouling for this mAb appears to be consistent with the occurrence of conformational changes that occur upon binding, which, in turn, facilitate association of lipophilic proteins with the mAb. Upon desorption at low pH, these destabilized mAb complexes are deposited on and within the resin growing with each cycle and eventually leading to significant degradation of process performance.

  20. Plutonium utilisation in future UK PWRs

    SciTech Connect

    Thomas, G. M.; Worrall, A.

    2006-07-01

    Plutonium recycling in the form of Mixed Oxide (MOX) fuels is already a reality in over 30 reactors in Europe (in Belgium, Switzerland, Germany and France). Japan also plans to use MOX in approximately 30% of its reactors in the near future[1]. This paper describes potential near to mid-term disposition strategies for the United Kingdom's stockpile of plutonium. In order to be confident that MOX fuel can be utilised effectively in Pressurised Water Reactors (PWRs) in the UK, details are given of studies carried out recently at Nexia Solutions on PWR cores loaded with MOX containing typical UK plutonium isotopic compositions. Three dimensional steady state neutronic models of a standard Westinghouse four loop PWR design are constructed using state of the art tools (Studsvik of America's Core Management System[2, 3, 4]). Initially, a standard 18-month equilibrium UO{sub 2} fuel cycle is generated, followed by safety analyses and fuel performance calculations to demonstrate its feasibility. This equilibrium UO{sub 2} core is then gradually transitioned through loading patterns containing increasing MOX core loading fractions. Finally, an equilibrium MOX core loading pattern is determined. Technical safety analyses are also carried out on the transition cores and the final equilibrium scenario to ensure that all of the MOX cores are robust from a technical and safety viewpoint. Once these studies are completed the annual fuel throughputs for each scenario can be determined and used to produce options for managing the UK's plutonium stockpile. This work is part of a wider exercise currently being carried out by Nexia Solutions to explore the options for the safe disposition of the UK civil stockpile of separated PUO{sub 2}. (authors)

  1. Plutonium Multiple Recycling In PWRs

    SciTech Connect

    Nigon, Jean-Louis; Lenain, Richard; Zaetta, Alain

    2002-07-01

    Reprocessing and recycling open the road to a sustainable management of nuclear materials and an environment friendly management of nuclear waste. However, long or very long term recycling implies fast neutron reactors. High burn-ups of irradiated standard UO{sub 2} fuel as well as recycling of plutonium fuel in thermal reactors lead to a 'degradation' of plutonium that means a low fissile content, which is hardly compatible with recycling in LWRs. Thus the question of plutonium management has been raised; although there are some limitations, a truly large variety of options do exist; no one of the presently selected ways of plutonium management is a dead end road. Among these various options, some are fully compatible with the existing reactors and may be considered for the mid term future; they offer a competitive management of plutonium during the transition from thermal to fast reactors. (authors)

  2. Probing Phonons in Plutonium

    NASA Astrophysics Data System (ADS)

    Wong, Joe

    2004-03-01

    The phonon spectra of plutonium and its alloys have been sought after in the past few decades following the discovery of this actinide element in 1941, but with no success. This was due to a combination of the high neutron absorption cross section of 239Pu, the common isotope, and non-availability of large single crystals of any Pu-bearing materials. We have recent designed a high resolution inelastic x-ray scattering experiment using a bright synchrotron x-ray beam at the European Sychrotron Radiation Facility (ESRF), Grenoble and mapped the full phonon dispersion curves of an fcc delta-phase polycrystalline Pu-Ga alloy (1). Several unusual features including, a large elastic anisotropy, a small shear elastic modulus C', a Kohn-like anomaly in the T1[011] branch, and a pronounced softening of the [111] transverse modes are found. These features can be related to the phase transitions of plutonium and to strong coupling between the lattice structure and the 5f valence instabilities. Our results also provide a critical test for theoretical treatments of highly correlated 5f electron systems as exemplified by recent dynamical mean field theory (DMFT) calculations for d-plutonium.(2) This work was performed in collaboration with Dr. M. Krisch (ESRF)) and Prof. T.-C. Chiang (UIU), and under the auspices of the U. S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under Contract No. W-7405-Eng-48. 1. Joe Wong et al. Science, vol.301, 1078 (2003) 2. X. Dai et al. Science, vol.300, 953 (2003)

  3. PLUTONIUM-HYDROGEN REACTION PRODUCT, METHOD OF PREPARING SAME AND PLUTONIUM POWDER THEREFROM

    DOEpatents

    Fried, S.; Baumbach, H.L.

    1959-12-01

    A process is described for forming plutonlum hydride powder by reacting hydrogen with massive plutonium metal at room temperature and the product obtained. The plutonium hydride powder can be converted to plutonium powder by heating to above 200 deg C.

  4. Radiochemistry of uranium, neptunium and plutonium: an updating

    SciTech Connect

    Roberts, R.A.; Choppin, G.R.; Wild, J.F.

    1986-02-01

    This report presents some procedures used in the radiochemical isolation, purification and/or analysis of uranium, neptunium, and plutonium. In this update of the procedures, we have not attempted to discuss the developments in the chemistry of U, Np, and Pu but have restricted the report to the newer procedures, most of which have resulted from the increased emphasis in environmental concern which requires analysis of extremely small amounts of the actinide element in quite complex matrices. The final section of this report describes several schemes for isolation of actinides by oxidation state.

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

    SciTech Connect

    Lumetta, Gregg J.; Meier, David E.; Tingey, Joel M.; Casella, Amanda J.; Delegard, Calvin H.; Edwards, Matthew K.; Orton, Robert D.; Rapko, Brian M.; Smart, John E.

    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.

  6. Low temperature oxidation of plutonium

    SciTech Connect

    Nelson, Art J.; Roussel, Paul

    2013-05-15

    The initial oxidation of gallium stabilized {delta}-plutonium metal at 193 K has been followed using x-ray photoelectron spectroscopy. On exposure to Langmuir quantities of oxygen, plutonium rapidly forms a trivalent oxide followed by a tetravalent plutonium oxide. The growth modes of both oxides have been determined. Warming the sample in vacuum, the tetravalent oxide reduces to the trivalent oxide. The kinetics of this reduction reaction have followed and the activation energy has been determined to be 38.8 kJ mol{sup -1}.

  7. SOLVENT EXTRACTION PROCESS FOR PLUTONIUM

    DOEpatents

    Seaborg, G.T.

    1959-04-14

    The separation of plutonium from aqueous inorganic acid solutions by the use of a water immiscible organic extractant liquid is described. The plutonium must be in the oxidized state, and the solvents covered by the patent include nitromethane, nitroethane, nitropropane, and nitrobenzene. The use of a salting out agents such as ammonium nitrate in the case of an aqueous nitric acid solution is advantageous. After contacting the aqueous solution with the organic extractant, the resulting extract and raffinate phases are separated. The plutonium may be recovered by any suitable method.

  8. Probing phonons in plutonium

    SciTech Connect

    Wong, Joe; Krisch, M.; Farber, D.; Occelli, F.; Schwartz, A.; Chiang, T.C.; Wall, M.; Boro, C.; Xu, Ruqing

    2010-11-16

    Plutonium (Pu) is well known to have complex and unique physico-chemical properties. Notably, the pure metal exhibits six solid-state phase transformations with large volume expansions and contractions along the way to the liquid state: {alpha} {yields} {beta} {yields} {gamma} {yields} {delta} {yields} {delta}{prime} {yields} {var_epsilon} {yields} liquid. Unalloyed Pu melts at a relatively low temperature {approx}640 C to yield a higher density liquid than that of the solid from which it melts, (Figure 1). Detailed understanding of the properties of plutonium and plutonium-based alloys is critical for the safe handling, utilization, and long-term storage of these important, but highly toxic materials. However, both technical and and safety issues have made experimental observations extremely difficult. Phonon dispersion curves (PDCs) are key experimenta l data to the understanding of the basic properties of Pu materials such as: force constants, sound velocities, elastic constants, thermodynamics, phase stability, electron-phonon coupling, structural relaxation, etc. However, phonon dispersion curves (PDCs) in plutonium (Pu) and its alloys have defied measurement for the past few decades since the discovery of this element in 1941. This is due to a combination of the high thermal-neutron absorption cross section of plutonium and the inability to grow the large single crystals (with dimensions of a few millimeters) necessary for inelastic neutron scattering. Theoretical simulations of the Pu PDC continue to be hampered by the lack of suitable inter -atomic potentials. Thus, until recently the PDCs for Pu and its alloys have remained unknown experimentally and theoretically. The experimental limitations have recently been overcome by using a tightly focused undulator x-ray micro-beam scattered from single -grain domains in polycrystalline specimens. This experimental approach has been applied successfully to map the complete PDCs of an fcc d-Pu-Ga alloy using the

  9. Measurement of impurities in plutonium metal by rapid ion exchange/direct current argon plasma spectrometry

    SciTech Connect

    Maxwell, S.L. III; Coleman, J.T.

    1989-01-01

    A rapid ion exchange/direct current argon plasma (DCAP) spectrometry method is now being applied at the Savannah River Site to provide faster, more reliable assay of key metallic impurities in plutonium metal. These measurements are essential for nuclear materials accountability and enhanced process control. Impurity assays must be performed to ensure that plutonium product specifications are met and to determine the 100% -- impurities plutonium assay used in shipper/receiver calculations. Separation of impurities from plutonium metal is required prior to measurement by spectral techniques since the complex emission spectra of plutonium interferes with the impurity emission lines. A modified commercial vacuum system is used to perform the ion exchange separation in a glovebox. Since column flow rates are 10--15 times that of conventional ion exchange, purification time is relatively short. Separation efficiency is maintained by using small particle resin. The DCAP method is faster and provides much better accuracy and precision than the previously used carrier distillation dc arc spectrographic technique. The DCAP instrument has a much greater linear dynamic range than dc arc, does not require plutonium matrix standards for instrument calibration, and requires much less space than a dc arc spectrograph. Sixteen key metallic impurities are routinely measured using the ion exchange/DCAP spectrometry method. 11 refs., 6 figs., 1 tab.

  10. Options for converting excess plutonium to feed for the MOX fuel fabrication facility

    SciTech Connect

    Watts, Joe A; Smith, Paul H; Psaras, John D; Jarvinen, Gordon D; Costa, David A; Joyce, Jr., Edward L

    2009-01-01

    The storage and safekeeping of excess plutonium in the United States represents a multibillion-dollar lifecycle cost to the taxpayers and poses challenges to National Security and Nuclear Non-Proliferation. Los Alamos National Laboratory is considering options for converting some portion of the 13 metric tons of excess plutonium that was previously destined for long-term waste disposition into feed for the MOX Fuel Fabrication Facility (MFFF). This approach could reduce storage costs and security ri sks, and produce fuel for nuclear energy at the same time. Over the course of 30 years of weapons related plutonium production, Los Alamos has developed a number of flow sheets aimed at separation and purification of plutonium. Flow sheets for converting metal to oxide and for removing chloride and fluoride from plutonium residues have been developed and withstood the test oftime. This presentation will address some potential options for utilizing processes and infrastructure developed by Defense Programs to transform a large variety of highly impure plutonium into feedstock for the MFFF.

  11. Climate regulation, energy provisioning and water purification: Quantifying ecosystem service delivery of bioenergy willow grown on riparian buffer zones using life cycle assessment.

    PubMed

    Styles, David; Börjesson, Pål; D'Hertefeldt, Tina; Birkhofer, Klaus; Dauber, Jens; Adams, Paul; Patil, Sopan; Pagella, Tim; Pettersson, Lars B; Peck, Philip; Vaneeckhaute, Céline; Rosenqvist, Håkan

    2016-12-01

    Whilst life cycle assessment (LCA) boundaries are expanded to account for negative indirect consequences of bioenergy such as indirect land use change (ILUC), ecosystem services such as water purification sometimes delivered by perennial bioenergy crops are typically neglected in LCA studies. Consequential LCA was applied to evaluate the significance of nutrient interception and retention on the environmental balance of unfertilised energy willow planted on 50-m riparian buffer strips and drainage filtration zones in the Skåne region of Sweden. Excluding possible ILUC effects and considering oil heat substitution, strategically planted filter willow can achieve net global warming potential (GWP) and eutrophication potential (EP) savings of up to 11.9 Mg CO2e and 47 kg PO4e ha(-1) year(-1), respectively, compared with a GWP saving of 14.8 Mg CO2e ha(-1) year(-1) and an EP increase of 7 kg PO4e ha(-1) year(-1) for fertilised willow. Planting willow on appropriate buffer and filter zones throughout Skåne could avoid 626 Mg year(-1) PO4e nutrient loading to waters.

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

    NASA Astrophysics Data System (ADS)

    Zhang, Yingjie; Gregg, Daniel J.; Kong, Linggen; Jovanovich, Miodrag; Triani, Gerry

    2017-07-01

    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 CaZrTi2O7 zirconolite with the Ca-site cation vacancies and the incorporation of Al3+ 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.

  13. TERNARY ALLOY-CONTAINING PLUTONIUM

    DOEpatents

    Waber, J.T.

    1960-02-23

    Ternary alloys of uranium and plutonium containing as the third element either molybdenum or zirconium are reported. Such alloys are particularly useful as reactor fuels in fast breeder reactors. The alloy contains from 2 to 25 at.% of molybdenum or zirconium, the balance being a combination of uranium and plutonium in the ratio of from 1 to 9 atoms of uranlum for each atom of plutonium. These alloys are prepared by melting the constituent elements, treating them at an elevated temperature for homogenization, and cooling them to room temperature, the rate of cooling varying with the oomposition and the desired phase structure. The preferred embodiment contains 12 to 25 at.% of molybdenum and is treated by quenching to obtain a body centered cubic crystal structure. The most important advantage of these alloys over prior binary alloys of both plutonium and uranium is the lack of cracking during casting and their ready machinability.

  14. METHOD OF PREPARING PLUTONIUM TETRAFLUORIDE

    DOEpatents

    Beede, R.L.; Hopkins, H.H. Jr.

    1959-11-17

    C rystalline plutonium tetrafluoride is precipitated from aqueous up to 1.6 N mineral acid solutions of a plutorium (IV) salt with fluosilicic acid anions, preferably at room temperature. Hydrogen fluoride naay be added after precipitation to convert any plutonium fluosilicate to the tetrafluoride and any silica to fluosilicic acid. This process results in a purer product, especially as to iron and aluminum, than does the precipitation by the addition of hydrogen fluoride.

  15. IODATE METHOD FOR PURIFYING PLUTONIUM

    DOEpatents

    Stoughton, R.W.; Duffield, R.B.

    1958-10-14

    A method is presented for removing radioactive fission products from aqueous solutions containing such fission products together with plutonium. This is accomplished by incorporating into such solutions a metal iodate precipitate to remove fission products which form insoluble iodates. Suitable metal iodates are those of thorium and cerium. The plutonium must be in the hexavalent state and the pH of the solution must be manintained at less than 2.

  16. Testing New Inert Matrix and Thoria Fuels for Plutonium Incineration

    SciTech Connect

    Vettraino, F.; Padovan, E.; Tverberg, T.

    2002-07-01

    One major issue for nuclear power continues to be the public concern about rad-waste and proliferation risk induced by large plutonium stockpiles accumulated worldwide. In this context, nuclear fuels which exhibit no-plutonium production, and possibly allow for an efficient utilization of the plutonium to get rid of, are of great interest. This is the basic reason for the efforts that many international institutions are devoting to R and D on such new U-free fuel concepts as Inert Matrix (IMF) and Thorium fuels. At the moment the major merit of such innovative fuels is primarily related to the safe closure of the nuclear fuel cycle as especially expected from those new concepts like ADS (Accelerated Driven System) for the transmutation of plutonium, minor actinides and LLFP. Both ceramic inert matrix (IM) and thoria (T) fuels have been identified as suitable to the scope of burning weapon and civilian plutonium and to act also as possible carrier for transmutation of minor actinides. For testing the irradiation behaviour of these new materials, three kinds of fuels have been selected: inert matrix (IM) fuel, inert matrix thoria-doped (IMT) fuel, and thoria (T) fuel. A first experiment, IFA-652, 40 MWD/kg burnup target, including high enriched uranium (HEU) as fissile phase, instead of plutonium, is currently underway in the Halden HWBR. The reason for this choice was that manufacturing of Pu containing fuels is more complex and there was no fabrication facility available at the needed time for the Pu fuel. It is expected, however, that the relative behaviour of the different kind of matrices would be only slightly dependent on the adopted fissile material. So, the comparison of the in-pile performance of the three fuels will constitute a significant common database also for plutonium bearing fuels. The primary aim for the IFA-652 experiment is the measurement of basic characteristics under LWR irradiation conditions over a period of 4-5 years. The design of a

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

    SciTech Connect

    DeMuth, S.F.

    1997-09-30

    The disposition of plutonium from decommissioned nuclear weapons, by incorporation into commercial UO{sub 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{sub 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.

  18. Capability to Recover Plutonium-238 in H-Canyon/HB-Line - 13248

    SciTech Connect

    Fuller, Kenneth S. Jr.; Smith, Robert H. Jr.; Goergen, Charles R.

    2013-07-01

    Plutonium-238 is used in Radioisotope Thermoelectric Generators (RTGs) to generate electrical power and in Radioisotope Heater Units (RHUs) to produce heat for electronics and environmental control for deep space missions. The domestic supply of Pu-238 consists of scrap material from previous mission production or material purchased from Russia. Currently, the United States has no significant production scale operational capability to produce and separate new Pu-238 from irradiated neptunium-237 targets. The Department of Energy - Nuclear Energy is currently evaluating and developing plans to reconstitute the United States capability to produce Pu-238 from irradiated Np-237 targets. The Savannah River Site had previously produced and/or processed all the Pu-238 utilized in Radioisotope Thermoelectric Generators (RTGs) for deep space missions up to and including the majority of the plutonium for the Cassini Mission. The previous full production cycle capabilities included: Np- 237 target fabrication, target irradiation, target dissolution and Np-237 and Pu-238 separation and purification, conversion of Np-237 and Pu-238 to oxide, scrap recovery, and Pu-238 encapsulation. The capability and equipment still exist and could be revitalized or put back into service to recover and purify Pu-238/Np-237 or broken General Purpose Heat Source (GPHS) pellets utilizing existing process equipment in HB-Line Scrap Recovery, and H-Canyon Frame Waste Recovery processes. The conversion of Np-237 and Pu-238 to oxide can be performed in the existing HB-Line Phase-2 and Phase- 3 Processes. Dissolution of irradiated Np-237 target material, and separation and purification of Np-237 and Pu-238 product streams would be possible at production rates of ∼2 kg/month of Pu-238 if the existing H-Canyon Frames Process spare equipment were re-installed. Previously, the primary H-Canyon Frames equipment was removed to be replaced: however, the replacement project was stopped. The spare equipment

  19. CAPABILITY TO RECOVER PLUTONIUM-238 IN H-CANYON/HB-LINE

    SciTech Connect

    Fuller, Kenneth S. Jr.; Smith, Robert H. Jr.; Goergen, Charles R.

    2013-01-09

    Plutonium-238 is used in Radioisotope Thermoelectric Generators (RTGs) to generate electrical power and in Radioisotope Heater Units (RHUs) to produce heat for electronics and environmental control for deep space missions. The domestic supply of Pu-238 consists of scrap material from previous mission production or material purchased from Russia. Currently, the United States has no significant production scale operational capability to produce and separate new Pu-238 from irradiated neptunium-237 targets. The Department of Energy - Nuclear Energy is currently evaluating and developing plans to reconstitute the United States capability to produce Pu-238 from irradiated Np-237 targets. The Savannah River Site had previously produced and/or processed all the Pu-238 utilized in Radioisotope Thermoelectric Generators (RTGs) for deep space missions up to and including the majority of the plutonium for the Cassini Mission. The previous full production cycle capabilities included: Np-237 target fabrication, target irradiation, target dissolution and Np-237 and Pu-238 separation and purification, conversion of Np-237 and Pu-238 to oxide, scrap recovery, and Pu-238 encapsulation. The capability and equipment still exist and could be revitalized or put back into service to recover and purify Pu-238/Np-237 or broken General Purpose Heat Source (GPHS) pellets utilizing existing process equipment in HB-Line Scrap Recovery, and H-anyon Frame Waste Recovery processes. The conversion of Np-237 and Pu-238 to oxide can be performed in the existing HB-Line Phase-2 and Phase-3 Processes. Dissolution of irradiated Np-237 target material, and separation and purification of Np-237 and Pu-238 product streams would be possible at production rates of ~ 2 kg/month of Pu-238 if the existing H-Canyon Frames Process spare equipment were re-installed. Previously, the primary H-Canyon Frames equipment was removed to be replaced: however, the replacement project was stopped. The spare equipment is

  20. DOE plutonium disposition study: Pu consumption in ALWRs. Volume 2, Final report

    SciTech Connect

    Not Available

    1993-05-15

    The Department of Energy (DOE) has contracted with Asea Brown Boveri-Combustion Engineering (ABB-CE) to provide information on the capability of ABB-CE`s System 80 + Advanced Light Water Reactor (ALWR) to transform, through reactor burnup, 100 metric tonnes (MT) of weapons grade plutonium (Pu) into a form which is not readily useable in weapons. This information is being developed as part of DOE`s Plutonium Disposition Study, initiated by DOE in response to Congressional action. This document Volume 2, provides a discussion of: Plutonium Fuel Cycle; Technology Needs; Regulatory Considerations; Cost and Schedule Estimates; and Deployment Strategy.

  1. Determination of origin and intended use of plutonium metal using nuclear forensic techniques.

    PubMed

    Rim, Jung H; Kuhn, Kevin J; Tandon, Lav; Xu, Ning; Porterfield, Donivan R; Worley, Christopher G; Thomas, Mariam R; Spencer, Khalil J; Stanley, Floyd E; Lujan, Elmer J; Garduno, Katherine; Trellue, Holly R

    2017-04-01

    Nuclear forensics techniques, including micro-XRF, gamma spectrometry, trace elemental analysis and isotopic/chronometric characterization were used to interrogate two, potentially related plutonium metal foils. These samples were submitted for analysis with only limited production information, and a comprehensive suite of forensic analyses were performed. Resulting analytical data was paired with available reactor model and historical information to provide insight into the materials' properties, origins, and likely intended uses. Both were super-grade plutonium, containing less than 3% (240)Pu, and age-dating suggested that most recent chemical purification occurred in 1948 and 1955 for the respective metals. Additional consideration of reactor modeling feedback and trace elemental observables indicate plausible U.S. reactor origin associated with the Hanford site production efforts. Based on this investigation, the most likely intended use for these plutonium foils was (239)Pu fission foil targets for physics experiments, such as cross-section measurements, etc.

  2. Electrorefining of cerium: Part 2, Cerium as a surrogate for plutonium electrorefining studies

    SciTech Connect

    Raraz, A.G.; Mishra, B.; Olson, D.L.; Moore, J.J.

    1992-01-01

    The plutonium metal produced by the Direct Oxide Reduction Process is associated with other metallic impurities that have to be removed. The purification of plutonium is achieved using electrorefining process through a molten salt medium. The optimization of process parameters involved in electrorefining is required to make the process effective, in terms of the metal purity, cell efficiency and overall process reliability. Since the study of strategic and radioactive metals requires the use of a surrogate, it is important to choose surrogates that simulate the process as closely as possible. Cerium has been chosen to study the electrorefining behavior of plutonium. The differences that exist in the physico-chemical properties between the two metals have been critically examined and appropriate models have been developed to study the behavior. Cerium is a justified choice for the investigation.

  3. Separation of Plutonium from Irradiated Fuels and Targets

    SciTech Connect

    Gray, Leonard W.; Holliday, Kiel S.; Murray, Alice; Thompson, Major; Thorp, Donald T.; Yarbro, Stephen; Venetz, Theodore J.

    2015-09-30

    Spent nuclear fuel from power production reactors contains moderate amounts of transuranium (TRU) actinides and fission products in addition to the still slightly enriched uranium. Originally, nuclear technology was developed to chemically separate and recover fissionable plutonium from irradiated nuclear fuel for military purposes. Military plutonium separations had essentially ceased by the mid-1990s. Reprocessing, however, can serve multiple purposes, and the relative importance has changed over time. In the 1960’s the vision of the introduction of plutonium-fueled fast-neutron breeder reactors drove the civilian separation of plutonium. More recently, reprocessing has been regarded as a means to facilitate the disposal of high-level nuclear waste, and thus requires development of radically different technical approaches. In the last decade or so, the principal reason for reprocessing has shifted to spent power reactor fuel being reprocessed (1) so that unused uranium and plutonium being recycled reduce the volume, gaining some 25% to 30% more energy from the original uranium in the process and thus contributing to energy security and (2) to reduce the volume and radioactivity of the waste by recovering all long-lived actinides and fission products followed by recycling them in fast reactors where they are transmuted to short-lived fission products; this reduces the volume to about 20%, reduces the long-term radioactivity level in the high-level waste, and complicates the possibility of the plutonium being diverted from civil use – thereby increasing the proliferation resistance of the fuel cycle. In general, reprocessing schemes can be divided into two large categories: aqueous/hydrometallurgical systems, and pyrochemical/pyrometallurgical systems. Worldwide processing schemes are dominated by the aqueous (hydrometallurgical) systems. This document provides a historical review of both categories of reprocessing.

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

    SciTech Connect

    Dvoeglazov, K.; Volk, V.; Zverev, D.; Veselov, S.; Krivitskiy, Y.; Alekseenko, S.; Alekseenko, V.

    2013-07-01

    For the purpose of removing plutonium from uranium- plutonium extract it is proposed to employ concentration stripping process with the use of separator and a new reducing reagent: Carbohydrazide CO(N{sub 2}H{sub 3}){sub 2}. Using plutonium stripping from solution simulating the composition of extract of spent nuclear fuel from VVER-1000 reactor (without γ-emitting isotopes), with O: A ratio of = 28, a product solution was obtained containing 17.8 g/l of plutonium, 29.2 g/l of uranium and more than 1 g/l of technetium. The experiment on real spent fuel from VVER-1000 with burn-up of more than 50 GW*d/t of uranium after 17 year exposure, performed in the shielded box of FSUE 'MCP', confirmed the effectiveness and feasibility of the proposed process. Through concentration stripping (O:A = 20), a plutonium product solution was obtained with a part of uranium with the following composition: [U] = 150 g/l; [Pu] = 23,5 g/l; [Np] = 1,7 g/l, [Tc] = 1.5 g/l; gamma exposure rate - 0,022 mR/s*l. Direct extraction of plutonium in this operation was 95.3%, the rest of plutonium is refluxing to the preceding stage of the extraction cycle. A process flow diagram with organization of plutonium recycling is proposed, allowing for its complete removal into a single stream. Carbohydrazide is an effective reducing agent of plutonium (IV), ensuring the stability of uranium-plutonium separation process. (authors)

  5. Spectrophotometric determination of plutonium with chlorophosphonazo III in n-pentanol

    SciTech Connect

    Saponara, N.M.; Marsh, S.F.

    1982-03-01

    Microgram amounts of plutonium are measured spectrophotometrically as the plutonium-chlorophosphonazo III complex after extraction into n-pentanol from 1.5 M HCl. The relative standard deviation is 1.5% for the range of 2.5 to 17.5 ..mu..g. The tolerance is excellent for many metals and nonmetals present in nuclear fuel-cycle materials. A preceding anion-exchange-column separation increases tolerance for certain metals and nonmetals.

  6. Plutonium and americium separation from salts

    DOEpatents

    Hagan, Paul G.; Miner, Frend 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.

  7. Plutonium focus area

    SciTech Connect

    1996-08-01

    To ensure research and development programs focus on the most pressing environmental restoration and waste management problems at the U.S. Department of Energy (DOE), the Assistant Secretary for the Office of Environmental Management (EM) established a working group in August 1993 to implement a new approach to research and technology development. As part of this new approach, EM developed a management structure and principles that led to the creation of specific Focus Areas. These organizations were designed to focus the scientific and technical talent throughout DOE and the national scientific community on the major environmental restoration and waste management problems facing DOE. The Focus Area approach provides the framework for intersite cooperation and leveraging of resources on common problems. After the original establishment of five major Focus Areas within the Office of Technology Development (EM-50, now called the Office of Science and Technology), the Nuclear Materials Stabilization Task Group (EM-66) followed the structure already in place in EM-50 and chartered the Plutonium Focus Area (PFA). The following information outlines the scope and mission of the EM, EM-60, and EM-66 organizations as related to the PFA organizational structure.

  8. Plutonium solution analyzer

    SciTech Connect

    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).

  9. Report by a special panel of the American Nuclear Society: Protection and management of plutonium

    SciTech Connect

    Bengelsdorf, H.

    1996-07-01

    The American Nuclear Society (ANS) established an independent and prestigious panel several months ago to take the matter up where the US National Academy of Science (NAS) left off. The challenge was to look at the broader issue of what to do with civil plutonium, as well as excess weapons material. In terms of approach, the report focused on several short- and long-term issues. The short-term focus was on the disposition of excess weapons plutonium, while the longer-range issue concerned the disposition of the plutonium being produced in the civil nuclear fuel cycle. For the short term, the ANS panel strongly endorsed the concept that all plutonium scheduled for release from the US and Russian weapons stocks should be converted to a form that is intensively radioactive in order to protect the plutonium from theft of seizure (the spent fuel standard). However, since the conversion will at best take several years to complete, the panel has concluded that immediate emphasis should be placed on the assurance that all unconverted materials are protected as securely as when they were part of the active weapon stockpiles. More importantly, the panel also recommended prompt implementation of the so-called reactor option for disposing of surplus US and Russian weapons plutonium. The longer-term issues covered by the panel were those posed by the growing stocks of both separated plutonium and spent fuel generated in the world`s civil nuclear power programs. These issues included what fuel cycle policies should be prudently pursued in light of proliferation risks and likely future energy needs, what steps should be taken in regard to the increase in the demand for nuclear power in the future, and how civil plutonium in its various forms should be protected and managed to minimize proliferation. Overall, the panel concluded that plutonium is an energy resource that should be used and not a waste material to be disposed of.

  10. PROCESS OF SEPARATING PLUTONIUM FROM URANIUM

    DOEpatents

    Brown, H.S.; Hill, O.F.

    1958-09-01

    A process is presented for recovering plutonium values from aqueous solutions. It comprises forming a uranous hydroxide precipitate in such a plutonium bearing solution, at a pH of at least 5. The plutonium values are precipitated with and carried by the uranium hydroxide. The carrier precipitate is then redissolved in acid solution and the pH is adjusted to about 2.5, causing precipitation of the uranous hydroxide but leaving the still soluble plutonium values in solution.

  11. Plutonium Proliferation: The Achilles Heel of Disarmament

    SciTech Connect

    Leventhal, Paul

    2001-02-07

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

  12. Selecting a plutonium vitrification process

    SciTech Connect

    Jouan, A.

    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.

  13. Tags to Track Illicit Uranium and Plutonium

    SciTech Connect

    Haire, M. Jonathan; Forsberg, Charles W.

    2007-07-01

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

  14. 49 CFR 175.704 - Plutonium shipments.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 2 2013-10-01 2013-10-01 false Plutonium shipments. 175.704 Section 175.704... Regulations Applicable According to Classification of Material § 175.704 Plutonium shipments. Shipments of plutonium which are subject to 10 CFR 71.88(a)(4) must comply with the following: (a) Each...

  15. 49 CFR 175.704 - Plutonium shipments.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 2 2014-10-01 2014-10-01 false Plutonium shipments. 175.704 Section 175.704... Regulations Applicable According to Classification of Material § 175.704 Plutonium shipments. Shipments of plutonium which are subject to 10 CFR 71.88(a)(4) must comply with the following: (a) Each...

  16. 49 CFR 175.704 - Plutonium shipments.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Plutonium shipments. 175.704 Section 175.704... Regulations Applicable According to Classification of Material § 175.704 Plutonium shipments. Shipments of plutonium which are subject to 10 CFR 71.88(a)(4) must comply with the following: (a) Each...

  17. 49 CFR 175.704 - Plutonium shipments.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 2 2011-10-01 2011-10-01 false Plutonium shipments. 175.704 Section 175.704... Regulations Applicable According to Classification of Material § 175.704 Plutonium shipments. Shipments of plutonium which are subject to 10 CFR 71.88(a)(4) must comply with the following: (a) Each...

  18. Rapid Radiochemical Method for Plutonium-238 and ...

    EPA Pesticide Factsheets

    Technical Fact Sheet Technique: Alpha spectrometry Method Developed for: Plutonium-238 and plutonium-239 in building materials Method Selected for: SAM lists this method for qualitative analysis of plutonium-238 and -239 in concrete or brick building materials. Summary of subject analytical method which will be posted to the SAM website to allow access to the method.

  19. 49 CFR 175.704 - Plutonium shipments.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 49 Transportation 2 2012-10-01 2012-10-01 false Plutonium shipments. 175.704 Section 175.704... Regulations Applicable According to Classification of Material § 175.704 Plutonium shipments. Shipments of plutonium which are subject to 10 CFR 71.88(a)(4) must comply with the following: (a) Each...

  20. Plutonium Oxide Process Capability Work Plan

    SciTech Connect

    Meier, David E.; Tingey, Joel M.

    2014-02-28

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

  1. Plutonium stabilization and packaging system

    SciTech Connect

    1996-05-01

    This document describes the functional design of the Plutonium Stabilization and Packaging System (Pu SPS). The objective of this system is to stabilize and package plutonium metals and oxides of greater than 50% wt, as well as other selected isotopes, in accordance with the requirements of the DOE standard for safe storage of these materials for 50 years. This system will support completion of stabilization and packaging campaigns of the inventory at a number of affected sites before the year 2002. The package will be standard for all sites and will provide a minimum of two uncontaminated, organics free confinement barriers for the packaged material.

  2. Anthropogenic plutonium-244 in the environment: Insights into plutonium’s longest-lived isotope

    PubMed Central

    Armstrong, Christopher R.; Brant, Heather A.; Nuessle, Patterson R.; Hall, Gregory; Cadieux, James R.

    2016-01-01

    Owing to the rich history of heavy element production in the unique high flux reactors that operated at the Savannah River Site, USA (SRS) decades ago, trace quantities of plutonium with highly unique isotopic characteristics still persist today in the SRS terrestrial environment. Development of an effective sampling, processing, and analysis strategy enables detailed monitoring of the SRS environment, revealing plutonium isotopic compositions, e.g., 244Pu, that reflect the unique legacy of plutonium production at SRS. This work describes the first long-term investigation of anthropogenic 244Pu occurrence in the environment. Environmental samples, consisting of collected foot borne debris, were taken at SRS over an eleven year period, from 2003 to 2014. Separation and purification of trace plutonium was carried out followed by three stage thermal ionization mass spectrometry (3STIMS) measurements for plutonium isotopic content and isotopic ratios. Significant 244Pu was measured in all of the years sampled with the highest amount observed in 2003. The 244Pu content, in femtograms (fg = 10−15 g) per gram, ranged from 0.31 fg/g to 44 fg/g in years 2006 and 2003 respectively. In all years, the 244Pu/239Pu atom ratios were significantly higher than global fallout, ranging from 0.003 to 0.698 in years 2014 and 2003 respectively. PMID:26898531

  3. Anthropogenic plutonium-244 in the environment: Insights into plutonium’s longest-lived isotope

    NASA Astrophysics Data System (ADS)

    Armstrong, Christopher R.; Brant, Heather A.; Nuessle, Patterson R.; Hall, Gregory; Cadieux, James R.

    2016-02-01

    Owing to the rich history of heavy element production in the unique high flux reactors that operated at the Savannah River Site, USA (SRS) decades ago, trace quantities of plutonium with highly unique isotopic characteristics still persist today in the SRS terrestrial environment. Development of an effective sampling, processing, and analysis strategy enables detailed monitoring of the SRS environment, revealing plutonium isotopic compositions, e.g., 244Pu, that reflect the unique legacy of plutonium production at SRS. This work describes the first long-term investigation of anthropogenic 244Pu occurrence in the environment. Environmental samples, consisting of collected foot borne debris, were taken at SRS over an eleven year period, from 2003 to 2014. Separation and purification of trace plutonium was carried out followed by three stage thermal ionization mass spectrometry (3STIMS) measurements for plutonium isotopic content and isotopic ratios. Significant 244Pu was measured in all of the years sampled with the highest amount observed in 2003. The 244Pu content, in femtograms (fg = 10‑15 g) per gram, ranged from 0.31 fg/g to 44 fg/g in years 2006 and 2003 respectively. In all years, the 244Pu/239Pu atom ratios were significantly higher than global fallout, ranging from 0.003 to 0.698 in years 2014 and 2003 respectively.

  4. MA Doping Analysis on Breeding Capability and Protected Plutonium Production of Large FBR

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Suzuki, Mitsutoshi; Kuno, Yusuke

    2010-06-01

    breeding region for protected plutonium production. Breeding capability of the reactor can be increased effectively by increasing MA doping rate while criticality condition of the reactor is reduced by doping MA. Adopting MA cycle is also effective to increase the isotopic Pu-238 production in plutonium vector composition for denaturing purpose of plutonium.

  5. MA Doping Analysis on Breeding Capability and Protected Plutonium Production of Large FBR

    SciTech Connect

    Permana, Sidik; Suzuki, Mitsutoshi; Kuno, Yusuke

    2010-06-22

    breeding region for protected plutonium production. Breeding capability of the reactor can be increased effectively by increasing MA doping rate while criticality condition of the reactor is reduced by doping MA. Adopting MA cycle is also effective to increase the isotopic Pu-238 production in plutonium vector composition for denaturing purpose of plutonium.

  6. PROCESS FOR SEPARATING PLUTONIUM FROM IMPURITIES

    DOEpatents

    Wahl, A.C.

    1957-11-12

    A method is described for separating plutonium from aqueous solutions containing uranium. It has been found that if the plutonium is reduced to its 3+ valence state, and the uranium present is left in its higher valence state, then the differences in solubility between certain salts (e.g., oxalates) of the trivalent plutonium and the hexavalent uranium can be used to separate the metals. This selective reduction of plutonium is accomplished by adding iodide ion to the solution, since iodide possesses an oxidation potential sufficient to reduce plutonium but not sufficient to reduce uranium.

  7. Plutonium inventory characterization technical evaluation report

    SciTech Connect

    Wittman, G.R., Westinghouse Hanford

    1996-07-10

    This is a technical report on the data, gathered to date, under WHC- SD-CP-TP-086, Rev. 1, on the integrity of the food pack cans currently being used to store plutonium or plutonium compounds at the Plutonium Finishing Plant. Workplan PFP-96-VO-009, `Inspection of Special Nuclear Material Using X-ray`, was used to gather data on material and containment conditions using real time radiography. Some of those images are included herein. A matrix found in the `Plutonium Inventory Characterization Implementation Plan` was used to categorize different plutonium items based upon the type of material being stored and the life expectancy of the containers.

  8. Method of separating thorium from plutonium

    DOEpatents

    Clifton, D.G.; Blum, T.W.

    A method of chemically separating plutonium from thorium is claimed. Plutonium and thorium to be separated are dissolved in an aqueous feed solution, preferably as the nitrate salts. The feed solution is acidified and sodium nitrite is added to the solution to adjust the valence of the plutonium to the +4 state. A chloride salt, preferably sodium chloride, is then added to the solution to induce formation of an anionic plutonium chloride complex. The anionic plutonium chloride complex and the thorium in solution are then separated by ion exchange on a strong base anion exchange column.

  9. Plutonium immobilization feed batching system concept report

    SciTech Connect

    Erickson, S.

    2000-07-19

    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 high level waste glass for permanent storage. Feed batching is one of the first process steps involved with first stage plutonium immobilization. It will blend plutonium oxide powder before it is combined with other materials to make pucks. This report discusses the Plutonium Immobilization feed batching process preliminary concept, batch splitting concepts, and includes a process block diagram, concept descriptions, a preliminary equipment list, and feed batching development areas.

  10. Method of separating thorium from plutonium

    DOEpatents

    Clifton, D.G.; Blum, T.W.

    1984-07-10

    A method is described for chemically separating plutonium from thorium. Plutonium and thorium to be separated are dissolved in an aqueous feed solution, preferably as the nitrate salts. The feed solution is acidified and sodium nitrite is added to the solution to adjust the valence of the plutonium to the +4 state. A chloride salt, preferably sodium chloride, is then added to the solution to induce formation of an anionic plutonium chloride complex. The anionic plutonium chloride complex and the thorium in solution are then separated by ion exchange on a strong base anion exchange column.

  11. Method of separating thorium from plutonium

    DOEpatents

    Clifton, David G.; Blum, Thomas W.

    1984-01-01

    A method of chemically separating plutonium from thorium. Plutonium and thorium to be separated are dissolved in an aqueous feed solution, preferably as the nitrate salts. The feed solution is acidified and sodium nitrite is added to the solution to adjust the valence of the plutonium to the +4 state. A chloride salt, preferably sodium chloride, is then added to the solution to induce formation of an anionic plutonium chloride complex. The anionic plutonium chloride complex and the thorium in solution are then separated by ion exchange on a strong base anion exchange column.

  12. Plutonium inventories for stabilization and stabilized materials

    SciTech Connect

    Williams, A.K.

    1996-05-01

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

  13. The First Weighing of Plutonium

    DOE R&D Accomplishments Database

    Seaborg, Glenn T.

    1967-09-10

    Recollections and reminiscences at the 25th Anniversary of the First Weighing of Plutonium, Chicago, IL, September 10, 1967, tell an important part of the story of this fascinating new element that is destined to play an increasingly significant role in the future of man.

  14. Plutonium disposition and storage model

    SciTech Connect

    Krupa, J.F.

    2000-03-01

    An EXTEND/SDI-Industry model of DOE plutonium disposition and storage has been created which can easily accommodate changes in scenarios by changing input parameters. It matches well with hand-crafted spreadsheet analyses, and has the advantage that it shows system logic and can be documented easily.

  15. Plutonium disposition and storage model

    SciTech Connect

    Krupa, J.F.

    1999-12-06

    An EXTEND/SDI-Industry model of DOE plutonium disposition and storage has been created which can easily accommodate changes in scenarios by changing input parameters. It matches well with hand-crafted spreadsheet analyses, and has the advantage that it shows system logic and can be documented easily.

  16. Plutonium Recycle: The Fateful Step

    ERIC Educational Resources Information Center

    Speth, J. Gustave; And Others

    1974-01-01

    Calls attention to the fact that if the Atomic Energy Commission proceeds with its plans to authorize the nuclear power industry to use plutonium as a fuel in commercial nuclear reactors around the country, this will result in a dramatic escalation in the risks posed by nuclear power. (PEB)

  17. Plutonium Recycle: The Fateful Step

    ERIC Educational Resources Information Center

    Speth, J. Gustave; And Others

    1974-01-01

    Calls attention to the fact that if the Atomic Energy Commission proceeds with its plans to authorize the nuclear power industry to use plutonium as a fuel in commercial nuclear reactors around the country, this will result in a dramatic escalation in the risks posed by nuclear power. (PEB)

  18. Development program to recycle and purify plutonium-238 oxide fuel from scrap

    NASA Astrophysics Data System (ADS)

    Schulte, Louis D.; Silver, Gary L.; Avens, Larry R.; Jarvinen, Gordon D.; Espinoza, Jacob; Foltyn, Elizabeth M.; Rinehart, Gary H.

    1997-01-01

    Nuclear Materials Technology (NMT) Division of Los Alamos National Laboratory (LANL) has initiated a development program to recover & purify plutonium-238 oxide from impure sources. A glove box line has been designed and a process flowsheet developed to perform this task on a large scale. Our initial effort has focused on purification of 238PuO2 fuel that fails to meet General Purpose Heat Source (GPHS) specifications because of impurities. The most notable non-actinide impurity was silicon, but aluminum, chromium, iron and nickel were also near or in excess of limits specified by GPHS fuel powder specifications. 234U was by far the largest actinide impurity observed in the feed material because it is the daughter product of 238Pu by alpha decay. An aqueous method based on nitric acid was selected for purification of the 238PuO2 fuel. All aqueous processing used high purity reagents, and was performed in PTFE apparatus to minimize introduction of new contaminants. Impure 238PuO2 was finely milled, then dissolved in refluxing HNO3/HF and the solution filtered. The dissolved 238Pu was adjusted to the trivalent state by an excess of reducing reagents to compensate for radiolytic effects, precipitated as plutonium(III) oxalate, and recovered by filtration. The plutonium(III) oxalate was subsequently calcined to convert the plutonium to the oxide. Decontamination factors for silicon, phosphorus and uranium were excellent. Decontamination factors for aluminum, chromium, iron and nickel were very good. The purity of the 238PuO2 recovered from this operation was significantly better than specifications. Efforts continue to develop the capability for efficient, safe, cost-effective, and environmentally acceptable methods to recover and purify 238PuO2 fuel in a glove box environment. Plutonium-238 materials targeted for recovery includes impure oxide and scrap items that are lean in 238Pu values.

  19. Analysis on Reactor Criticality Condition and Fuel Conversion Capability Based on Different Loaded Plutonium Composition in FBR Core

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Saputra, Geby; Suzuki, Mitsutoshi; Saito, Masaki

    2017-01-01

    Reactor criticality condition and fuel conversion capability are depending on the fuel arrangement schemes, reactor core geometry and fuel burnup process as well as the effect of different fuel cycle and fuel composition. Criticality condition of reactor core and breeding ratio capability have been investigated in this present study based on fast breeder reactor (FBR) type for different loaded fuel compositions of plutonium in the fuel core regions. Loaded fuel of Plutonium compositions are based on spent nuclear fuel (SNF) of light water reactor (LWR) for different fuel burnup process and cooling time conditions of the reactors. Obtained results show that different initial fuels of plutonium gives a significant chance in criticality conditions and fuel conversion capability. Loaded plutonium based on higher burnup process gives a reduction value of criticality condition or less excess reactivity. It also obtains more fuel breeding ratio capability or more breeding gain. Some loaded plutonium based on longer cooling time of LWR gives less excess reactivity and in the same time, it gives higher breeding ratio capability of the reactors. More composition of even mass plutonium isotopes gives more absorption neutron which affects to decresing criticality or less excess reactivity in the core. Similar condition that more absorption neutron by fertile material or even mass plutonium will produce more fissile material or odd mass plutonium isotopes to increase the breeding gain of the reactor.

  20. Hamiltonian purification

    SciTech Connect

    Orsucci, Davide; Burgarth, Daniel; Facchi, Paolo; Pascazio, Saverio; Nakazato, Hiromichi; Yuasa, Kazuya; Giovannetti, Vittorio

    2015-12-15

    The problem of Hamiltonian purification introduced by Burgarth et al. [Nat. Commun. 5, 5173 (2014)] is formalized and discussed. Specifically, given a set of non-commuting Hamiltonians (h{sub 1}, …, h{sub m}) operating on a d-dimensional quantum system ℋ{sub d}, the problem consists in identifying a set of commuting Hamiltonians (H{sub 1}, …, H{sub m}) operating on a larger d{sub E}-dimensional system ℋ{sub d{sub E}} which embeds ℋ{sub d} as a proper subspace, such that h{sub j} = PH{sub j}P with P being the projection which allows one to recover ℋ{sub d} from ℋ{sub d{sub E}}. The notions of spanning-set purification and generator purification of an algebra are also introduced and optimal solutions for u(d) are provided.

  1. Accelerated purification of colloidal silica sols

    NASA Technical Reports Server (NTRS)

    Bahnsen, E. B.; Garofalini, S.; Pechman, A.

    1979-01-01

    Accelerated purification process for colloidal sols using heat/deionization scheme, sharply reduces waiting time between deionization cycles from several months to a few days. Process produces same high purity silica sols as conventional methods.

  2. Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

    SciTech Connect

    Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.; Baryshnikov, M.V.; Kryukov, O.V.; Khaperskaya, A.V.

    2013-07-01

    REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

  3. PRECIPITATION METHOD FOR THE SEPARATION OF PLUTONIUM AND RARE EARTHS

    DOEpatents

    Thompson, S.G.

    1960-04-26

    A method of purifying plutonium is given. Tetravalent plutonium is precipitated with thorium pyrophosphate, the plutonium is oxidized to the tetravalent state, and then impurities are precipitated with thorium pyrophosphate.

  4. Surprising coordination for plutonium in the first plutonium(III) borate.

    PubMed

    Wang, Shuao; Alekseev, Evgeny V; Depmeier, Wulf; Albrecht-Schmitt, Thomas E

    2011-03-21

    The first plutonium(III) borate, Pu(2)[B(12)O(18)(OH)(4)Br(2)(H(2)O)(3)]·0.5H(2)O, has been prepared by reacting plutonium(III) with molten boric acid under strictly anaerobic conditions. This compound contains a three-dimensional polyborate network with triangular holes that house the plutonium(III) sites. The plutonium sites in this compound are 9- and 10-coordinate and display atypical geometries.

  5. Air transport of plutonium metal: content expansion initiative for the plutonium air transportable (PAT01) packaging

    SciTech Connect

    Caviness, Michael L; Mann, Paul T

    2010-01-01

    The National Nuclear Security Administration (NNSA) has submitted an application to the Nuclear Regulatory Commission (NRC) for the air shipment of plutonium metal within the Plutonium Air Transportable (PAT-1) packaging. The PAT-1 packaging is currently authorized for the air transport of plutonium oxide in solid form only. The INMM presentation will provide a limited overview of the scope of the plutonium metal initiative and provide a status of the NNSA application to the NRC.

  6. SEPARATION OF PLUTONIUM HYDROXIDE FROM BISMUTH HYDROXIDE

    DOEpatents

    Watt, G.W.

    1958-08-19

    An tmproved method is described for separating plutonium hydroxide from bismuth hydroxide. The end product of the bismuth phosphate processes for the separation amd concentration of plutonium is a inixture of bismuth hydroxide amd plutonium hydroxide. It has been found that these compounds can be advantageously separated by treatment with a reducing agent having a potential sufficient to reduce bismuth hydroxide to metalltc bisinuth but not sufficient to reduce the plutonium present. The resulting mixture of metallic bismuth and plutonium hydroxide can then be separated by treatment with a material which will dissolve plutonium hydroxide but not metallic bismuth. Sodiunn stannite is mentioned as a preferred reducing agent, and dilute nitric acid may be used as the separatory solvent.

  7. Multi-generational stewardship of plutonium

    SciTech Connect

    Pillay, K.K.S.

    1997-10-01

    The post-cold war era has greatly enhanced the interest in the long-term stewardship of plutonium. The management of excess plutonium from proposed nuclear weapons dismantlement has been the subject of numerous intellectual discussions during the past several years. In this context, issues relevant to long-term management of all plutonium as a valuable energy resource are also being examined. While there are differing views about the future role of plutonium in the economy, there is a recognition of the environmental and health related problems and proliferation potentials of weapons-grade plutonium. The long-term management of plutonium as an energy resource will require a new strategy to maintain stewardship for many generations to come.

  8. Biokinetics of Plutonium in Nonhuman Primates.

    PubMed

    Poudel, Deepesh; Guilmette, Raymond A; Gesell, Thomas F; Harris, Jason T; Brey, Richard R

    2016-10-01

    A major source of data on metabolism, excretion and retention of plutonium comes from experimental animal studies. Although old world monkeys are one of the closest living relatives to humans, certain physiological differences do exist between these nonhuman primates and humans. The objective of this paper was to describe the metabolism of plutonium in nonhuman primates using the bioassay and retention data obtained from macaque monkeys injected with plutonium citrate. A biokinetic model for nonhuman primates was developed by adapting the basic model structure and adapting the transfer rates described for metabolism of plutonium in adult humans. Significant changes to the parameters were necessary to explain the shorter retention of plutonium in liver and skeleton of the nonhuman primates, differences in liver to bone partitioning ratio, and significantly higher excretion of plutonium in feces compared to that in humans.

  9. Global plutonium management: A security option

    SciTech Connect

    Sylvester, K.W.B.

    1998-12-31

    The US surplus plutonium disposition program was created to reduce the proliferation risk posed by the fissile material from thousands of retired nuclear weapons. The Department of Energy has decided to process its Put into a form as secure as Pu in civilian spent fuel. While implementation issues have been considered, a major one (Russian reciprocity) remains unresolved. Russia has made disposition action conditional on extracting the fuel value of its Pu but lacks the infrastructure to do so. Assistance in the construction of the required facilities would conflict with official US policy opposing the development of a Pu fuel cycle. The resulting stagnation provides impetus for a reevaluation of US nonproliferation objectives and Pu disposition options. A strategy for satisfying Russian fuel value concerns and reducing the proliferation risk posed by surplus weapons-grade plutonium (WGPu) is proposed. The effectiveness of material alteration (e.g., isotopic, chemical, etc.{hor_ellipsis}) at reducing the desire, ability and opportunity for proliferation is assessed. Virtually all the security benefits attainable by material processing can be obtained by immobilizing Pu in large unit size/mass monoliths without a radiation barrier. Russia would be allowed to extract the Pu at a future date for use as fuel in a verifiable manner. Remote tracking capability, if proven feasible, would further improve safeguarding capability. As an alternate approach, the US could compensate Russia for its Pu, allowing it to be disposed of or processed elsewhere. A market based method for pricing Pu is proposed. Surplus Pu could represent access to nuclear fuel at a fixed price at a future date. This position can be replicated in the uranium market and priced using derivative theory. The proposed strategy attempts to meet nonproliferation objectives by recognizing technical limitations and satisfying political constraints.

  10. PROCESS FOR THE RECOVERY OF PLUTONIUM

    DOEpatents

    Potratz, H.A.

    1958-12-16

    A process for the separation of plutonium from uranlum and other associated radioactlve fission products ls descrlbed conslstlng of contacting an acid solution containing plutonium in the tetravalent state and uranium in the hexavalent state with enough ammonium carbonate to form an alkaline solution, adding cupferron to selectlvely form plutonlum cupferrlde, then recoverlng the plutonium cupferride by extraction with a water lmmiscible organic solvent such as chloroform.

  11. PROCESS OF SEPARATING PLUTONIUM VALUES BY ELECTRODEPOSITION

    DOEpatents

    Whal, A.C.

    1958-04-15

    A process is described of separating plutonium values from an aqueous solution by electrodeposition. The process consists of subjecting an aqueous 0.1 to 1.0 N nitric acid solution containing plutonium ions to electrolysis between inert metallic electrodes. A current density of one milliampere io one ampere per square centimeter of cathode surface and a temperature between 10 and 60 d C are maintained. Plutonium is electrodeposited on the cathode surface and recovered.

  12. WET METHOD OF PREPARING PLUTONIUM TRIBROMIDE

    DOEpatents

    Davidson, N.R.; Hyde, E.K.

    1958-11-11

    S> The preparation of anhydrous plutonium tribromide from an aqueous acid solution of plutonium tetrabromide is described, consisting of adding a water-soluble volatile bromide to the tetrabromide to provide additional bromide ions sufficient to furnish an oxidation-reduction potential substantially more positive than --0.966 volt, evaporating the resultant plutonium tribromides to dryness in the presence of HBr, and dehydrating at an elevated temperature also in the presence of HBr.

  13. PLUTONIUM METAL: OXIDATION CONSIDERATIONS AND APPROACH

    SciTech Connect

    Estochen, E.

    2013-03-20

    Plutonium is arguably the most unique of all metals when considered in the combined context of metallurgical, chemical, and nuclear behavior. Much of the research in understanding behavior and characteristics of plutonium materials has its genesis in work associated with nuclear weapons systems. However, with the advent of applications in fuel materials, the focus in plutonium science has been more towards nuclear fuel applications, as well as long term storage and disposition. The focus of discussion included herein is related to preparing plutonium materials to meet goals consistent with non-proliferation. More specifically, the emphasis is on the treatment of legacy plutonium, in primarily metallic form, and safe handling, packaging, and transport to meet non-proliferation goals of safe/secure storage. Elevated temperature oxidation of plutonium metal is the treatment of choice, due to extensive experiential data related to the method, as the oxide form of plutonium is one of only a few compounds that is relatively simple to produce, and stable over a large temperature range. Despite the simplicity of the steps required to oxidize plutonium metal, it is important to understand the behavior of plutonium to ensure that oxidation is conducted in a safe and effective manner. It is important to understand the effect of changes in environmental variables on the oxidation characteristics of plutonium. The primary purpose of this report is to present a brief summary of information related to plutonium metal attributes, behavior, methods for conversion to oxide, and the ancillary considerations related to processing and facility safety. The information provided is based on data available in the public domain and from experience in oxidation of such materials at various facilities in the United States. The report is provided as a general reference for implementation of a simple and safe plutonium metal oxidation technique.

  14. PLUTONIUM-CUPFERRON COMPLEX AND METHOD OF REMOVING PLUTONIUM FROM SOLUTION

    DOEpatents

    Potratz, H.A.

    1959-01-13

    A method is presented for separating plutonium from fission products present in solutions of neutronirradiated uranium. The process consists in treating such acidic solutions with cupferron so that the cupferron reacts with the plutonium present to form an insoluble complex. This plutonium cupferride precipitates and may then be separated from the solution.

  15. Disposition options for separated plutonium

    SciTech Connect

    Hippel, F. von; Feiveson, H. )

    1993-01-01

    Russia and the United States expect to dismantle [approximately]50,000 nuclear warheads containing [approximately]150 tonnes of plutonium as a result of the drastic reductions in tactical nuclear weapons announced by Presidents Bush and Gorbachev during the fall of 1991 and the reductions in strategic weapons agreed to in the START I and START II Treaties. In addition, if current plans for reprocessing spent light water reactor (LWR) fuel are carried out (mainly in Britain and France) [approximately]200 tonnes of civilian plutonium will be separated during the 1990s. This paper addresses the public-policy issues in the U.S. and abroad regarding disposition options as well as some technical aspects for options.

  16. Plutonium Immobilization Can Loading Concepts

    SciTech Connect

    Kriikku, E.; Ward, C.; Stokes, M.; Randall, B.; Steed, J.; Jones, R.; Hamilton, L.; Rogers, L.; Fiscus, J.; Dyches, G.

    1998-05-01

    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 five can loading conceptual designs and the lists the advantages and disadvantages for each concept. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas. The can loading welder and cutter are very similar to the existing Savannah River Site (SRS) FB-Line bagless transfer welder and cutter and thus they are a low priority development item.

  17. Plutonium Immobilization Project Baseline Formulation

    SciTech Connect

    Ebbinghaus, B.

    1999-02-01

    A key milestone for the Immobilization Project (AOP Milestone 3.2a) in Fiscal Year 1998 (FY98) is the definition of the baseline composition or formulation for the plutonium ceramic form. The baseline formulation for the plutonium ceramic product must be finalized before the repository- and plant-related process specifications can be determined. The baseline formulation that is currently specified is given in Table 1.1. In addition to the baseline formulation specification, this report provides specifications for two alternative formulations, related compositional specifications (e.g., precursor compositions and mixing recipes), and other preliminary form and process specifications that are linked to the baseline formulation. The preliminary specifications, when finalized, are not expected to vary tremendously from the preliminary values given.

  18. Study of the IDGS technique for mixed plutonium-uranium (MOX) samples

    SciTech Connect

    Li, T. K.; Vo, Duc T.; Sumi, M.; Suzuki, T.

    2004-01-01

    The isotope dilution gamma-ray spectrometry (IDGS) technique has been demonstrated for simultaneously measuring concentrations and isotopic compositions of plutonium in spent-fuel input dissolver solutions. For timely analyzing nuclear materials on the purpose of material accountancy and quality control/assurance, we have performed a feasibility study to implement the IDGS for measuring mixed plutonium-uranium oxide (MOX) samples at the Plutonium Fuel Center (PFC) of Japan Nuclear Cycle Development Institute (JNC). Proof-of-principle experiments and analysis have been conducted for developing simultaneous plutonium and uranium measurements in MOX samples with wide variation of Pu/U ratios including powder, pellets and process scraps from the MOX fuel fabrication plant at PFC. We have shown that FRAM can be used with the IDGS technique to simultaneously determine plutonium and uranium isotopic compositions and concentrations in MOX samples at PFC, JNC. The uncertainties of the results are somewhat large due to weak statistics. If better statistics are obtained by either using more plutonium in the measurements, acquire the data for longer time, or using higher efficiency detector then the results can be better. The accuracy of the results can also be improved by a factor of 2-3 by using the generalized IDGS technique instead of this traditional IDGS.

  19. NON-AQUEOUS DISSOLUTION OF MASSIVE PLUTONIUM

    DOEpatents

    Reavis, J.G.; Leary, J.A.; Walsh, K.A.

    1959-05-12

    A method is presented for obtaining non-aqueous solutions or plutonium from massive forms of the metal. In the present invention massive plutonium is added to a salt melt consisting of 10 to 40 weight per cent of sodium chloride and the balance zinc chloride. The plutonium reacts at about 800 deg C with the zinc chloride to form a salt bath of plutonium trichloride, sodium chloride, and metallic zinc. The zinc is separated from the salt melt by forcing the molten mixture through a Pyrex filter.

  20. METHOD OF REDUCING PLUTONIUM WITH FERROUS IONS

    DOEpatents

    Dreher, J.L.; Koshland, D.E.; Thompson, S.G.; Willard, J.E.

    1959-10-01

    A process is presented for separating hexavalent plutonium from fission product values. To a nitric acid solution containing the values, ferrous ions are added and the solution is heated and held at elevated temperature to convert the plutonium to the tetravalent state via the trivalent state and the plutonium is then selectively precipitated on a BiPO/sub 4/ or LaF/sub 3/ carrier. The tetravalent plutonium formed is optionally complexed with fluoride, oxalate, or phosphate anion prior to carrier precipitation.

  1. OXIDATIVE METHOD OF SEPARATING PLUTONIUM FROM NEPTUNIUM

    DOEpatents

    Beaufait, L.J. Jr.

    1958-06-10

    A method is described of separating neptunium from plutonium in an aqueous solution containing neptunium and plutonium in valence states not greater than +4. This may be accomplished by contacting the solution with dichromate ions, thus oxidizing the neptunium to a valence state greater than +4 without oxidizing any substantial amount of plutonium, and then forming a carrier precipitate which carries the plutonium from solution, leaving the neptunium behind. A preferred embodiment of this invention covers the use of lanthanum fluoride as the carrier precipitate.

  2. ION EXCHANGE ADSORPTION PROCESS FOR PLUTONIUM SEPARATION

    DOEpatents

    Boyd, G.E.; Russell, E.R.; Taylor, M.D.

    1961-07-11

    Ion exchange processes for the separation of plutonium from fission products are described. In accordance with these processes an aqueous solution containing plutonium and fission products is contacted with a cation exchange resin under conditions favoring adsorption of plutonium and fission products on the resin. A portion of the fission product is then eluted with a solution containing 0.05 to 1% by weight of a carboxylic acid. Plutonium is next eluted with a solution containing 2 to 8 per cent by weight of the same carboxylic acid, and the remaining fission products on the resin are eluted with an aqueous solution containing over 10 per cent by weight of sodium bisulfate.

  3. Plutonium Oxidation State Distribution under Aerobic and Anaerobic Subsurface Conditions for Metal-Reducing Bacteria

    NASA Astrophysics Data System (ADS)

    Reed, D. T.; Swanson, J.; Khaing, H.; Deo, R.; Rittmann, B.

    2009-12-01

    The fate and potential mobility of plutonium in the subsurface is receiving increased attention as the DOE looks to cleanup the many legacy nuclear waste sites and associated subsurface contamination. Plutonium is the near-surface contaminant of concern at several DOE sites and continues to be the contaminant of concern for the permanent disposal of nuclear waste. The mobility of plutonium is highly dependent on its redox distribution at its contamination source and along its potential migration pathways. This redox distribution is often controlled, especially in the near-surface where organic/inorganic contaminants often coexist, by the direct and indirect effects of microbial activity. The redox distribution of plutonium in the presence of facultative metal reducing bacteria (specifically Shewanella and Geobacter species) was established in a concurrent experimental and modeling study under aerobic and anaerobic conditions. Pu(VI), although relatively soluble under oxidizing conditions at near-neutral pH, does not persist under a wide range of the oxic and anoxic conditions investigated in microbiologically active systems. Pu(V) complexes, which exhibit high chemical toxicity towards microorganisms, are relatively stable under oxic conditions but are reduced by metal reducing bacteria under anaerobic conditions. These facultative metal-reducing bacteria led to the rapid reduction of higher valent plutonium to form Pu(III/IV) species depending on nature of the starting plutonium species and chelating agents present in solution. Redox cycling of these lower oxidation states is likely a critical step in the formation of pseudo colloids that may lead to long-range subsurface transport. The CCBATCH biogeochemical model is used to explain the redox mechanisms and final speciation of the plutonium oxidation state distributions observed. These results for microbiologically active systems are interpreted in the context of their importance in defining the overall migration

  4. Determination of filter pore size for use in HB line phase II production of plutonium oxide

    SciTech Connect

    Shehee, T.; Crowder, M.; Rudisill, T.

    2014-08-01

    H-Canyon and HB-Line are tasked with the production of plutonium oxide (PuO2) from a feed of plutonium (Pu) metal. The PuO2 will provide feed material for the Mixed Oxide (MOX) Fuel Fabrication Facility. After dissolution of the Pu metal in H-Canyon, plans are to transfer the solution to HB-Line for purification by anion exchange. Anion exchange will be followed by plutonium(IV) oxalate precipitation, filtration, and calcination to form PuO2. The filtrate solutions, remaining after precipitation, contain low levels of Pu ions, oxalate ions, and may include solids. These solutions are transferred to H-Canyon for disposition. To mitigate the criticality concern of Pu solids in a Canyon tank, past processes have used oxalate destruction or have pre-filled the Canyon tank with a neutron poison. The installation of a filter on the process lines from the HB-Line filtrate tanks to H-Canyon Tank 9.6 is proposed to remove plutonium oxalate solids. This report describes SRNL’s efforts to determine the appropriate pore size for the filters needed to perform this function. Information provided in this report aids in developing the control strategies for solids in the process.

  5. Pyrochemical process for extracting plutonium from an electrolyte salt

    DOEpatents

    Mullins, Lawrence J.; Christensen, Dana C.

    1984-01-01

    A pyrochemical process for extracting plutonium from a plutonium-bearing salt is disclosed. The process is particularly useful in the recovery of plutonium from electrolyte salts which are left over from the electrorefining of plutonium. In accordance with the process, the plutonium-bearing salt is melted and mixed with metallic calcium. The calcium reduces ionized plutonium in the salt to plutonium metal, and also causes metallic plutonium in the salt, which is typically present as finely dispersed metallic shot, to coalesce. The reduced and coalesced plutonium separates out on the bottom of the reaction vessel as a separate metallic phase which is readily separable from the overlying salt upon cooling of the mixture. Yields of plutonium are typically on the order of 95%. The stripped salt is virtually free of plutonium and may be discarded to low-level waste storage.

  6. Pyrochemical process for extracting plutonium from an electrolyte salt

    DOEpatents

    Mullins, L.J.; Christensen, D.C.

    1982-09-20

    A pyrochemical process for extracting plutonium from a plutonium-bearing salt is disclosed. The process is particularly useful in the recovery of plutonium for electrolyte salts which are left over from the electrorefining of plutonium. In accordance with the process, the plutonium-bearing salt is melted and mixed with metallic calcium. The calcium reduces ionized plutonium in the salt to plutonium metal, and also causes metallic plutonium in the salt, which is typically present as finely dispersed metallic shot, to coalesce. The reduced and coalesced plutonium separates out on the bottom of the reaction vessel as a separate metallic phase which is readily separable from the overlying salt upon cooling of the mixture. Yields of plutonium are typically on the order of 95%. The stripped salt is virtually free of plutonium and may be discarded to low-level waste storage.

  7. Plutonium oxalate precipitation for trace elemental determination in plutonium materials

    SciTech Connect

    Xu, Ning; Gallimore, David; Lujan, Elmer; Garduno, Katherine; Walker, Laurie; Taylor, Fiona; Thompson, Pam; Tandon, Lav

    2015-05-26

    In this study, an analytical chemistry method has been developed that removes the plutonium (Pu) matrix from the dissolved Pu metal or oxide solution prior to the determination of trace impurities that are present in the metal or oxide. In this study, a Pu oxalate approach was employed to separate Pu from trace impurities. After Pu(III) was precipitated with oxalic acid and separated by centrifugation, trace elemental constituents in the supernatant were analyzed by inductively coupled plasma-optical emission spectroscopy with minimized spectral interferences from the sample matrix.

  8. Plutonium oxalate precipitation for trace elemental determination in plutonium materials

    DOE PAGES

    Xu, Ning; Gallimore, David; Lujan, Elmer; ...

    2015-05-26

    In this study, an analytical chemistry method has been developed that removes the plutonium (Pu) matrix from the dissolved Pu metal or oxide solution prior to the determination of trace impurities that are present in the metal or oxide. In this study, a Pu oxalate approach was employed to separate Pu from trace impurities. After Pu(III) was precipitated with oxalic acid and separated by centrifugation, trace elemental constituents in the supernatant were analyzed by inductively coupled plasma-optical emission spectroscopy with minimized spectral interferences from the sample matrix.

  9. Pool Purification

    NASA Technical Reports Server (NTRS)

    1988-01-01

    Caribbean Clear, Inc. used NASA's silver ion technology as a basis for its automatic pool purifier. System offers alternative approach to conventional purification chemicals. Caribbean Clear's principal markets are swimming pool owners who want to eliminate chlorine and bromine. Purifiers in Caribbean Clear System are same silver ions used in Apollo System to kill bacteria, plus copper ions to kill algae. They produce spa or pool water that exceeds EPA Standards for drinking water.

  10. Polonium purification

    SciTech Connect

    Baker, J.D.

    1996-09-01

    Three processes for the purification of {sup 210}Po from irradiated bismuth targets are described. Safety equipment includes shielded hotcells for the initial separation from other activation products, gloveboxes for handling the volatile and highly toxic materials, and provisions for ventilation. All chemical separations must be performed under vacuum or in inerted systems. Two of the processes require large amounts of electricity; the third requires vessels made from exotic materials.

  11. RECOVERY OF PLUTONIUM BY CARRIER PRECIPITATION

    DOEpatents

    Goeckermann, R.H.

    1961-04-01

    A process is given for recovering plutonium from an aqueous nitric acid zirconium-containing solution of an acidity between 0.2 and 1 N by adding fluoride anions (1.5 to 5 mg/l) and precipitating the plutonium with an excess of hydrogen peroxide at from 53 to 65 deg C.

  12. URANOUS IODATE AS A CARRIER FOR PLUTONIUM

    DOEpatents

    Miller, D.R.; Seaborg, G.T.; Thompson, S.G.

    1959-12-15

    A process is described for precipitating plutonium on a uranous iodate carrier from an aqueous acid solution conA plutonium solution more concentrated than the original solution can then be obtained by oxidizing the uranium to the hexavalent state and dissolving the precipitate, after separating the latter from the original solution, by means of warm nitric acid.

  13. Surplus Plutonium Disposition Final Environmental Impact Statement

    SciTech Connect

    N /A

    1999-11-19

    In December 1996, the U.S. Department of Energy (DOE) published the ''Storage and Disposition of Weapons-Usable Fissile Materials Final Programmatic Environmental Impact Statement (Storage and Disposition PEIS)'' (DOE 1996a). That PEIS analyzes the potential environmental consequences of alternative strategies for the long-term storage of weapons-usable plutonium and highly enriched uranium (HEU) and the disposition of weapons-usable plutonium that has been or may be declared surplus to national security needs. The Record of Decision (ROD) for the ''Storage and Disposition PEIS'', issued on January 14, 1997 (DOE 1997a), outlines DOE's decision to pursue an approach to plutonium disposition that would make surplus weapons-usable plutonium inaccessible and unattractive for weapons use. DOE's disposition strategy, consistent with the Preferred Alternative analyzed in the ''Storage and Disposition PEIS'', allows for both the immobilization of some (and potentially all) of the surplus plutonium and use of some of the surplus plutonium as mixed oxide (MOX) fuel in existing domestic, commercial reactors. The disposition of surplus plutonium would also involve disposal of both the immobilized plutonium and the MOX fuel (as spent nuclear fuel) in a potential geologic repository.

  14. MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

    DOEpatents

    Kiehn, R.M.; King, L.D.P.; Peterson, R.E.; Swickard, E.O. Jr.

    1962-06-26

    A description is given of a nuclear fast reactor fueled with molten plutonium containing about 20 kg of plutonium in a tantalum container, cooled by circulating liquid sodium at about 600 to 650 deg C, having a large negative temperature coefficient of reactivity, and control rods and movable reflector for criticality control. (AEC)

  15. Uses for plutonium: Weapons, reactors, and other

    SciTech Connect

    Condit, R.H.

    1994-05-01

    This document begins with a introduction on criticality and supercriticality. Then, types and components, design and engineering, yields, and disassembly of nuclear weapons are discussed. Plutonium is evaluated as a reactor fuel, including neutronics and chemistry considerations. Finally, other uses of plutonium are analyzed.

  16. Source Book on Plutonium and Its Decontamination

    DTIC Science & Technology

    1973-09-24

    Data Entered) UNCLASIFIED 20. ABSTRACT (Continued) |development of the coupled differential equations, based on the 1965 and the proposed 1973...61 XV Some Foreign Plutonium Decontamination Standards . . ...... 63 XVI Variability of Sol Sampling Data .... ..... .... 64 XVII Criteria for...Scheduling Feces Samples . . .......... 66 XVIII Types of Data which may be Coliected for Plutonium Inhalation Incidents . 66 XIX Percent Efficiencies for

  17. Proceedings of the Plutonium Futures ? The Science 2006 Conference

    SciTech Connect

    Fluss, M; Hobart, D; Allan, P; Jarvinen, G

    2007-07-12

    Plutonium Futures--The Science 2006 provided opportunities to examine present knowledge of the chemical and physical properties of plutonium and other actinides in complex media and materials; to discuss the current and emerging science (chemistry, physics, materials science, nuclear science, and environmental effects) of plutonium and actinides relevant to enhancing global nuclear security; and to exchange ideas. This international conference also provided a forum for illustrating and enhancing capabilities and interests, and assessing issues in these areas. U.S. and international scientists, engineers, faculty, and students from universities, national laboratories, and DOE's nuclear complex were encouraged to participate and make technical contributions. The Conference ran from Sunday, July 9th through Thursday, July 13th. A popular aspect of the conference was the opening tutorial session on Sunday afternoon intended for students and scientists new to the area of plutonium research. The tutorial was well attended by novices and veterans alike, and featured such diverse topics as; plutonium metallurgy, plutonium in the environment, and international arms control and nonproliferation. Two plenary lectures began each morning and each afternoon session and highlighted the breakout sessions on coordination/organometallic chemistry, solid-state physics, environmental chemistry, materials science, separations and reprocessing, advanced fuels and waste forms, phase transformations, solution and gas-phase chemistry, compounds and complexes, electronic structure and physical properties, and more. Chemistry Highlights--Among the many chemistry highlights presented in this proceedings are the overview of concepts and philosophies on inert nuclear fuel matrices and concerns about the ever-increasing amounts of minor actinides and plutonium generated in the fuel cycle. The various ideas involve multiple reduction schemes for these materials, suggesting fuels for 'burning' or

  18. Nondestructive assay methods for solids containing plutonium

    SciTech Connect

    Macmurdo, K.W.; Gray, L.W.; Gibbs, A.

    1984-06-01

    Specific nondestructive assay (NDA) methods, e.g. calorimetry, coincidence neutron counting, singles neutron counting, and gamma ray spectrometry, were studied to provide the Savannah River Plant with an NDA method to measure the plutonium content of solid scrap (slag and crucible) generated in the JB-Line plutonium metal production process. Results indicate that calorimetry can be used to measure the plutonium content to within about 3% in 4 to 6 hours by using computerized equilibrium sample power predictive models. Calorimetry results confirm that a bias exists in the present indirect measurement method used to estimate the plutonium content of slag and crucible. Singles neutron counting of slag and crucible can measure plutonium to only +-30%, but coincidence neutron counting methods improve measurement precision to better than +-10% in less than ten minutes. Only four portions of a single slag and crucible sample were assayed, and further study is recommended.

  19. Application of PGNAA to plutonium surveillance

    SciTech Connect

    Prettyman, T.H.; Foster, L.A.; Staples, P.

    1997-12-01

    Prompt gamma-ray neutron activation analysis (PGNAA) is a well-established tool for nondestructive elemental analysis of bulk samples. At Los Alamos National Laboratory we are investigating the use of PGNAA as a diagnostic tool for a number of applications, particularly matrix characterization for nondestructive assay and plutonium surveillance. Surveillance is an essential feature of most plutonium facility operations, including routine material processing and research, short-term storage, and processing operations prior to disposal or long-term storage. The ability to identify and assay specific elements from gamma-ray-produced active neutron interrogation (e.g., by neutron capture, nonelastic scattering, and the decay of activation products) makes PGNAA an ideal tool for surveillance. For example, PGNAA can help confirm item descriptions (for example, plutonium chloride versus plutonium oxide). This feature is particularly important in operations involving poorly characterized legacy materials where the material form could adversely impact plutonium-processing operations.

  20. PROCESS FOR THE RECOVERY OF PLUTONIUM

    DOEpatents

    Ritter, D.M.

    1959-01-13

    An improvement is presented in the process for recovery and decontamination of plutonium. The carrier precipitate containing plutonium is dissolved and treated with an oxidizing agent to place the plutonium in a hexavalent oxidation state. A lanthanum fluoride precipitate is then formed in and removed from the solution to carry undesired fission products. The fluoride ions in the reniaining solution are complexed by addition of a borate sueh as boric acid, sodium metaborate or the like. The plutonium is then reduced and carried from the solution by the formation of a bismuth phosphate precipitate. This process effects a better separation from unwanted flssion products along with conccntration of the plutonium by using a smaller amount of carrier.

  1. New Fecal Method for Plutonium and Americium

    SciTech Connect

    Maxwell, S.L. III

    2000-06-27

    A new fecal analysis method that dissolves plutonium oxide was developed at the Westinghouse Savannah River Site. Diphonix Resin (Eichrom Industries), is used to pre-concentrate the actinides from digested fecal samples. A rapid microwave digestion technique is used to remove the actinides from the Diphonix Resin, which effectively extracts plutonium and americium from acidic solutions containing hydrofluoric acid. After resin digestion, the plutonium and americium are recovered in a small volume of nitric acid that is loaded onto small extraction chromatography columns, TEVA Resin and TRU Resin (Eichrom Industries). The method enables complete dissolution of plutonium oxide and provides high recovery of plutonium and americium with good removal of thorium isotopes such as thorium-228.

  2. REMOVAL OF LEGACY PLUTONIUM MATERIALS FROM SWEDEN

    SciTech Connect

    Dunn, Kerry A.; Bellamy, J. Steve; Chandler, Greg T.; Iyer, Natraj C.; Koenig, Rich E.; Leduc, D.; Hackney, B.; Leduc, Dan R.; McClard, J. W.

    2013-08-18

    U.S. Department of Energy’s National Nuclear Security Administration (NNSA) Office of Global Threat Reduction (GTRI) recently removed legacy plutonium materials from Sweden in collaboration with AB SVAFO, Sweden. This paper details the activities undertaken through the U.S. receiving site (Savannah River Site (SRS)) to support the characterization, stabilization, packaging and removal of legacy plutonium materials from Sweden in 2012. This effort was undertaken as part of GTRI’s Gap Materials Program and culminated with the successful removal of plutonium from Sweden as announced at the 2012 Nuclear Security Summit. The removal and shipment of plutonium materials to the United States was the first of its kind under NNSA’s Global Threat Reduction Initiative. The Environmental Assessment for the U.S. receipt of gap plutonium material was approved in May 2010. Since then, the multi-year process yielded many first time accomplishments associated with plutonium packaging and transport activities including the application of the of DOE-STD-3013 stabilization requirements to treat plutonium materials outside the U.S., the development of an acceptance criteria for receipt of plutonium from a foreign country, the development and application of a versatile process flow sheet for the packaging of legacy plutonium materials, the identification of a plutonium container configuration, the first international certificate validation of the 9975 shipping package and the first intercontinental shipment using the 9975 shipping package. This paper will detail the technical considerations in developing the packaging process flow sheet, defining the key elements of the flow sheet and its implementation, determining the criteria used in the selection of the transport package, developing the technical basis for the package certificate amendment and the reviews with multiple licensing authorities and most importantly integrating the technical activities with the Swedish partners.

  3. Rapid determination of (237)Np and plutonium isotopes in urine by inductively-coupled plasma mass spectrometry and alpha spectrometry.

    PubMed

    Maxwell, Sherrod L; Culligan, Brian K; Jones, Vernon D; Nichols, Sheldon T; Noyes, Gary W; Bernard, Maureen A

    2011-08-01

    A new rapid separation method was developed for the measurement of plutonium and neptunium in urine samples by inductively-coupled plasma mass spectrometry (ICP-MS) and/or alpha spectrometry with enhanced uranium removal. This method allows separation and preconcentration of plutonium and neptunium in urine samples using stacked extraction chromatography cartridges and vacuum box flow rates to facilitate rapid separations. There is an increasing need to develop faster analytical methods for emergency response samples. There is also enormous benefit to having rapid bioassay methods in the event that a nuclear worker has an uptake (puncture wound, etc.) to assess the magnitude of the uptake and guide efforts to mitigate dose (e.g., tissue excision and chelation therapy). This new method focuses only on the rapid separation of plutonium and neptunium with enhanced removal of uranium. For ICP-MS, purified solutions must have low salt content and low concentration of uranium due to spectral interference of (238)U(1)H(+) on m/z 239. Uranium removal using this method is enhanced by loading plutonium and neptunium initially onto TEVA resin, then moving plutonium to DGA resin where additional purification from uranium is performed with a decontamination factor of almost 1×10(5). If UTEVA resin is added to the separation scheme, a decontamination factor of ~3 × 10(6) can be achieved.

  4. ADSORPTION-BISMUTH PHOSPHATE METHOD FOR SEPARATING PLUTONIUM

    DOEpatents

    Russell, E.R.; Adamson, A.W.; Boyd, G.E.

    1960-06-28

    A process is given for separating plutonium from uranium and fission products. Plutonium and uranium are adsorbed by a cation exchange resin, plutonium is eluted from the adsorbent, and then, after oxidation to the hexavalent state, the plutonium is contacted with a bismuth phosphate carrier precipitate.

  5. Plutonium Uptake and Distribution in Mammalian Cells: Molecular vs Polymeric Plutonium

    PubMed Central

    ARYAL, BAIKUNTHA P.; GORMAN-LEWIS, DREW; PAUNESKU, TATJANA; WILSON, RICHARD E.; LAI, BARRY; VOGT, STEFAN; WOLOSCHAK, GAYLE E.; JENSEN, MARK P.

    2013-01-01

    Purpose To study the cellular responses to molecular and polymeric forms of plutonium using PC12 cells derived from rat adrenal glands. Materials and methods Serum starved PC12 cells were exposed to polymeric and molecular forms of plutonium for three hours. Cells were washed with 10 mM EGTA, 100 mM NaCl at pH 7.4 to remove surface sorbed plutonium. Localization of plutonium in individual cell was quantitatively analyzed by synchrotron X-ray fluorescence (XRF) microscopy. Results Molecular plutonium complexes introduced to cell growth media in the form of NTA, citrate, or transferrin complexes were taken up by PC12 cells, and mostly co-localized with iron within the cells. Polymeric plutonium prepared separately was not internalized by PC12 cells but it was always found on the cell surface as big agglomerates; however polymeric plutonium formed in situ was mostly found within the cells as agglomerates. Conclusions PC12 cells can differentiate molecular and polymeric forms of plutonium. Molecular plutonium is taken up by PC12 cells and mostly co-localized with iron but aged polymeric plutonium is not internalized by the cells. PMID:21770702

  6. Plutonium uptake and distribution in mammalian cells: molecular vs. polymeric plutonium.

    PubMed

    Aryal, Baikuntha P; Gorman-Lewis, Drew; Paunesku, Tatjana; Wilson, Richard E; Lai, Barry; Vogt, Stefan; Woloschak, Gayle E; Jensen, Mark P

    2011-10-01

    To study the cellular responses to molecular and polymeric forms of plutonium using PC12 cells derived from a rat pheochromocytoma. Serum starved PC12 cells were exposed to polymeric and molecular forms of plutonium for 3 h. Cells were washed with 10 mM ethylene glycol tetraacetic acid (EGTA), 100 mM NaCl at pH 7.4 to remove surface sorbed plutonium. Localization of plutonium in individual cell was quantitatively analyzed by synchrotron X-ray fluorescence (XRF) microscopy. Molecular plutonium complexes introduced to cell growth media in the form of nitrilotriacetic acid (NTA), citrate, or transferrin complexes were taken up by PC12 cells, and mostly colocalized with iron within the cells. Aged polymeric plutonium prepared separately was not internalized by PC12 cells but it was always found on the cell surface as big agglomerates; however, polymeric plutonium formed in situ was mostly found within the cells as agglomerates. PC12 cells can differentiate molecular and polymeric forms of plutonium. Molecular plutonium is taken up by PC12 cells and mostly co-localizes with iron but aged polymeric plutonium is not internalized by the cells.

  7. Potentiometric determination of plutonium by sodium bismuthate oxidation.

    PubMed

    Charyulu, M M; Rao, V K; Natarajan, P R

    1984-12-01

    A potentiometric method for the determination of plutonium is described, in which the plutonium is quantitatively oxidized to plutonium(VI) with sodium bismuthate in nitric acid medium, the excess of oxidant is destroyed chemically and plutonium(VI) is reduced to plutonium(IV) with a measured excess of iron(II), the surplus of which is back-titrated with dichromate. For 3-5 mg of plutonium the error is less than 0.2%. For submilligram quantities of plutonium in presence of macro-amounts of uranium the error is below 2.0%.

  8. Laboratory-scale evaluations of alternative plutonium precipitation methods

    SciTech Connect

    Martella, L.L.; Saba, M.T.; Campbell, G.K.

    1984-02-08

    Plutonium(III), (IV), and (VI) carbonate; plutonium(III) fluoride; plutonium(III) and (IV) oxalate; and plutonium(IV) and (VI) hydroxide precipitation methods were evaluated for conversion of plutonium nitrate anion-exchange eluate to a solid, and compared with the current plutonium peroxide precipitation method used at Rocky Flats. Plutonium(III) and (IV) oxalate, plutonium(III) fluoride, and plutonium(IV) hydroxide precipitations were the most effective of the alternative conversion methods tested because of the larger particle-size formation, faster filtration rates, and the low plutonium loss to the filtrate. These were found to be as efficient as, and in some cases more efficient than, the peroxide method. 18 references, 14 figures, 3 tables.

  9. Ceramification: A plutonium immobilization process

    SciTech Connect

    Rask, W.C.; Phillips, A.G.

    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.

  10. Plutonium focus area. Technology summary

    SciTech Connect

    1997-09-01

    The Assistant Secretary for the Office of Environmental Management (EM) at the U.S. Department of Energy (DOE) chartered the Plutonium Focus Area (PFA) in October 1995. The PFA {open_quotes}...provides for peer and technical reviews of research and development in plutonium stabilization activities...{close_quotes} In addition, the PFA identifies and develops relevant research and technology. The purpose of this document is to focus attention on the requirements used to develop research and technology for stabilization, storage, and preparation for disposition of nuclear materials. The PFA Technology Summary presents the approach the PFA uses to identify, recommend, and review research. It lists research requirements, research being conducted, and gaps where research is needed. It also summarizes research performed by the PFA in the traditional research summary format. This document encourages researchers and commercial enterprises to do business with PFA by submitting research proposals or {open_quotes}white papers.{close_quotes} In addition, it suggests ways to increase the likelihood that PFA will recommend proposed research to the Nuclear Materials Stabilization Task Group (NMSTG) of DOE.

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

    SciTech Connect

    Persiani, P.J.

    1995-08-01

    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.

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

    SciTech Connect

    Persiani, P.J.

    1995-12-31

    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.

  13. 10 CFR 140.108 - Appendix H-Form of indemnity agreement with licensees possessing plutonium for use in plutonium...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... possessing plutonium for use in plutonium processing and fuel fabrication plants and furnishing proof of... Appendixes to Part 140 § 140.108 Appendix H—Form of indemnity agreement with licensees possessing plutonium for use in plutonium processing and fuel fabrication plants and furnishing proof of...

  14. 10 CFR 140.108 - Appendix H-Form of indemnity agreement with licensees possessing plutonium for use in plutonium...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... possessing plutonium for use in plutonium processing and fuel fabrication plants and furnishing proof of... Appendixes to Part 140 § 140.108 Appendix H—Form of indemnity agreement with licensees possessing plutonium for use in plutonium processing and fuel fabrication plants and furnishing proof of...

  15. 10 CFR 140.108 - Appendix H-Form of indemnity agreement with licensees possessing plutonium for use in plutonium...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... possessing plutonium for use in plutonium processing and fuel fabrication plants and furnishing proof of... Appendixes to Part 140 § 140.108 Appendix H—Form of indemnity agreement with licensees possessing plutonium for use in plutonium processing and fuel fabrication plants and furnishing proof of...

  16. 10 CFR 140.108 - Appendix H-Form of indemnity agreement with licensees possessing plutonium for use in plutonium...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... possessing plutonium for use in plutonium processing and fuel fabrication plants and furnishing proof of... Appendixes to Part 140 § 140.108 Appendix H—Form of indemnity agreement with licensees possessing plutonium for use in plutonium processing and fuel fabrication plants and furnishing proof of...

  17. Water Purification

    NASA Technical Reports Server (NTRS)

    1992-01-01

    Silver ionization water purification technology was originally developed for Apollo spacecraft. It was later used to cleanse swimming pools and has now been applied to industrial cooling towers and process coolers. Sensible Technologies, Inc. has added two other technologies to the system, which occupies only six square feet. It is manufactured in three capacities, and larger models are custom built on request. The system eliminates scale, corrosion, algae, bacteria and debris, and because of the NASA technology, viruses and waterborne bacteria are also destroyed. Applications include a General Motors cooling tower, amusement parks, ice manufacture and a closed-loop process cooling system.

  18. PURIFICATION PROCESS

    DOEpatents

    Wibbles, H.L.; Miller, E.I.

    1958-01-14

    This patent deals with the separation of uranium from molybdenum compounds, and in particular with their separation from ether solutions containing the molybdenum in the form of acids, such as silicomolybdic and phosphomolybdic acids. After the nitric acid leach of pitchblende, the molybdenum values present in the ore are found in the leach solution in the form of complex acids. The uranium bearing solution may be purified of this molybdenum content by comtacting it with activated charcoal. The purification is improved when the acidity of the solution is low ad agitation is also beneficial. The molybdenum may subsequently be recovered from the charcosl ad the charcoal reused.

  19. Volatile fluoride process for separating plutonium from other materials

    DOEpatents

    Spedding, F. H.; Newton, A. S.

    1959-04-14

    The separation of plutonium from uranium and/or fission products by formation of the higher fluorides off uranium and/or plutonium is described. Neutronirradiated uranium metal is first converted to the hydride. This hydrided product is then treated with fluorine at about 315 deg C to form and volatilize UF/sub 6/ leaving plutonium behind. Thc plutonium may then be separated by reacting the residue with fluorine at about 5004DEC and collecting the volatile plutonium fluoride thus formed.

  20. VOLATILE FLUORIDE PROCESS FOR SEPARATING PLUTONIUM FROM OTHER MATERIALS

    DOEpatents

    Spedding, F.H.; Newton, A.S.

    1959-04-14

    The separation of plutonium from uranium and/or tission products by formation of the higher fluorides of uranium and/or plutonium is discussed. Neutronirradiated uranium metal is first convcrted to the hydride. This hydrided product is then treatced with fluorine at about 315 deg C to form and volatilize UF/sup 6/ leaving plutonium behind. The plutonium may then be separated by reacting the residue with fluorine at about 500 deg C and collecting the volatile plutonium fluoride thus formed.

  1. Plutonium in the atmosphere: A global perspective.

    PubMed

    Thakur, P; Khaing, H; Salminen-Paatero, S

    2017-09-01

    A number of potential source terms have contributed plutonium isotopes to the atmosphere. The atmospheric nuclear weapon tests conducted between 1945 and 1980 and the re-entry of the burned SNAP-9A satellite in 1964, respectively. It is generally believed that current levels of plutonium in the stratosphere are negligible and compared with the levels generally found at surface-level air. In this study, the time trend analysis and long-term behavior of plutonium isotopes ((239+240)Pu and (238)Pu) in the atmosphere were assessed using historical data collected by various national and international monitoring networks since 1960s. An analysis of historical data indicates that (239+240)Pu concentration post-1984 is still frequently detectable, whereas (238)Pu is detected infrequently. Furthermore, the seasonal and time-trend variation of plutonium concentration in surface air followed the stratospheric trends until the early 1980s. After the last Chinese test of 1980, the plutonium concentrations in surface air dropped to the current levels, suggesting that the observed concentrations post-1984 have not been under stratospheric control, but rather reflect the environmental processes such as resuspension. Recent plutonium atmospheric air concentrations data show that besides resuspension, other environmental processes such as global dust storms and biomass burning/wildfire also play an important role in redistributing plutonium in the atmosphere. Copyright © 2017 Elsevier Ltd. All rights reserved.

  2. How much plutonium does North Korea have?

    SciTech Connect

    Albright, D.

    1994-09-01

    U.S. intelligence discovered in the 1980s that North Korea was building a small nuclear reactor. The reactor was described as a gas-cooled, graphite-moderated model similar to those Britian and France used to produce electric power as well as plutonium for nuclear weapons. When Western nations expressed concern about the reactor Russia pressed North Korea to sign the Non-Proliferation Treaty (NPT) which it did on December 12, 1985. However, North Korea stalled on signing the required safeguards agreement that allows the International Atomic Energy Agency (IAEA) to inspect nuclear facilities until January 1992. Inspections by the IAEA revealed discrepancies with the amounts of plutonium separated as declared by the North Koreans. The IAEA also received reports that two North Korean waste sites were hidden. By February 1993 the IAEA and the North Koreans has reached an impasse: North Koreas initial declarations of plutonium inventory could not be confirmed and North Korea refused to cooperate. At the least, North Korea admits to having separated 100 grams of plutonium. At the most, worst case estimate, they could have a total of 6 - 13 kilograms of separated plutonium. A first nuclear weapon can require up to 10 kilograms of weapon-grade plutonium. Any settlement needs to include a way to insure that the IAEA can verify North Korea`s past nuclear activities and determine the amount of plutonium that may have been separated in the past. 2 refs.

  3. Recycle of scrap plutonium-238 oxide fuel to support future radioisotope applications

    SciTech Connect

    Schulte, L.D.; Espinoza, J.M.; Ramsey, K.B.; Rinehart, G.H.; Silver, G.L.; Purdy, G.M.; Jarvinen, G.D.

    1997-11-01

    The Nuclear Materials Technology (NMT) Division of Los Alamos National Laboratory has initiated a development program to recover and purify plutonium-238 oxide from impure feed sources in a glove box environment. A glove box line has been designed and a chemistry flowsheet developed to perform this recovery task at large scale. The initial demonstration effort focused on purification of {sup 238}PuO{sub 2} fuel by HNO{sub 3}/HF dissolution, followed by plutonium(III) oxalate precipitation and calcination to an oxide. Decontamination factors for most impurities of concern in the fuel were very good, producing {sup 238}PuO{sub 2} fuel significantly better in purity than specified by General Purpose Heat Source (GPHS) fuel powder specifications. The results are encouraging for recycle of relatively impure plutonium-238 oxide and scrap residue items into fuel for useful applications. A sufficient quantity of purified {sup 238}PuO{sub 2} fuel was recovered from the process to allow fabrication of a GPHS unit for testing. The high specific activity of plutonium-238 magnifies the consequences and concerns of radioactive waste generation. This work places an emphasis on development of waste minimization technologies to complement the aqueous processing operation. Results from experiments allowing more time for neutralized solutions of plutonium-238 to precipitate resulted in decontamination to about 1 millicurie/L. Combining ultrafiltration treatment with addition of a water-soluble polymer designed to coordinate Pu, allowed solutions to be decontaminated to about 1 microcurie/L. Efforts continue to develop a capability for efficient, safe, cost-effective, and environmentally acceptable methods to recover and purify {sup 238}PuO{sub 2} fuel.

  4. Measurement of achievable plutonium decontamination from gallium by means of PUREX solvent extraction

    SciTech Connect

    Collins, E.D.; Campbell, D.O.; Felker, L.K.

    2000-01-01

    The objective of the work described herein was to measure, experimentally, the achievable decontamination of plutonium from gallium by means of the PUREX solvent extraction process. Gallium is present in surplus weapons-grade plutonium (WG-Pu) at a concentration of approximately 1 wt%. Plans are to dispose of surplus WG-Pu by converting it to UO{sub 2}-PuO{sub 2} mixed oxide (MOX) fuel and irradiating it in commercial power reactors. However, the presence of high concentrations of gallium in plutonium is a potential corrosion problem during the process of MOX fuel irradiation. The batch experiments performed in this study were designed to measure the capability of the PUREX solvent extraction process to separate gallium from plutonium under idealized conditions. Radioactive tracing of the gallium with {sup 72}Ga enabled the accurate measurement of low concentrations of extractable gallium. The experiments approximated the proposed flowsheet for WG-Pu purification, except that only one stage was used for each process: extraction, scrubbing, and stripping. With realistic multistage countercurrent systems, much more efficient separations are generally obtained. The gallium decontamination factor (DF) obtained after one extraction stage was about 3 x 10{sup 6}. After one scrub stage, all gallium measurements were less than the detection limit, which corresponded to DFs >5 x 10{sup 6}. All these values exceed a 10{sup 6} DF needed to meet a hypothetical 10-ppb gallium impurity limit in MOX fuel. The results of this study showed no inherent or fundamental problem with regard to removing gallium from plutonium.

  5. Recycle of scrap plutonium-238 oxide fuel to support future radioisotope applications

    NASA Astrophysics Data System (ADS)

    Schulte, Louis D.; Purdy, Geraldine M.; Jarvinen, Gordon D.; Ramsey, Kevin; Silver, Gary L.; Espinoza, Jacob; Rinehart, Gary H.

    1998-01-01

    The Nuclear Materials Technology (NMT) Division of Los Alamos National Laboratory has initiated a development program to recover & purify plutonium-238 oxide from impure feed sources in a glove box environment. A glove box line has been designed and a chemistry flowsheet developed to perform this recovery task at large scale. The initial demonstration effort focused on purification of 238PuO2 fuel by HNO3/HF dissolution, followed by plutonium(III) oxalate precipitation and calcination to an oxide. Decontamination factors for most impurities of concern in the fuel were very good, producing 238PuO2 fuel significantly better in purity than specified by General Purpose Heat Source (GPHS) fuel powder specifications. A sufficient quantity of purified 238PuO2 fuel was recovered from the process to allow fabrication of a GPHS unit for testing. The results are encouraging for recycle of relatively impure plutonium-238 oxide and scrap residue items into fuel for useful applications. The high specific activity of plutonium-238 magnifies the consequences and concerns of radioactive waste generation. This work places an emphasis on development of waste minimization technologies to complement the aqueous processing operation. Results from experiments on neutralized solutions of plutonium-238 resulted in decontamination to about 1 millicurie/L. Combining ultrafiltration treatment with addition of a water-soluble polymer designed to coordinate Pu, allowed solutions to be decontaminated to about 1 microcurie/L. Efforts continue to develop a capability for efficient, safe, cost-effective, and environmentally acceptable methods to recover and purify 238PuO2 fuel.

  6. Accelerator Mass Spectrometric (AMS) Measurements of Plutonium Activity Concentrations and 240Pu/239Pu Atom Ratios In Soil Extracts Supplied by the Carlsbad Environmental Monitoring & Research Center

    SciTech Connect

    Hamilton, T F; Brown, T A; Marchetti, A A; Martinelli, R E; Kehl, S R

    2005-02-28

    Plutonium-239 ({sup 239}Pu) and plutonium-239+240 ({sup 239+240}Pu) activities concentrations and {sup 240}Pu/{sup 239}Pu atom ratios are reported for a series of chemically purified soil extracts received from the Carlsbad Environmental Monitoring & Research Center (CEMRC) in New Mexico. Samples were analyzed without further purification at the Lawrence Livermore National Laboratory (LLNL) using accelerator mass spectrometry (AMS). This report also includes a brief description of the AMS system and internal laboratory procedures used to ensure the quality and reliability of the measurement data.

  7. Plutonium speciation, solubilization, and migration in soils. 1998 annual progress report

    SciTech Connect

    Haire, R.G.; Neu, M.; Runde, W.

    1998-06-01

    The DOE is currently conducting cleanup activities at its nuclear weapons development sites, many of which have accumulated plutonium in soils for 50 years. To properly control Pu migration in soils within Federal sites and onto public lands, better evaluate the public risk, and design effective remediation strategies, a fundamental understanding of Pu speciation, transport, and release mechanisms is needed. Key scientific goals include: determine Pu concentrations and speciation at a contaminated DOE site; study the formation, stability, and structural and spectroscopic features of environmentally relevant Pu (III, IV, and V) species; determine the mechanism of interaction between Pu and Mn/Fe minerals and the potential release of Pu via redox cycling; and model the environmental behavior of plutonium. This report summarizes work after seven months of a three-year project. In the first year of this project the authors are focusing on the origin, speciation, and mobility of plutonium at the Rocky Flats Environmental Test Site (RFETS).

  8. Excess Weapons Plutonium Immobilization in Russia

    SciTech Connect

    Jardine, L.; Borisov, G.B.

    2000-04-15

    The joint goal of the Russian work is to establish a full-scale plutonium immobilization facility at a Russian industrial site by 2005. To achieve this requires that the necessary engineering and technical basis be developed in these Russian projects and the needed Russian approvals be obtained to conduct industrial-scale immobilization of plutonium-containing materials at a Russian industrial site by the 2005 date. This meeting and future work will provide the basis for joint decisions. Supporting R&D projects are being carried out at Russian Institutes that directly support the technical needs of Russian industrial sites to immobilize plutonium-containing materials. Special R&D on plutonium materials is also being carried out to support excess weapons disposition in Russia and the US, including nonproliferation studies of plutonium recovery from immobilization forms and accelerated radiation damage studies of the US-specified plutonium ceramic for immobilizing plutonium. This intriguing and extraordinary cooperation on certain aspects of the weapons plutonium problem is now progressing well and much work with plutonium has been completed in the past two years. Because much excellent and unique scientific and engineering technical work has now been completed in Russia in many aspects of plutonium immobilization, this meeting in St. Petersburg was both timely and necessary to summarize, review, and discuss these efforts among those who performed the actual work. The results of this meeting will help the US and Russia jointly define the future direction of the Russian plutonium immobilization program, and make it an even stronger and more integrated Russian program. The two objectives for the meeting were to: (1) Bring together the Russian organizations, experts, and managers performing the work into one place for four days to review and discuss their work with each other; and (2) Publish a meeting summary and a proceedings to compile reports of all the excellent

  9. RECOVERY OF PLUTONIUM FROM AQUEOUS SOLUTIONS

    DOEpatents

    Reber, E.J.

    1959-09-01

    A process is described for recovering plutonium values from aqueous solutions by precipitation on bismuth phosphate. The plutonium is secured in its tetravalent state. bismuth salt is added to the solution, and ant excess of phosphoric acid anions is added to the solution in two approximately equal installments. The rate of addition of the first installment is about two to three times as high as the rate of addition of the second installment, whereby a precipitate of bismuth phosphate forms, the precipitate carrying the plutonium values. The precipitate is separated from the solution.

  10. SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS

    DOEpatents

    Nicholls, C.M.; Wells, I.; Spence, R.

    1959-10-13

    The separation of uranium and plutonium from neutronirradiated uranium is described. The neutron-irradiated uranium is dissolved in nitric acid to provide an aqueous solution 3N in nitric acid. The fission products of the solution are extruded by treating the solution with dibutyl carbitol substantially 1.8N in nitric acid. The organic solvent phase is separated and neutralized with ammonium hydroxide and the plutonium reduced with hydroxylamine base to the trivalent state. Treatment of the mixture with saturated ammonium nitrate extracts the reduced plutonium and leaves the uranium in the organic solvent.

  11. Immobilization of excess weapons plutonium in Russia

    SciTech Connect

    Borisov, G B; Jardine, L J; Mansourov, O A

    1999-01-25

    In this paper, we examine the logic and framework for the development of a capability to immobilize excess Russian weapons plutonium by the year 2004. The initial activities underway in Russia, summarized here, include engineering feasibility studies of the immobilization of plutonium-containing materials at the Krasnoyarsk and Mayak industrial sites. In addition, research and development (R&D) studies are underway at Russian institutes to develop glass and ceramic forms suitable for the immobilization of plutonium-containing materials, residues, and wastes and for their geologic disposal.

  12. Removal of plutonium from hepatic tissue

    DOEpatents

    Lindenbaum, Arthur; Rosenthal, Marcia W.

    1979-01-01

    A method is provided for removing plutonium from hepatic tissues by introducing into the body and blood stream a solution of the complexing agent DTPA and an adjunct thereto. The adjunct material induces aberrations in the hepatic tissue cells and removes intracellularly deposited plutonium which is normally unavailable for complexation with the DTPA. Once the intracellularly deposited plutonium has been removed from the cell by action of the adjunct material, it can be complexed with the DTPA present in the blood stream and subsequently removed from the body by normal excretory processes.

  13. HENC performance evaluation and plutonium calibration

    SciTech Connect

    Menlove, H.O.; Baca, J.; Pecos, J.M.; Davidson, D.R.; McElroy, R.D.; Brochu, D.B.

    1997-10-01

    The authors have designed a high-efficiency neutron counter (HENC) to increase the plutonium content in 200-L waste drums. The counter uses totals neutron counting, coincidence counting, and multiplicity counting to determine the plutonium mass. The HENC was developed as part of a Cooperative Research and Development Agreement between the Department of Energy and Canberra Industries. This report presents the results of the detector modifications, the performance tests, the add-a-source calibration, and the plutonium calibration at Los Alamos National Laboratory (TA-35) in 1996.

  14. Plutonium Immobilization Can Loading Conceptual Design

    SciTech Connect

    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.'

  15. NON-CORROSIVE PLUTONIUM FUEL SYSTEMS

    DOEpatents

    Coffinberry, A.S.; Waber, J.T.

    1962-10-23

    An improved plutonium reactor liquid fuel is described for utilization in a nuclear reactor having a tantalum fuel containment vessel. The fuel consists of plutonium and a diluent such as iron, cobalt, nickel, cerium, cerium-- iron, cerium--cobalt, cerium--nickel, and cerium--copper, and an additive of carbon and silicon. The carbon and silicon react with the tantalum container surface to form a coating that is self-healing and prevents the corrosive action of liquid plutonium on the said tantalum container. (AEC)

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

    SciTech Connect

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

    1993-06-01

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

  17. Spectrophotometric determination of plutonium-239 based on the spectrum of plutonium(III) chloride

    SciTech Connect

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

    1994-07-01

    This report describes a spectrophotometric method for determining plutonium-239 (Pu-239) based on the spectrum of Pu(III) chloride. The authors used the sealed-reflux technique for the dissolution of plutonium oxide with hydrochloric acid (HCl) and small amounts of nitric and hydrofluoric acids. To complex the fluoride, they added zirconium, and to reduce plutonium to Pu(III), they added ascorbic acid. They then adjusted the solution to a concentration of 2 M HCl and measured the absorbances at five wavelengths of the Pu(III) chloride spectrum. This spectrophotometric determination can also be applied to samples of plutonium metal dissolved in HCl.

  18. PRODUCTION OF PLUTONIUM FLUORIDE FROM BISMUTH PHOSPHATE PRECIPITATE CONTAINING PLUTONIUM VALUES

    DOEpatents

    Brown, H.S.; Bohlmann, E.G.

    1961-05-01

    A process is given for separating plutonium from fission products present on a bismuth phosphate carrier. The dried carrier is first treated with hydrogen fluoride at between 500 and 600 deg C whereby some fission product fluorides volatilize away from plutonium tetrafluoride, and nonvolatile fission product fluorides are formed then with anhydrous fluorine at between 400 and 500 deg C. Bismuth and plutonium distill in the form of volatile fluorides away from the nonvolatile fission product fluorides. The bismuth and plutonium fluorides are condensed at below 290 deg C.

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

    SciTech Connect

    Chodak, III, Paul

    1996-05-01

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

  20. Opportunities in Plutonium Metallurgical Research

    SciTech Connect

    Schwartz, A J

    2006-12-19

    This is an exciting time to be involved in plutonium metallurgical research. Over the past few years, there have been significant advances in our understanding of the fundamental materials science of this unusual metal, particularly in the areas of self-irradiation induced aging of Pu, the equilibrium phase diagram, the homogenization of {delta}-phase alloys, the crystallography and morphology of the {alpha}{prime}-phase resulting from the isothermal martensitic phase transformation, and the phonon dispersion curves, among many others. In addition, tremendous progress has been made, both experimentally and theoretically, in our understanding of the condensed matter physics and chemistry of the actinides, particularly in the area of electronic structure. Although these communities have made substantial progress, many challenges still remain. This brief overview will address a number of important challenges that we face in fully comprehending the metallurgy of Pu with a specific focus on aging and phase transformations.

  1. PLUTONIUM METALLOGRAPHY AT LOS ALAMOS

    SciTech Connect

    PEREYRA, RAMIRO A.; LOVATO, DARRYL

    2007-01-08

    From early days of the Manhattan program to today, scientists and engineers have continued to investigate the metallurgical properties of plutonium (Pu). Although issues like aging was not a concern to the early pioneers, today the reliability of our aging stockpile is of major focus. And as the country moves toward a new generation of weapons similar problems that the early pioneers faced such as compatibility, homogeneity and malleability have come to the forefront. And metallography will continue to be a principle tool for the resolution of old and new issues. Standard metallographic techniques are used for the preparation of plutonium samples. The samples are first cut with a slow speed idamond saw. After mounting in Epon 815 epoxy resin, the samples are ground through 600 grit silicon carbide paper. PF 5070 (a Freon substitute) is used as a coolant, lubricant, and solvent for most operations. Rough mechanical polished is done with 9-{mu} diamond using a nap less cloth, for example nylon or cotton. Final polish is done with 1-{mu} diamond on a nappy cloth such as sylvet. Ethyl alcohol is then used ultrasonically to clean the samples before electro polishing. The sample is then electro-polished and etched in an electrolyte containing 10% nitric acid, and 90% dimethyleneformalmide. Ethyl alcohol is used as a final cleaning agent. Although standard metallographic preparation techniques are used, there are several reasons why metallography of Pu is difficult and challenging. Firstly, because of the health hazards associated with its radioactive properties, sample preparation is conducted in glove boxes. Figure 1 shows the metallography line, in an R and D facility. Since they are designed to be negative in pressure to the laboratory, cross-contamination of abrasives is a major problem. In addition, because of safety concerns and waste issues, there is a limit to the amount of solvent that can be used. Secondly, Pu will readily hydride or oxidize when in contact

  2. A Plutonium Storage Container Pressure Measurement Technique

    SciTech Connect

    Grim, T.J.

    2002-05-10

    Plutonium oxide and metal awaiting final disposition are currently stored at the Savannah River Site in crimp sealed food pack cans. Surveillances to ensure continued safe storage of the cans include periodic lid deflection measurements using a mechanical device.

  3. SEPARATION OF URANIUM, PLUTONIUM, AND FISSION PRODUCTS

    DOEpatents

    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.

  4. IMPROVED PROCESS OF PLUTONIUM CARRIER PRECIPITATION

    DOEpatents

    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.

  5. Recent plutonium metal production experience at Hanford

    SciTech Connect

    Gibson, M.W.; Nyman, D.H. )

    1989-11-01

    Plutonium metal is produced at the Hanford site in the remote mechanical C (RMC) line. The line is housed in the plutonium finishing plant (PFP). The PFP is operated by the Westinghouse Hanford Company for the U.S. Department of Energy. The RMC line was built in the early 1960s and operated until 1973 when it was shut down. The line was restarted in 1985 and has operated on a campaign basis since that time. The RMC line converts plutonium nitrate solution to plutonium metal in the classic precipitation/calcination/fluorination/reduction process. The operations are contained in glove boxes with a dry air atmosphere. Most of the process is remotely controlled from a central control room. Numerous process improvements were made in the line before initiating operations in 1985 and in 1988. These changes, in conjunction with improved conduct of operations, have resulted in improved yields.

  6. Pulmonary carcinogenesis from plutonium-containing particles

    SciTech Connect

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

    1980-01-01

    Plutonium administered as an alpha radiation source to the respiratory tracts of Syrian hamsters has resulted in various incidences of neoplasia. Adenomas are the primary lung tumor observed, but adenocarcinomas are also prevalent.

  7. Design and evaluation of plutonium electrorefining cells

    SciTech Connect

    Not Available

    1987-01-01

    A plutonium electrorefining cell was designed for stationary furnace operation. This cell and the LANL electrorefining cell were evaluated. Results of this evaluation and comparison to existing production electrorefining at Rocky Flats are presented.

  8. Plutonium finishing plant dangerous waste training plan

    SciTech Connect

    ENTROP, G.E.

    1999-05-24

    This training plan describes general requirements, worker categories, and provides course descriptions for operation of the Plutonium Finish Plant (PFP) waste generation facilities, permitted treatment, storage and disposal (TSD) units, and the 90-Day Accumulation Areas.

  9. Leaching behavior of particulate plutonium oxide

    SciTech Connect

    Kosiewicz, S.T.; Heaton, R.C.

    1985-08-01

    Different size cuts of /sup 238/PuO/sub 2/ particles were mixed with deionized water at two temperatures in a shaker bath. The gross plutonium concentration in the water was measured, as well as that portion of the plutonium retained on a 0.1-..mu..m pore filter. The concentration of the plutonium released was primarily a function of the surface area of the particles. The release rate of plutonium into the water for the size cut with particles having diameters between 30 and 20 ..mu..m was 3 ng/m/sup 2//s; this rate is within the range observed in past experiments involving aquatic environments. The amount of material retained by the 0.1-..mu..m filters decreased with increasing time, suggesting that size reduction or removal processes occurred. 6 refs., 3 figs., 9 tabs.

  10. Plutonium focus area: Technology summary

    SciTech Connect

    1996-03-01

    To ensure research and development programs focus on the most pressing environmental restoration and waste management problems at the U.S. Department of Energy (DOE), the Assistant Secretary for the Office of Environmental Management (EM) established a working group in August 1993 to implement a new approach to research and technology development. As part of this approach, EM developed a management structure and principles that led to creation of specific focus areas. These organizations were designed to focus scientific and technical talent throughout DOE and the national scientific community on major environmental restoration and waste management problems facing DOE. The focus area approach provides the framework for inter-site cooperation and leveraging of resources on common problems. After the original establishment of five major focus areas within the Office of Technology Development (EM-50), the Nuclear Materials Stabilization Task Group (NMSTG, EM-66) followed EM-50`s structure and chartered the Plutonium Focus Area (PFA). NMSTG`s charter to the PFA, described in detail later in this book, plays a major role in meeting the EM-66 commitments to the Defense Nuclear Facilities Safety Board (DNFSB). The PFA is a new program for FY96 and as such, the primary focus of revision 0 of this Technology Summary is an introduction to the Focus Area; its history, development, and management structure, including summaries of selected technologies being developed. Revision 1 to the Plutonium Focus Area Technology Summary is slated to include details on all technologies being developed, and is currently planned for release in August 1996. The following report outlines the scope and mission of the Office of Environmental Management, EM-60, and EM-66 organizations as related to the PFA organizational structure.

  11. PLUTONIUM CARRIER METATHESIS WITH ORGANIC REAGENT

    DOEpatents

    Thompson, S.G.

    1958-07-01

    A method is described for converting a plutonium containing bismuth phosphate carrier precipitate Into a compositton more readily soluble in acid. The method consists of dissolving the bismuth phosphate precipitate in an aqueous solution of alkali metal hydroxide, and adding one of a certaia group of organic compounds, e.g., polyhydric alcohols or a-hydrorycarboxylic acids. The mixture is then heated causiing formation of a bismuth hydroxide precipitate containing plutonium which may be readily dissolved in nitric acid for further processing.

  12. Plutonium: The first 50 years. United States plutonium production, acquisition, and utilization from 1944 through 1994

    SciTech Connect

    1996-02-01

    The report contains important newly declassified information regarding the US production, acquisition, and removals of plutonium. This new information, when combined with previously declassified data, has allowed the DOE to issue, for the first time, a truly comprehensive report on the total DOE plutonium inventory. At the December 7, 1993, Openness Press Conference, the DOE declassified the plutonium inventories at eight locations totaling 33.5 metric tons (MT). This report declassifies the remainder of the DOE plutonium inventory. Newly declassified in this report is the quantity of plutonium at the Pantex Site, near Amarillo, Texas, and in the US nuclear weapons stockpile of 66.1 MT, which, when added to the previously released inventory of 33.5 MT, yields a total plutonium inventory of 99.5 MT. This report will document the sources which built up the plutonium inventory as well as the transactions which have removed plutonium from that inventory. This report identifies four sources that add plutonium to the DOE/DoD inventory, and seven types of transactions which remove plutonium from the DOE/DoD inventory. This report also discusses the nuclear material control and accountability system which records all nuclear material transactions, compares records with inventory and calculates material balances, and analyzes differences to verify that nuclear materials are in quantities as reported. The DOE believes that this report will aid in discussions in plutonium storage, safety, and security with stakeholders as well as encourage other nations to declassify and release similar data. These data will also be available for formulating policies with respect to disposition of excess nuclear materials. The information in this report is based on the evaluation of available records. The information contained in this report may be updated or revised in the future should additional or more detailed data become available.

  13. A DGT technique for plutonium bioavailability measurements.

    PubMed

    Cusnir, Ruslan; Steinmann, Philipp; Bochud, François; Froidevaux, Pascal

    2014-09-16

    The toxicity of heavy metals in natural waters is strongly dependent on the local chemical environment. Assessing the bioavailability of radionuclides predicts the toxic effects to aquatic biota. The technique of diffusive gradients in thin films (DGT) is largely exploited for bioavailability measurements of trace metals in waters. However, it has not been applied for plutonium speciation measurements yet. This study investigates the use of DGT technique for plutonium bioavailability measurements in chemically different environments. We used a diffusion cell to determine the diffusion coefficients (D) of plutonium in polyacrylamide (PAM) gel and found D in the range of 2.06-2.29 × 10(-6) cm(2) s(-1). It ranged between 1.10 and 2.03 × 10(-6) cm(2) s(-1) in the presence of fulvic acid and in natural waters with low DOM. In the presence of 20 ppm of humic acid of an organic-rich soil, plutonium diffusion was hindered by a factor of 5, with a diffusion coefficient of 0.50 × 10(-6) cm(2) s(-1). We also tested commercially available DGT devices with Chelex resin for plutonium bioavailability measurements in laboratory conditions and the diffusion coefficients agreed with those from the diffusion cell experiments. These findings show that the DGT methodology can be used to investigate the bioaccumulation of the labile plutonium fraction in aquatic biota.

  14. Disposal of Surplus Weapons Grade Plutonium

    SciTech Connect

    H. Alsaed; P. Gottlieb

    2000-09-12

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

  15. Complementary technologies for verification of excess plutonium

    SciTech Connect

    Langner, , D.G.; Nicholas, N.J.; Ensslin, N.; Fearey, B.L.; Mitchell, D.J.; Marlow, K.W.; Luke, S.J.; Gosnell, T.B.

    1998-12-31

    Three complementary measurement technologies have been identified as candidates for use in the verification of excess plutonium of weapons origin. These technologies: high-resolution gamma-ray spectroscopy, neutron multiplicity counting, and low-resolution gamma-ray spectroscopy, are mature, robust technologies. The high-resolution gamma-ray system, Pu-600, uses the 630--670 keV region of the emitted gamma-ray spectrum to determine the ratio of {sup 240}Pu to {sup 239}Pu. It is useful in verifying the presence of plutonium and the presence of weapons-grade plutonium. Neutron multiplicity counting is well suited for verifying that the plutonium is of a safeguardable quantity and is weapons-quality material, as opposed to residue or waste. In addition, multiplicity counting can independently verify the presence of plutonium by virtue of a measured neutron self-multiplication and can detect the presence of non-plutonium neutron sources. The low-resolution gamma-ray spectroscopic technique is a template method that can provide continuity of knowledge that an item that enters the a verification regime remains under the regime. In the initial verification of an item, multiple regions of the measured low-resolution spectrum form a unique, gamma-radiation-based template for the item that can be used for comparison in subsequent verifications. In this paper the authors discuss these technologies as they relate to the different attributes that could be used in a verification regime.

  16. Recovery of plutonium from nitric acid waste

    SciTech Connect

    Muscatello, A.C.; Saba, M.T.; Navratil, J.D.

    1986-12-21

    Seven candidate materials, each contained in a static-bed column, have been evaluated for removing plutonium from nitric acid waste. Three materials have insufficient capacity for plutonium: TBP (tri-n-butylphosphate) sorbed on Amberlite XAD-4 resin, O phi D(IB)CMPO (octylphenyl-N, N-diisobutylcarbamoylmethylphosphine oxide) sorbed on XAD-4, and Amberlite IRA-938 anion exchange resin. The remaining four materials reduced 10/sup -3/ g/l plutonium in 7.2M HNO/sub 3/ to low 10/sup -5/ g/l for 80 bed volumes (BV). The 10% breakthrough capacities for 3 x 10/sup -2/ g/l plutonium are: TOPO (tri-n-octylphosphine oxide) on XAD-4 - 1800 BV, DHDECMP (dihexyl-N, N-diethylcarbamoylmethylphosphonate) on XAD-4 - 960 BV, Dowex 1 x 4 - 840 BV, and DHDECMP + TBP - 640 BV. Based on these results and generally poor elution of all materials, TOPO on XAD-4 is recommended as the best candidate for recovery of plutonium followed by acid digestion or combustion of the TOPO to recover the concentrated plutonium.

  17. Structural and functional characteristics of virgin and fouled Protein A MabSelect resin cycled in a monoclonal antibody purification process.

    PubMed

    Zhang, Shaojie; Xu, Kerui; Daniels, William; Salm, Jeffrey; Glynn, Judy; Martin, Joseph; Gallo, Christopher; Godavarti, Ranga; Carta, Giorgio

    2016-02-01

    The structural and functional characteristics of the Protein A MabSelect resin are determined for a virgin sample and for samples removed from a column that had been operated in an antibody capture process which had shown losses in product recovery over fewer than 20 cycles. Compared to the virgin resin, the cycled samples show reduced porosity and apparent pore size based on inverse size exclusion chromatography while transmission electron microscopy (TEM) shows accumulation of foulants on the cycled resin. Adsorption isotherms, batch adsorption kinetics, and batch desorption kinetics, obtained using the antibody in purified form, show that the cycled samples have about 10% lower binding capacity and slower mass transfer. Confocal scanning laser microscopy shows, however, that different degrees of fouling exist for different beads in the cycled samples, which may correspond to the existence of areas exposed to minimal or no flow in the process column. Replacing the standard cleaning procedure with an improved multi-step cleaning protocol prevented the accumulation of foulants in the resin beads, as evident from TEM, and resulted in a stable operation with high recovery.

  18. PROCESS OF FORMING PLUOTONIUM SALTS FROM PLUTONIUM EXALATES

    DOEpatents

    Garner, C.S.

    1959-02-24

    A process is presented for converting plutonium oxalate to other plutonium compounds by a dry conversion method. According to the process, lower valence plutonium oxalate is heated in the presence of a vapor of a volatile non- oxygenated monobasic acid, such as HCl or HF. For example, in order to produce plutonium chloride, the pure plutonium oxalate is heated to about 700 deg C in a slow stream of hydrogen plus HCl. By the proper selection of an oxidizing or reducing atmosphere, the plutonium halide product can be obtained in either the plus 3 or plus 4 valence state.

  19. SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS

    DOEpatents

    Boyd, G.E.; Adamson, A.W.; Schubert, J.; Russell, E.R.

    1958-10-01

    A chromatographic adsorption process is presented for the separation of plutonium from other fission products formed by the irradiation of uranium. The plutonium and the lighter element fission products are adsorbed on a sulfonated phenol-formaldehyde resin bed from a nitric acid solution containing the dissolved uranium. Successive washes of sulfuric, phosphoric, and nitric acids remove the bulk of the fission products, then an eluate of dilute phosphoric and nitric acids removes the remaining plutonium and fission products. The plutonium is selectively removed by passing this solution through zirconium phosphate, from which the plutonium is dissolved with nitric acid. This process provides a convenient and efficient means for isolating plutonium.

  20. Determination of origin and intended use of plutonium metal using nuclear forensic techniques

    DOE PAGES

    Rim, Jung H.; Kuhn, Kevin J.; Tandon, Lav; ...

    2017-04-01

    Nuclear forensics techniques, including micro-XRF, gamma spectrometry, trace elemental analysis and isotopic/chronometric characterization were used to interrogate two, potentially related plutonium metal foils. These samples were submitted for analysis with only limited production information, and a comprehensive suite of forensic analyses were performed. Resulting analytical data was paired with available reactor model and historical information to provide insight into the materials’ properties, origins, and likely intended uses. Both were super-grade plutonium, containing less than 3% 240Pu, and age-dating suggested that most recent chemical purification occurred in 1948 and 1955 for the respective metals. Additional consideration of reactor modelling feedback andmore » trace elemental observables indicate plausible U.S. reactor origin associated with the Hanford site production efforts. In conclusion, based on this investigation, the most likely intended use for these plutonium foils was 239Pu fission foil targets for physics experiments, such as cross-section measurements, etc.« less

  1. Environmental Management Science Program Report Progress Report Plutonium Speciation, Solubilization, and Migration in Soils

    SciTech Connect

    Haire, Richard G.

    2000-06-01

    The DOE is currently conducting cleanup activities at its nuclear weapons development sites, many of which have accumulated plutonium in soils for 50 years. To properly control Pu migration in soils within Federal sites and onto public lands, better evaluate the public risk, and design effective remediation strategies, a fundamental understanding of Pu speciation and environmental transport is needed. The key scientific goals of this project are: to determine Pu concentrations and speciation at a contaminated DOE site; to study the formation, stability, and structural and spectroscopic features of environmentally relevant Pu species; to determine the mechanism(s) of interaction between Pu and Mn/Fe minerals and the potential release of Pu via redox cycling; and to model the environmental behavior of plutonium. Our long-term goal is to use characterization, thermodynamic, mineral interaction, and mobility data to develop better models of radionuclide transport and risk assessment, and to enable the development of science based decontamination strategies. This research will fill important gaps between basic actinide science and the problems impeding site clean-up, plutonium disposition, and accurate risk assessment. Information gained will allow for the development of technologies and clean-up approaches targeting particular plutonium contaminants and improved assessment of risks associated with actinide migration, site remediation, and decontamination. By combining very specific study of plutonium at the Rocky Flats Environmental Technology Site (RFETS), a well-characterized contaminated site, with laboratory studies on the most important plutonium and mineral component systems, we will provide essential knowledge of contaminant characteristics and distinguish critical geochemical processes and mechanisms.

  2. The thermodynamics of pyrochemical processes for liquid metal reactor fuel cycles

    SciTech Connect

    Johnson, I.

    1987-01-01

    The thermodynamic basis for pyrochemical processes for the recovery and purification of fuel for the liquid metal reactor fuel cycle is described. These processes involve the transport of the uranium and plutonium from one liquid alloy to another through a molten salt. The processes discussed use liquid alloys of cadmium, zinc, and magnesium and molten chloride salts. The oxidation-reduction steps are done either chemically by the use of an auxiliary redox couple or electrochemically by the use of an external electrical supply. The same basic thermodynamics apply to both the salt transport and the electrotransport processes. Large deviations from ideal solution behavior of the actinides and lanthanides in the liquid alloys have a major influence on the solubilities and the performance of both the salt transport and electrotransport processes. Separation of plutonium and uranium from each other and decontamination from the more noble fission product elements can be achieved using both transport processes. The thermodynamic analysis is used to make process design computations for different process conditions.

  3. 16. VIEW OF GLOVE BOX WORKSTATIONS WITHIN THE PLUTONIUM BUTTON ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    16. VIEW OF GLOVE BOX WORKSTATIONS WITHIN THE PLUTONIUM BUTTON BREAKOUT ROOM. (9/82) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

  4. 69. INTERIOR, BUILDING 272 (PLUTONIUM STORAGE BUILDING) LOOKING SOUTHWEST THROUGH ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    69. INTERIOR, BUILDING 272 (PLUTONIUM STORAGE BUILDING) LOOKING SOUTHWEST THROUGH DOOR-WAY INTO PLUTONIUM STORAGE AREA. - Loring Air Force Base, Weapons Storage Area, Northeastern corner of base at northern end of Maine Road, Limestone, Aroostook County, ME

  5. 71. INTERIOR, BUILDING 272 (PLUTONIUM STORAGE BUILDING) LOOKING NORTHEAST INTO ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    71. INTERIOR, BUILDING 272 (PLUTONIUM STORAGE BUILDING) LOOKING NORTHEAST INTO PLUTONIUM STORAGE ROOM SHOWING CUBICLES FOR STORAGE. - Loring Air Force Base, Weapons Storage Area, Northeastern corner of base at northern end of Maine Road, Limestone, Aroostook County, ME

  6. 17. VIEW OF THE FIRST PLUTONIUM BUTTON PRODUCED FROM THE ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    17. VIEW OF THE FIRST PLUTONIUM BUTTON PRODUCED FROM THE BUILDING 371 AQUEOUS RECOVERY OPERATION. (9/30/83) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

  7. Magnetic excitations in plutonium monoantimonide

    SciTech Connect

    Lander, G.H.; Stirling, W.G.; Mignod, J.R.; Spirlet, J.C.; Rebezant, J.; Vogt, O.

    1985-01-01

    Neutron inelastic scattering studies of uranium compounds have illustrated the complexity of the interactions in this part of periodic table. Recently, large crystals of plutonium compounds have been grown at the Transuranium Institute, and we report here the first neutron inelastic scattering on PuSb. A monodomain sample of ferromagnetic PuSb was prepared by cooling in a field with H vector parallel to (001). We observe an almost dispersionless magnetic transition, assigned to arise within the GAMMA/sub 8/ ground state, with an energy of 4.3 THz. Surprisingly, the lowest frequency (3.5 THz) excitation occurs at the zone boundary, with a polarization along the wavevector. Theoretical predictions give the spin-wave gap and GAMMA/sub 8/-GAMMA/sub 7/ crystal field splitting (> 12 THz experimentally), but are presently unable to explain the lifting of the degeneracy of the longitudinal and transverse modes and the minimum of the former at the zone boundary. 7 refs., 3 figs.

  8. Hydrogen gas purification apparatus

    SciTech Connect

    Yanagihara, N.; Gamo, T.; Iwaki, T.; Moriwaki, Y.

    1984-04-24

    A hydrogen gas purification apparatus which includes at least one set of two hydrogen purification containers coupled to each other for heat exchanging therebetween, each of the hydrogen purification containers containing a hydrogen absorbing alloy. The hydrogen gas purification apparatus is so arranged as to cause hydrogen gas to be selectively desorbed from and absorbed into the hydrogen absorbing alloy by the amount of heat produced when the hydrogen gas is selectively absorbed into and desorbed from the hydrogen absorbing alloy.

  9. Exhaust gas purification device

    SciTech Connect

    Fujiwara, H.; Hibi, T.; Sayo, S.; Sugiura, Y.; Ueda, K.

    1980-02-19

    The exhaust gas purification device includes an exhaust manifold , a purification cylinder connected with the exhaust manifold through a first honey-comb shaped catalyst, and a second honeycomb shaped catalyst positioned at the rear portion of the purification cylinder. Each catalyst is supported by steel wool rings including coarse and dense portions of steel wool. The purification device further includes a secondary air supplying arrangement.

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

    NASA Astrophysics Data System (ADS)

    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. PROCESS OF ELIMINATING HYDROGEN PEROXIDE IN SOLUTIONS CONTAINING PLUTONIUM VALUES

    DOEpatents

    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.

  12. COLUMBIC OXIDE ADSORPTION PROCESS FOR SEPARATING URANIUM AND PLUTONIUM IONS

    DOEpatents

    Beaton, R.H.

    1959-07-14

    A process is described for separating plutonium ions from a solution of neutron irradiated uranium in which columbic oxide is used as an adsorbert. According to the invention the plutonium ion is selectively adsorbed by Passing a solution containing the plutonium in a valence state not higher than 4 through a porous bed or column of granules of hydrated columbic oxide. The adsorbed plutonium is then desorbed by elution with 3 N nitric acid.

  13. Dispersion of plutonium from contaminated pond sediments

    USGS Publications Warehouse

    Rees, T.F.; Cleveland, J.M.; Carl, Gottschall W.

    1978-01-01

    Sediment-water distributions of plutonium as a function of pH and contact time are investigated in a holding pond at the Rocky Flats plant of the Department of Energy. Although plutonium has been shown to sorb from natural waters onto sediments, the results of this study indicate that under the proper conditions it can be redispersed at pH 9 and above. Concentrations greater than 900 pCi Pu/L result after 34 h contact at pH 11 or 12 and the distribution coefficient, defined as the ratio of concentration in the sediment to that in the liquid, decreases from 1.1 ?? 105 at pH 7 to 1.2 ?? 103 at pH 11. The plutonium is probably dispersed as discrete colloids or as hydrolytic species adsorbed onto colloidal sediment particles whose average size decreases with increasing pH above pH 9. About 5% of the total plutonium is dispersed at pH 12, and the dispersion seems to readsorb on the sediment with time. Consequently, migration of plutonium from the pond should be slow, and it would be difficult to remove this element completely from pond sediment by leaching with high pH solutions. ?? 1978 American Chemical Society.

  14. Plutonium Chemistry in the UREX+ Separation Processes

    SciTech Connect

    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.

  15. Lindbladian purification

    NASA Astrophysics Data System (ADS)

    Arenz, Christian; Burgarth, Daniel; Giovannetti, Vittorio; Nakazato, Hiromichi; Yuasa, Kazuya

    2017-06-01

    In a recent work (Burgarth et al 2014, Nat. Commun. 5 5173), it was shown that a series of frequent measurements can project the dynamics of a quantum system onto a subspace in which the dynamics can be more complex. In this subspace, even full controllability can be achieved, although the controllability over the system before the projection is very poor since the control Hamiltonians commute with each other. We can also think of the opposite: any Hamiltonians of a quantum system, which are in general noncommutative with each other, can be made commutative by embedding them in an extended Hilbert space, thus the dynamics in the extended space becomes trivial and simple. This idea of making noncommutative Hamiltonians commutative is called ‘Hamiltonian purification.’ The original noncommutative Hamiltonians are recovered by projecting the system back onto the original Hilbert space through frequent measurements. Here, we generalise this idea to open-system dynamics by presenting a simple construction to make Lindbladians, as well as Hamiltonians, commutative on a larger space with an auxiliary system. We show that the original dynamics can be recovered through frequently measuring the auxiliary system in a non-selective way. Moreover, we provide a universal pair of Lindbladians that describe an ‘accessible’ open quantum system for generic system sizes. This allows us to conclude that through a series of frequent non-selective measurements a nonaccessible open quantum system generally becomes accessible. This sheds further light on the role of measurement backaction on the control of quantum systems.

  16. Microwave calcination for plutonium immobilization and residue stabilization

    SciTech Connect

    Harris, M.J.; Rising, T.L.; Roushey, W.J.; Sprenger, G.S.

    1995-12-01

    In the late 1980`s development was begun on a process using microwave energy to vitrify low level mixed waste sludge and transuranic mixed waste sludge generated in Building 374 at Rocky Flats. This process was shown to produce a dense, highly durable waste form. With the cessation of weapons production at Rocky Flats, the emphasis has changed from treatment of low level and TRU wastes to stabilizaiton of plutonium oxide and residues. This equipment is versatile and can be used as a heat source to calcine, react or vitrify many types of residues and oxides. It has natural economies in that it heats only the material to be treated, significantly reducing cycle times over conventional furnaces. It is inexpensive to operate in that most of the working components remain outside of any necessary contamination enclosure and therefore can easily be maintained. Limited testing has been successfully performed on cerium oxide (as a surrogate for plutonium oxide), surrogate electrorefining salts, surrogate residue sludge and residue ash. Future plans also include tests on ion exchange resins. In an attempt to further the usefullness of this technology, a mobile, self-contained microwave melting system is currently under development and expected to be operational at Rocky Flats Enviromental Technology Site by the 4th quarter of FY96.

  17. Hematological effects of inhaled plutonium dioxide in beagles

    SciTech Connect

    Weller, R.E.; Buschbom, R.L.; Park, J.F.

    1995-07-01

    A life-span study indicated that plutonium activity in the thoracic lymph nodes is a contributor to development of lymphopenia in beagles exposed to {sup 239}PuO{sub 2}. Significant lymphopenia was found in 67 (58%) beagles given a single nose-only exposure to {sup 239}PuO{sub 2} to result in mean initial lung depositions ranging from 0.69 to 213.3 kBq. Lymphoid atrophy and sclerosis of the thoracic lymph nodes and lymphopenia were observed in exposure-level groups with initial lung depositions {ge}2.5 kBq. Those dogs with final plutonium concentrations in the thoracic lymph nodes {ge}0.4 kBq/g and dose rates {ge}0.01 Gy/day developed lymphopenia. Marked differences existed between chronically lymphopenic dogs and intermittently lymphopenic dogs with regard to initial lung deposition, time to lymphopenic events and absolute lymphocyte concentrations. Linear regression analysis revealed moderate correlation between reduction in lymphocyte values and initial lung deposition, in both magnitude and time of appearance after exposure. Cumulative dose and dose rate appeared to act together to produce initial effects on lymphocyte populations, while dose rate alone appeared to be responsible for the maintenance and subsequent cycles of lymphopenia seen over the life span. No primary tumors were associated with the thoracic lymph nodes in this study, although 70% of the lymphopenic dogs developed lung tumors. 28 refs., 2 figs., 7 tabs.

  18. The biodistribution and toxicity of plutonium, americium and neptunium.

    PubMed

    Taylor, D M

    1989-07-15

    In the nuclear fuel cycle the transuranic radionuclides plutonium-239, americium-241 and neptunium-237 would probably present the most serious hazard to human health if released into the environment. Despite differences in their solution chemistry the three elements exhibit remarkable similarity in their biochemical behaviour, apparently sharing similar transport pathways in blood and cells. After entering the blood the elements deposit predominantly in liver and skeleton, where retention appears to be prolonged, with half-times of the order of years. The principal late effects of all three radionuclides are the induction of cancers of bone, lung or liver. For the latter tumours the induction risk per unit radiation dose appears similar for the three radionuclides. But in bone there are indications that, due to microscopic differences in the distribution of the alpha-particle radiation dose, the efficiency of bone cancer induction may increase in the order americium-241 less than plutonium-239 less than neptunium-237. No case of human cancer induced by these radionuclides is known.

  19. 15. VIEW OF THE SAFE GEOMETRY PLUTONIUM METAL STORAGE PALLETS ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    15. VIEW OF THE SAFE GEOMETRY PLUTONIUM METAL STORAGE PALLETS FROM THE INSIDE OF AN INPUT-OUTPUT STATION. INDIVIDUAL CONTAINERS OF PLUTONIUM ARE STORED IN THE WATER-FILLED, DOUBLE-WALLED STAINLESS STEEL TUBES THAT ARE WELDED ONTO THE PALLETS. (12/3/88) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

  20. 10 CFR 71.88 - Air transport of plutonium.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Air transport of plutonium. 71.88 Section 71.88 Energy... Controls and Procedures § 71.88 Air transport of plutonium. (a) Notwithstanding the provisions of any... citation of 49 CFR chapter I, as may be applicable, the licensee shall assure that plutonium in any...

  1. 10 CFR 71.63 - Special requirement for plutonium shipments.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Special requirement for plutonium shipments. 71.63 Section... MATERIAL Package Approval Standards § 71.63 Special requirement for plutonium shipments. Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than...

  2. VIEW OF THE INTERIOR OF THE PLUTONIUM LABORATORY IN BUILDING ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    VIEW OF THE INTERIOR OF THE PLUTONIUM LABORATORY IN BUILDING 559. THE LABORATORY WAS USED TO ANALYZE THE PURITY OF PLUTONIUM. PLUTONIUM SAMPLES WERE CONTAINED WITHIN GLOVE BOXES - Rocky Flats Plant, Chemical Analytical Laboratory, North-central section of Plant, Golden, Jefferson County, CO

  3. 10 CFR 71.63 - Special requirement for plutonium shipments.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Special requirement for plutonium shipments. 71.63 Section... MATERIAL Package Approval Standards § 71.63 Special requirement for plutonium shipments. Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than...

  4. 10 CFR 71.63 - Special requirement for plutonium shipments.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Special requirement for plutonium shipments. 71.63 Section... MATERIAL Package Approval Standards § 71.63 Special requirement for plutonium shipments. Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than...

  5. 10 CFR 71.63 - Special requirement for plutonium shipments.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Special requirement for plutonium shipments. 71.63 Section... MATERIAL Package Approval Standards § 71.63 Special requirement for plutonium shipments. Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than...

  6. 10 CFR 71.63 - Special requirement for plutonium shipments.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Special requirement for plutonium shipments. 71.63 Section... MATERIAL Package Approval Standards § 71.63 Special requirement for plutonium shipments. Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than...

  7. 10 CFR 71.88 - Air transport of plutonium.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Air transport of plutonium. 71.88 Section 71.88 Energy... Controls and Procedures § 71.88 Air transport of plutonium. (a) Notwithstanding the provisions of any... citation of 49 CFR chapter I, as may be applicable, the licensee shall assure that plutonium in any...

  8. A particulate isotopic standard of plutonium in an aluminosilicate matrix

    SciTech Connect

    Stoffels, J.J.; Cannon, W.C.; Robertson, D.M. )

    1991-01-01

    Plutonium isotopic microstandard particles have been produced for mass spectrometer calibration. The particles may also be useful as an elemental standard for calibration of electron and ion microprobe instruments. The standard consists of spherical, micrometer-size aluminosilicate particles loaded with plutonium of known isotopic distribution. The morphology, elemental composition, and plutonium isotopic composition of the particles have been characterized.

  9. OVERVIEW OF CALORIMETRIC ASSAY OF PLUTONIUM IN THE UNITED STATES

    SciTech Connect

    Rudy, C. R.

    2001-01-01

    Calorimetry is a primary measurement technique for assay of quantities of plutonium in the United States. It is the most accurate NDA technique for many forms of plutonium-bearing materials. This paper provides an overview of the use of calorimetry in combination with high-resolution gamma-ray spectroscopy for accurate plutonium mass determinations.

  10. Treatment of accidental intakes of plutonium and americium: guidance notes.

    PubMed

    Ménétrier, F; Grappin, L; Raynaud, P; Courtay, C; Wood, R; Joussineau, S; List, V; Stradling, G N; Taylor, D M; Bérard, Ph; Morcillo, M A; Rencova, J

    2005-06-01

    The scientific basis for the treatment of the contamination of the human body by plutonium, americium and other actinides is reviewed. Guidance Notes are presented for the assistance of physicians and others who may be called upon to treat workers or members of the public who may become contaminated internally with inhaled plutonium nitrate, plutonium tributyl phosphate, americium nitrate or americium oxide.

  11. Sorption/Desorption Interactions of Plutonium with Montmorillonite

    NASA Astrophysics Data System (ADS)

    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

  12. ESTIMATING IMPURITIES IN SURPLUS PLUTONIUM FOR DISPOSITION

    SciTech Connect

    Allender, J.; Moore, E.

    2013-07-17

    The United States holds at least 61.5 metric tons (MT) of plutonium that is permanently excess to use in nuclear weapons programs, including 47.2 MT of weapons-grade plutonium. Surplus inventories will be stored safely by the Department of Energy (DOE) and then transferred to facilities that will prepare the plutonium for permanent disposition. The Savannah River National Laboratory (SRNL) operates a Feed Characterization program for the Office of Fissile Materials Disposition of the National Nuclear Security Administration and the DOE Office of Environmental Management. Many of the items that require disposition are only partially characterized, and SRNL uses a variety of techniques to predict the isotopic and chemical properties that are important for processing through the Mixed Oxide Fuel Fabrication Facility and alternative disposition paths. Recent advances in laboratory tools, including Prompt Gamma Analysis and Peroxide Fusion treatment, provide data on the existing inventories that will enable disposition without additional, costly sampling and destructive analysis.

  13. Characterization of Delta Phase Plutonium Metal

    SciTech Connect

    Rudisill, T.S.

    2000-09-21

    The FB-Line facility has developed the capability to recast plutonium metal using an M-18 reduction furnace with a new casting chamber. Plutonium metal is recast by charging a standard FB-Line magnesia crucible and placing the charge in the casting chamber. The loaded casting chamber is raised into the M-18 reduction furnace and sealed against the furnace head using a copper gasket following the same procedure used for a bomb reduction run. The interior volume of the chamber is evacuated and backfilled with argon gas. The M-18 motor-generator set is used to heat the surface of the casting chamber to nominally 750 Degrees C. Within about 2 hr, the plutonium metal reaches its melting temperature of approximately 640 Degrees C.

  14. Excess plutonium disposition using ALWR technology

    SciTech Connect

    Phillips, A.; Buckner, M.R.; Radder, J.A.; Angelos, J.G.; Inhaber, H.

    1993-02-01

    The Office of Nuclear Energy of the Department of Energy chartered the Plutonium Disposition Task Force in August 1992. The Task Force was created to assess the range of practicable means of disposition of excess weapons-grade plutonium. Within the Task Force, working groups were formed to consider: (1) storage, (2) disposal,and(3) fission options for this disposition,and a separate group to evaluate nonproliferation concerns of each of the alternatives. As a member of the Fission Working Group, the Savannah River Technology Center acted as a sponsor for light water reactor (LWR) technology. The information contained in this report details the submittal that was made to the Fission Working Group of the technical assessment of LWR technology for plutonium disposition. The following aspects were considered: (1) proliferation issues, (2) technical feasibility, (3) technical availability, (4) economics, (5) regulatory issues, and (6) political acceptance.

  15. Method for dissolving delta-phase plutonium

    DOEpatents

    Karraker, David G.

    1992-01-01

    A process for dissolving plutonium, and in particular, delta-phase plutonium. The process includes heating a mixture of nitric acid, hydroxylammonium nitrate (HAN) and potassium fluoride to a temperature between 40.degree. and 70.degree. C., then immersing the metal in the mixture. Preferably, the nitric acid has a concentration of not more than 2M, the HAN approximately 0.66M, and the potassium fluoride 0.1M. Additionally, a small amount of sulfamic acid, such as 0.1M can be added to assure stability of the HAN in the presence of nitric acid. The oxide layer that forms on plutonium metal may be removed with a non-oxidizing acid as a pre-treatment step.

  16. The United States Plutonium Balance, 1944 - 2009

    SciTech Connect

    2012-06-01

    This report updates the report -Plutonium: The first 50 years- which was released by the U.S.Department of Energy (DOE) in 1996. The topic of both reports is plutonium, sometimes referred to as Pu-239, which is capable of sustaining a nuclear chain reaction and is used in nuclear weapons and for nuclear power production. This report updates 1994 data through 2009. The four most significant changes since 1994 include: (a) the completion of cleanup activities at the Rocky Flats Plant in 2005; (b) material consolidation and disposition activities, especially shipments from Hanford to the Savannah River Site; (c) the 2007 declaration of an additional 9.0 MT of weapons grade plutonium to be surplus to defense needs in the coming decades; and (d) the opening of the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico in 1999.

  17. Excess plutonium disposition using ALWR technology

    SciTech Connect

    Phillips, A.; Buckner, M.R.; Radder, J.A.; Angelos, J.G.; Inhaber, H.

    1993-02-01

    The Office of Nuclear Energy of the Department of Energy chartered the Plutonium Disposition Task Force in August 1992. The Task Force was created to assess the range of practicable means of disposition of excess weapons-grade plutonium. Within the Task Force, working groups were formed to consider: (1) storage, (2) disposal,and(3) fission options for this disposition,and a separate group to evaluate nonproliferation concerns of each of the alternatives. As a member of the Fission Working Group, the Savannah River Technology Center acted as a sponsor for light water reactor (LWR) technology. The information contained in this report details the submittal that was made to the Fission Working Group of the technical assessment of LWR technology for plutonium disposition. The following aspects were considered: (1) proliferation issues, (2) technical feasibility, (3) technical availability, (4) economics, (5) regulatory issues, and (6) political acceptance.

  18. Interaction of divalent plutonium and curium

    SciTech Connect

    Mikheev, N.B.; Kazakevich, M.Z.; Rumer, I.A.

    1988-11-01

    It has been established that at plutonium concentrations ranging from 10/sup -5/ to 10/sup -4/ mole % the oxidation potentials of the Pu/sup 3 +//Pu/sup 2 +/ and Cm/sup 3 +//Cm/sup 2 +/ pairs increased by 0.15-0.2 V due to the dimerization of Pu/sup 2 +/ and the formation of mixed dimers of plutonium and curium. Promethium(2+) does not have a similar ability to form mixed dimers owing to the fact that Pm/sup 2 +/ does not have a free d electron. The oxidation potential of the Pm/sup 3 +//Pm/sup 2 +/ pair does not vary in the presence of massive quantities of plutonium

  19. Disposition of plutonium in deep boreholes

    SciTech Connect

    Halsey, W.G.; Jardine, L.J.; Walter, C.E.

    1995-05-01

    Substantial inventories of excess plutonium are expected to result from dismantlement of U.S. and Russian nuclear weapons. Disposition of this material should be a high priority in both countries. A variety of disposition options are under consideration. One option is to place the plutonium either directly or in an immobilized form at the bottom of a deep borehole that is then sealed. Deep-borehole disposition involves placing plutonium several kilometers deep into old, stable, rock formations that have negligible free water present. Containment assurance is based on the presence of ancient groundwater indicating lack of migration and communication with the biosphere. Recovery would be extremely difficult (costly) and impossible to accomplish clandestinely.

  20. PLUTONIUM ALLOYS CONTAINING CONTROLLED AMOUNTS OF PLUTONIUM ALLOTROPES OBTAINED BY APPLICATION OF HIGH PRESSURES

    DOEpatents

    Elliott, R.O.; Gschneidner, K.A. Jr.

    1962-07-10

    A method of making stabilized plutonium alloys which are free of voids and cracks and have a controlled amount of plutonium allotropes is described. The steps include adding at least 4.5 at.% of hafnium, indium, or erbium to the melted plutonium metal, homogenizing the resulting alloy at a temperature of 450 deg C, cooling to room temperature, and subjecting the alloy to a pressure which produces a rapid increase in density with a negligible increase in pressure. The pressure required to cause this rapid change in density or transformation ranges from about 800 to 2400 atmospheres, and is dependent on the alloying element. (AEC)

  1. Study of the formation, prevention, and recovery of plutonium from plutonium esters in the Purex process

    SciTech Connect

    Gray, L. W.; Burney, G. A.

    1981-01-01

    The Savannah River Plant uses the basic Purex process to separate /sup 239/Pu from /sup 238/U and fission products. Dark-brown, dense solids containing up to 30% Pu have previously occurred in rotameters in the plutonium finishing operations. The kinetics of formation of this mixture of DBP- and MBP-Pu esters suggest two methods to prevent the formation of the solids. A selective dissolution method using NaOH metathesis has been developed to separate the phosphate ester from the plutonium before dissolution of the residual plutonium hydroxide in a HNO/sub 3/-HF medium.

  2. Alternating layers of plutonium and lead or indium as surrogate for plutonium

    SciTech Connect

    Rudin, Sven Peter

    2009-01-01

    Elemental plutonium (Pu) assumes more crystal structures than other elements, plausibly due to bonding f electrons becoming non-bonding. Complex geometries hamper understanding of the transition in Pu, but calculations predict this transition in a system with simpler geometry: alternating layers either of plutonium and lead or of plutonium and indium. Here the transition occurs via a pairing-up of atoms within Pu layers. Calculations stepping through this pairing-up reveal valuable details of the transition, for example that the transition from bonding to non-bonding proceeds smoothly.

  3. Determination of plutonium metal origins

    SciTech Connect

    Moody, K.J.

    1995-02-01

    Forensic signatures are present in any Pu sample that can determine the sample`s origin: isotopic ratio of Pu, progeny species that grow into the sample, and contaminant species left over from incomplete purification of the Pu in fuel reprocessing. In the context of intelligence information, this can result in attribution of responsibility for the product of clandestine proliferant operations or material smuggled from existing stockpiles. A list of signature elements and what can be determined from them have been developed. Work needs to be done in converting concentrations of signature species into a quantitative forensic analysis, particularly in regard to reactor performance, but this should require only a small effort. A radiochemical analysis scheme has been developed for measuring these nuclides; more work is needed, particularly for determining fission product concentrations. A sample of Pu metal has been analyzed and several parameters determined that are strong indicators of its point of origin.

  4. CHARACTERIZATION OF SURPLUS PLUTONIUM FOR DISPOSITION OPTIONS

    SciTech Connect

    Allender, J; Edwin Moore, E; Scott Davies, S

    2008-07-15

    The United States (U.S.) has identified 61.5 metric tons (MT) of plutonium that is permanently excess to use in nuclear weapons programs, including 47.2 MT of weapons-grade plutonium. Except for materials that remain in use for programs outside of national defense, including programs for nuclear-energy development, the surplus inventories will be stored safely by the Department of Energy (DOE) and then transferred to facilities that will prepare the plutonium for permanent disposition. Some items will be disposed as transuranic waste, low-level waste, or spent fuel. The remaining surplus plutonium will be managed through: (1) the Mixed Oxide (MOX) Fuel Fabrication Facility (FFF), to be constructed at the Savannah River Site (SRS), where the plutonium will be converted to fuel that will be irradiated in civilian power reactors and later disposed to a high-level waste (HLW) repository as spent fuel; (2) the SRS H-Area facilities, by dissolving and transfer to HLW systems, also for disposal to the repository; or (3) alternative immobilization techniques that would provide durable and secure disposal. From the beginning of the U.S. program for surplus plutonium disposition, DOE has sponsored research to characterize the surplus materials and to judge their suitability for planned disposition options. Because many of the items are stored without extensive analyses of their current chemical content, the characterization involves three interacting components: laboratory sample analysis, if available; non-destructive assay data; and rigorous evaluation of records for the processing history for items and inventory groups. This information is collected from subject-matter experts at inventory sites and from materials stabilization and surveillance programs, in cooperation with the design agencies for the disposition facilities. This report describes the operation and status of the characterization program.

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

    SciTech Connect

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

    2004-07-02

    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).

  6. REVIEW OF PLUTONIUM OXIDATION LITERATURE

    SciTech Connect

    Korinko, P.

    2009-11-12

    A brief review of plutonium oxidation literature was conducted. The purpose of the review was to ascertain the effect of oxidation conditions on oxide morphology to support the design and operation of the PDCF direct metal oxidation (DMO) furnace. The interest in the review was due to a new furnace design that resulted in oxide characteristics that are different than those of the original furnace. Very little of the published literature is directly relevant to the DMO furnace operation, which makes assimilation of the literature data with operating conditions and data a convoluted task. The oxidation behavior can be distilled into three regimes, a low temperature regime (RT to 350 C) with a relatively slow oxidation rate that is influenced by moisture, a moderate temperature regime (350-450 C) that is temperature dependent and relies on more or less conventional oxidation growth of a partially protective oxide scale, and high temperature oxidation (> 500 C) where the metal autocatalytically combusts and oxidizes. The particle sizes obtained from these three regimes vary with the finest being from the lowest temperature. It is surmised that the slow growth rate permits significant stress levels to be achieved that help break up the oxides. The intermediate temperatures result in a fairly compact scale that is partially protective and that grows to critical thickness prior to fracturing. The growth rate in this regime may be parabolic or paralinear, depending on the oxidation time and consequently the oxide thickness. The high temperature oxidation is invariant in quiescent or nearly quiescent conditions due to gas blanketing while it accelerates with temperature under flowing conditions. The oxide morphology will generally consist of fine particles (<15 {micro}m), moderately sized particles (15 < x < 250 {micro}m) and large particles (> 250 {micro}m). The particle size ratio is expected to be < 5%, 25%, and 70% for fine, medium and large particles, respectively, for

  7. Dehydration of plutonium or neptunium trichloride hydrate

    DOEpatents

    Foropoulos, J. Jr.; Avens, L.R.; Trujillo, E.A.

    1992-03-24

    A process is described for preparing anhydrous actinide metal trichlorides of plutonium or neptunium by reacting an aqueous solution of an actinide metal trichloride selected from the group consisting of plutonium trichloride or neptunium trichloride with a reducing agent capable of converting the actinide metal from an oxidation state of +4 to +3 in a resultant solution, evaporating essentially all the solvent from the resultant solution to yield an actinide trichloride hydrate material, dehydrating the actinide trichloride hydrate material by heating the material in admixture with excess thionyl chloride, and recovering anhydrous actinide trichloride.

  8. Dehydration of plutonium or neptunium trichloride hydrate

    DOEpatents

    Foropoulos, Jr., Jerry; Avens, Larry R.; Trujillo, Eddie A.

    1992-01-01

    A process of preparing anhydrous actinide metal trichlorides of plutonium or neptunium by reacting an aqueous solution of an actinide metal trichloride selected from the group consisting of plutonium trichloride or neptunium trichloride with a reducing agent capable of converting the actinide metal from an oxidation state of +4 to +3 in a resultant solution, evaporating essentially all the solvent from the resultant solution to yield an actinide trichloride hydrate material, dehydrating the actinide trichloride hydrate material by heating the material in admixture with excess thionyl chloride, and recovering anhydrous actinide trichloride is provided.

  9. Waste measurements at a plutonium facility

    SciTech Connect

    Wachter, J.R.

    1992-01-01

    Solid plutonium contaminated wastes are often highly heterogeneous, span a wide range of chemical compositions and matrix types, and are packaged in a variety of container sizes. NDA analysis of this waste depends on operator knowledge of these parameters so that proper segregation, instrument selection, quality assurance, and uncertainty estimation can take place. This report describes current waste measurement practices and uncertainty estimates at a US plutonium scrap recovery facility and presents a program for determining reproducibility and bias in NDA measurements. Following this, an operator's perspective on desirable NDA upgrades is offered.

  10. Measurement of Plutonium Isotopic Composition - MGA

    SciTech Connect

    Vo, Duc Ta

    2015-08-21

    In this module, we will use the Canberra InSpector-2000 Multichannel Analyzer with a high-purity germanium detector (HPGe) and the MGA isotopic anlysis software to assay a variety of plutonium samples. The module provides an understanding of the MGA method, its attributes and limitations. You will assess the system performance by measuring a range of materials similar to those you may assay in your work. During the final verification exercise, the results from MGA will be combined with the 240Pueff results from neutron coincidence or multiplicity counters so that measurements of the plutonium mass can be compared with the operator-declared (certified) values.

  11. Sequential Determination of Free Acidity and Plutonium Concentration in the Dissolver Solution of Fast-Breeder Reactor Spent Fuels in a Single Aliquot.

    PubMed

    Dhamodharan, K; Pius, Anitha

    2016-01-01

    A simple potentiometric method for determining the free acidity without complexation in the presence of hydrolysable metal ions and sequentially determining the plutonium concentration by a direct spectrophotometric method using a single aliquot was developed. Interference from the major fission products, which are susceptible to hydrolysis at lower acidities, had been investigated in the free acidity measurement. This method is applicable for determining the free acidity over a wide range of nitric acid concentrations as well as the plutonium concentration in the irradiated fuel solution prior to solvent extraction. Since no complexing agent is introduced during the measurement of the free acidity, the purification step is eliminated during the plutonium estimation, and the resultant analytical waste is free from corrosive chemicals and any complexing agent. Hence, uranium and plutonium can be easily recovered from analytical waste by the conventional solvent extraction method. The error involved in determining the free acidity and plutonium is within ±1% and thus this method is superior to the complexation method for routine analysis of plant samples and is also amenable for remote analysis.

  12. Multi-isotopic determination of plutonium (239Pu, 240Pu, 241Pu and 242Pu) in marine sediments using sector-field inductively coupled plasma mass spectrometry.

    PubMed

    Donard, O F X; Bruneau, F; Moldovan, M; Garraud, H; Epov, V N; Boust, D

    2007-03-28

    Among the transuranic elements present in the environment, plutonium isotopes are mainly attached to particles, and therefore they present a great interest for the study and modelling of particle transport in the marine environment. Except in the close vicinity of industrial sources, plutonium concentration in marine sediments is very low (from 10(-4) ng kg(-1) for (241)Pu to 10 ng kg(-1) for (239)Pu), and therefore the measurement of (238)Pu, (239)Pu, (240)Pu, (241)Pu and (242)Pu in sediments at such concentration level requires the use of very sensitive techniques. Moreover, sediment matrix contains huge amounts of mineral species, uranium and organic substances that must be removed before the determination of plutonium isotopes. Hence, an efficient sample preparation step is necessary prior to analysis. Within this work, a chemical procedure for the extraction, purification and pre-concentration of plutonium from marine sediments prior to sector-field inductively coupled plasma mass spectrometry (SF-ICP-MS) analysis has been optimized. The analytical method developed yields a pre-concentrated solution of plutonium from which (238)U and (241)Am have been removed, and which is suitable for the direct and simultaneous measurement of (239)Pu, (240)Pu, (241)Pu and (242)Pu by SF-ICP-MS.

  13. A review of research programs related to the behavior of plutonium in the environment

    SciTech Connect

    Bartram, Bart W.; Wilkinson, Martha J.

    1983-06-15

    Plutonium-fueled radioisotopic heat sources find application in a spectrum of space, terrestrial, and underseas applications to generate electrical power by thermoelectric or dynamic-cycle conversion. Such systems under postulated accident conditions could release radioactivity into the environment resulting in risks to the general population. The released radioactivity could be dispersed into various environmental media, such as air, soil, and water and interact with people through various exposure pathways leading to inhalation, ingestion, and external radiological doses and associated health effects. The authors developed short-term exposure (RISK II) and long-term exposure (RISK III) models for use in safety risk assessments of space missions utilizing plutonium-fueled electric power systems. To effectively use these models in risk assessments, representative input values must be selected for a spectrum of environmental transfer parameters that characterize the behavior of plutonium in the environment. The selection of appropriate transfer parameters to be used in a given analysis will depend on the accident scenarios to be modeled and the terrestrial and aquatic environments to be encountered. The authors reviewed the availability of plutonium in the environment. This report summarizes the research programs presently being conducted at six Department of Energy Laboratories and makes recommendations on areas where further research is needed to fill gaps in the data necessary for risk assessments

  14. PROGRESS IN REDUCING THE NUCLEAR THREAT: UNITED STATES PLUTONIUM CONSOLIDATION AND DISPOSITION

    SciTech Connect

    Allender, J.; Koenig, R.; Davies, S.

    2009-06-01

    Following the end of the Cold War, the United States identified 61.5 metric tons (MT) of plutonium and larger quantities of enriched uranium that are permanently excess to use in nuclear weapons programs. The Department of Energy (DOE) also began shutting down, stabilizing, and removing inventories from production facilities that were no longer needed to support weapons programs and non-weapons activities. The storage of 'Category I' nuclear materials at Rocky Flats, Sandia National Laboratories, and several smaller sites has been terminated to reduce costs and safeguards risks. De-inventory continues at the Hanford site and the Lawrence Livermore National Laboratory. Consolidation of inventories works in concert with the permanent disposition of excess inventories, including several tonnes of plutonium that have already been disposed to waste repositories and the preparation for transfers to the planned Mixed Oxide (MOX) Fuel Fabrication Facility (for the bulk of the excess plutonium) and alternative disposition methods for material that cannot be used readily in the MOX fuel cycle. This report describes status of plutonium consolidation and disposition activities and their impacts on continuing operations, particularly at the Savannah River Site.

  15. SEPARATION OF PLUTONIUM FROM LANTHANUM BY CHELATION-EXTRACTION

    DOEpatents

    James, R.A.; Thompson, S.G.

    1958-12-01

    Plutonium can be separated from a mixture of plutonlum and lanthanum in which the lanthanum to plutonium molal ratio ls at least five by adding the ammonium salt of N-nitrosoarylhydroxylamine to an aqueous solution having a pH between about 3 and 0.2 and containing the plutonium in a valence state of at least +3, to form a plutonium chelate compound of N-nitrosoarylhydroxylamine. The plutonium chelate compound may be recovered from the solution by extracting with an immiscible organic solvent such as chloroform.

  16. Plutonium dispersal in fires: Summary of what is known

    SciTech Connect

    Condit, R.H.

    1993-07-01

    In view of the great public apprehension about plutonium and nuclear weapons we should explore ways to prevent, limit, or mitigate possible plutonium dispersals. This review is primarily a tutorial on what is known about plutonium dispersal in fires. It concludes that in most types of fires involving plutonium the amount released will not be an immediate danger to life. Indeed, in many cases very few personnel will receive more than the lung burden allowed by current regulations for plutonium workers. However, the dangers may be significant in special situations, unusual terrains, certain meteorological conditions, and very high burn temperatures.

  17. Plutonium Finishing Plant (PFP) Standards/Requirements Identification Document (S/RID)

    SciTech Connect

    Maddox, B.S.

    1996-01-01

    This Standards/Requirements Identification Document (S/RID) sets forth the Environmental Safety and Health (ESH) standards/requirements for the Plutonium Finishing Plant (PFP). This S/RID is applicable to the appropriate life cycle phases of design, construction, operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.

  18. Using magnetization measurements to detect small amounts of plutonium hydride formation in plutonium metal

    SciTech Connect

    Kim, Jae Wook; Mielke, Charles H.; Zapf, Vivien; Baiardo, Joseph P.; Mitchell, Jeremy N.; Richmond, Scott; Schwartz, Daniel S.; Mun, Eun D.; Smith, Alice Iulia

    2014-10-20

    We report the formation of plutonium hydride in 2 at % Ga-stabilized δ-Pu, with 1 atomic % H charging. We show that magnetization measurements are a sensitive, quantitative measure of ferromagnetic plutonium hydride against the nonmagnetic background of plutonium. It was previously shown that at low hydrogen concentrations, hydrogen forms super-abundant vacancy complexes with plutonium, resulting in a bulk lattice contraction. Here we use magnetization, X-ray and neutron diffraction measurements to show that in addition to forming vacancy complexes, at least 30% of the H atoms bond with Pu to precipitate PuHx, largely on the surface of the sample with x ~ 1.9. We observe magnetic hysteresis loops below 40 K with magnetic remanence, consistent with precipitates of ferromagnetic PuH1.9.

  19. Excess Weapons Plutonium Disposition: Plutonium Packaging, Storage and Transportation and Waste Treatment, Storage and Disposal Activities

    SciTech Connect

    Jardine, L J; Borisov, G B

    2004-07-21

    A fifth annual Excess Weapons Plutonium Disposition meeting organized by Lawrence Livermore National Laboratory (LLNL) was held February 16-18, 2004, at the State Education Center (SEC), 4 Aerodromnya Drive, St. Petersburg, Russia. The meeting discussed Excess Weapons Plutonium Disposition topics for which LLNL has the US Technical Lead Organization responsibilities. The technical areas discussed included Radioactive Waste Treatment, Storage, and Disposal, Plutonium Oxide and Plutonium Metal Packaging, Storage and Transportation and Spent Fuel Packaging, Storage and Transportation. The meeting was conducted with a conference format using technical presentations of papers with simultaneous translation into English and Russian. There were 46 Russian attendees from 14 different Russian organizations and six non-Russian attendees, four from the US and two from France. Forty technical presentations were made. The meeting agenda is given in Appendix B and the attendance list is in Appendix C.

  20. Quantitative ion-exchange separation of plutonium from impurities

    SciTech Connect

    Pietri, C.E.; Freeman, B.P.; Weiss, J.R.

    1981-09-01

    The methods used at the New Brunswick Laboratory for the quantitative ion exchange separation of plutonium from impurities prior to plutonium assay are described. Other ion exchange separation procedures for impurity determination and for isotopic abundance measurements are given. The primary technique used consists of sorption of plutonium(IV) in 8N HNO/sub 3/ on Dowex-1 anion exchange resin and elution of the purified plutonium with 0.3N HCl-0.01N HF. Other methods consist of the anion exchange separation of plutonium(IV) in 12N HCl and the cation exchange separation of plutonium(III) in 0.2 N HNO/sub 3/. The application of these procedures to the subsequent assay of plutonium, isotopic analysis, and impurity determination is described.

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

    SciTech Connect

    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.

  2. Adsorption of plutonium oxide nanoparticles.

    PubMed

    Schmidt, Moritz; Wilson, Richard E; Lee, Sang Soo; Soderholm, L; Fenter, P

    2012-02-07

    Adsorption of monodisperse cubic plutonium oxide nanoparticles ("Pu-NP", [Pu(38)O(56)Cl(x)(H(2)O)(y)]((40-x)+), with a fluorite-related lattice, approximately 1 nm in edge size) to the muscovite (001) basal plane from aqueous solutions was observed in situ (in 100 mM NaCl background electrolyte at pH 2.6). Uptake capacity of the surface quantified by α-spectrometry was 0.92 μg Pu/cm(2), corresponding to 10.8 Pu per unit cell area (A(UC)). This amount is significantly larger than that of Pu(4+) needed for satisfying the negative surface charge (0.25 Pu(4+) for 1 e(-)/A(UC)). The adsorbed Pu-NPs cover 17% of the surface area, determined by X-ray reflectivity (XR). This correlates to one Pu-NP for every 14 unit cells of muscovite, suggesting that each particle compensates the charge of the unit cells onto which it adsorbs as well as those in its direct proximity. Structural investigation by resonant anomalous X-ray reflectivity distinguished two different sorption states of Pu-NPs on the surface at two different regimes of distance from the surface. A fraction of Pu is distributed within 11 Å from the surface. The distribution width matches the Pu-NP size, indicating that this species represents Pu-NPs adsorbed directly on the surface. Beyond the first layer, an additional fraction of sorbed Pu was observed to extend more broadly up to more than 100 Å from the surface. This distribution is interpreted as resulting from "stacking" or aggregation of the nanoparticles driven by sorption and accumulation of Pu-NPs at the interface although these Pu-NPs do not aggregate in the solution. These results are the first in situ observation of the interaction of nanoparticles with a charged mineral-water interface yielding information important to understanding the environmental transport of Pu and other nanophase inorganic species.

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

    SciTech Connect

    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

  4. Plutonium Immobilization Can Loading Equipment Review

    SciTech Connect

    Kriikku, E.; Ward, C.; Stokes, M.; Randall, B.; Steed, J.; Jones, R.; Hamilton, L.

    1998-05-01

    This report lists the operations required to complete the Can Loading steps on the Pu Immobilization Plant Flow Sheets and evaluates the equipment options to complete each operation. This report recommends the most appropriate equipment to support Plutonium Immobilization Can Loading operations.

  5. Overview of surplus weapons plutonium disposition

    SciTech Connect

    Rudy, G.

    1996-05-01

    The safe disposition of surplus weapons useable plutonium is a very important and urgent task. While the functions of long term storage and disposition directly relate to the Department`s weapons program and the environmental management program, the focus of this effort is particularly national security and nonproliferation.

  6. Recovery of Plutonium by Carrier Precipitation

    DOEpatents

    Goeckermann, R. H.

    1961-04-01

    The recovery of plutonium from an aqueous nitric acid Zr-containing solution of 0.2 to 1N acidity is accomplished by adding fluoride anions (1.5 to 5 mg/l), and precipitating the Pu with an excess of H/sub 2/0/sub 2/ at 53 to 65 deg C. (AEC)

  7. Plutonium isotope ratio variations in North America

    SciTech Connect

    Steiner, Robert E; La Mont, Stephen P; Eisele, William F; Fresquez, Philip R; Mc Naughton, Michael; Whicker, Jeffrey J

    2010-12-14

    Historically, approximately 12,000 TBq of plutonium was distributed throughout the global biosphere by thermo nuclear weapons testing. The resultant global plutonium fallout is a complex mixture whose {sup 240}Pu/{sup 239}Pu atom ratio is a function of the design and yield of the devices tested. The average {sup 240}Pu/{sup 239}Pu atom ratio in global fallout is 0.176 + 014. However, the {sup 240}Pu/{sup 239}Pu atom ratio at any location may differ significantly from 0.176. Plutonium has also been released by discharges and accidents associated with the commercial and weapons related nuclear industries. At many locations contributions from this plutonium significantly alters the {sup 240}Pu/{sup 239}Pu atom ratios from those observed in global fallout. We have measured the {sup 240}Pu/{sup 239}Pu atom ratios in environmental samples collected from many locations in North America. This presentation will summarize the analytical results from these measurements. Special emphasis will be placed on interpretation of the significance of the {sup 240}Pu/{sup 239}Pu atom ratios measured in environmental samples collected in the Arctic and in the western portions of the United States.

  8. NNSS Soils Monitoring: Plutonium Valley (CAU366)

    SciTech Connect

    Miller, Julianne J.; Mizell, Steve A.; Nikolich, George; Campbell, Scott

    2012-02-01

    The U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), Nevada Site Office (NSO), Environmental Restoration Soils Activity has authorized the Desert Research Institute (DRI) to conduct field assessments of potential sediment transport of contaminated soil from Corrective Action Unit (CAU) 366, Area 11 Plutonium Valley Dispersion Sites Contamination Area (CA) during precipitation runoff events.

  9. 233-S plutonium concentration facility hazards assessment

    SciTech Connect

    Broz, R.E.

    1994-12-19

    This document establishes the technical basis in support of Emergency Planning activities for the 233-S Plutonium Concentration Facility on the Hanford Site. The document represents an acceptable interpretation of the implementing guidance document for DOE ORDER 5500.3A. Through this document, the technical basis for the development of facility specific Emergency Action Levels and the Emergency Planning Zone is demonstrated.

  10. Method for calibration of plutonium NDA

    SciTech Connect

    Lemming, J.F.; Campbell, A.R.; Rodenburg, W.W.

    1980-01-01

    Calibration materials characterized by calorimetric assay can be a practical alternative to synthetic standards for the calibration of plutonium nondestructive assay. Calorimetric assay is an effective measurement system for the characterization because: it can give an absolute assay from first principles when the isotopic composition is known, it is insensitive to most matrix effects, and its traceability to international measurement systems has been demonstrated.

  11. Heterogeneity Effects in Plutonium Contaminated Soil

    DTIC Science & Technology

    2009-03-01

    masses up to one kilogram once the ratio of Americium -241 (Am-241) and plutonium concentrations was established (Rademacher, 2001). Alpha...with a sample number and tared weight with a non-smearing marker. A standard control was then set using a point source of Americium -241 on an aluminum

  12. Regulatory issues for deep borehole plutonium disposition

    SciTech Connect

    Halsey, W.G.

    1995-03-01

    As a result of recent changes throughout the world, a substantial inventory of excess separated plutonium is expected to result from dismantlement of US nuclear weapons. The safe and secure management and eventual disposition of this plutonium, and of a similar inventory in Russia, is a high priority. A variety of options (both interim and permanent) are under consideration to manage this material. The permanent solutions can be categorized into two broad groups: direct disposal and utilization. The deep borehole disposition concept involves placing excess plutonium deep into old stable rock formations with little free water present. Issues of concern include the regulatory, statutory and policy status of such a facility, the availability of sites with desirable characteristics and the technologies required for drilling deep holes, characterizing them, emplacing excess plutonium and sealing the holes. This white paper discusses the regulatory issues. Regulatory issues concerning construction, operation and decommissioning of the surface facility do not appear to be controversial, with existing regulations providing adequate coverage. It is in the areas of siting, licensing and long term environmental protection that current regulations may be inappropriate. This is because many current regulations are by intent or by default specific to waste forms, facilities or missions significantly different from deep borehole disposition of excess weapons usable fissile material. It is expected that custom regulations can be evolved in the context of this mission.

  13. Electrochemically Modulated Separation for Plutonium Safeguards

    SciTech Connect

    Pratt, Sandra H.; Breshears, Andrew T.; Arrigo, Leah M.; Schwantes, Jon M.; Duckworth, Douglas C.

    2013-12-31

    Accurate and timely analysis of plutonium in spent nuclear fuel is critical in nuclear safeguards for detection of both protracted and rapid plutonium diversions. Gamma spectroscopy is a viable method for accurate and timely measurements of plutonium provided that the plutonium is well separated from the interfering fission and activation products present in spent nuclear fuel. Electrochemically modulated separation (EMS) is a method that has been used successfully to isolate picogram amounts of Pu from nitric acid matrices. With EMS, Pu adsorption may be turned "on" and "off" depending on the applied voltage, allowing for collection and stripping of Pu without the addition of chemical reagents. In this work, we have scaled up the EMS process to isolate microgram quantities of Pu from matrices encountered in spent nuclear fuel during reprocessing. Several challenges have been addressed including surface area limitations, radiolysis effects, electrochemical cell performance stability, and chemical interferences. After these challenges were resolved, 6 µg Pu was deposited in the electrochemical cell with approximately an 800-fold reduction of fission and activation product levels from a spent nuclear fuel sample. Modeling showed that these levels of Pu collection and interference reduction may not be sufficient for Pu detection by gamma spectroscopy. The main remaining challenges are to achieve a more complete Pu isolation and to deposit larger quantities of Pu for successful gamma analysis of Pu. If gamma analyses of Pu are successful, EMS will allow for accurate and timely on-site analysis for enhanced Pu safeguards.

  14. Plutonium Immobilization Can Loading Preliminary Specifications

    SciTech Connect

    Kriikku, E.

    1998-11-25

    This report discusses the Plutonium Immobilization can loading preliminary equipment specifications and includes a process block diagram, process description, equipment list, preliminary equipment specifications, plan and elevation sketches, and some commercial catalogs. This report identifies loading pucks into cans and backfilling cans with helium as the top priority can loading development areas.

  15. In search of plutonium: A nonproliferation journey

    NASA Astrophysics Data System (ADS)

    Hecker, Siegfried

    2010-02-01

    In February 1992, I landed in the formerly secret city of Sarov, the Russian Los Alamos, followed a few days later by a visit to Snezhinsk, their Livermore. The briefings we received of the Russian nuclear weapons program and tours of their plutonium, reactor, explosives, and laser facilities were mind boggling considering the Soviet Union was dissolved only two months earlier. This visit began a 17-year, 41 journey relationship with the Russian nuclear complex dedicated to working with them in partnership to protect and safeguard their weapons and fissile materials, while addressing the plight of their scientists and engineers. In the process, we solved a forty-year disagreement about the plutonium-gallium phase diagram and began a series of fundamental plutonium science workshops that are now in their tenth year. At the Yonbyon reprocessing facility in January 2004, my North Korean hosts had hoped to convince me that they have a nuclear deterrent. When I expressed skepticism, they asked if I wanted to see their ``product.'' I asked if they meant the plutonium; they replied, ``Well, yes.'' Thus, I wound up holding 200 grams of North Korean plutonium (in a sealed glass jar) to make sure it was heavy and warm. So began the first of my six journeys to North Korea to provide technical input to the continuing North Korean nuclear puzzle. In Trombay and Kalpakkam a few years later I visited the Indian nuclear research centers to try to understand how India's ambitious plans for nuclear power expansion can be accomplished safely and securely. I will describe these and other attempts to deal with the nonproliferation legacy of the cold war and the new challenges ahead. )

  16. PLUTONIUM CONTAMINATION VALENCE STATE DETERMINATION USING X-RAY ABSORPTION FINE STRUCTURE PERMITS CONCRETE RECYCLE

    SciTech Connect

    Ervin, P. F.; Conradson, S. D.

    2002-02-25

    This paper describes the determination of the speciation of plutonium contamination present on concrete surfaces at the Rocky Flats Environmental Technology Site (RFETS). At RFETS, the plutonium processing facilities have been contaminated during multiple events over their 50 year operating history. Contamination has resulted from plutonium fire smoke, plutonium fire fighting water, milling and lathe operation aerosols, furnace operations vapors and plutonium ''dust'' diffusion.

  17. A Note on the Reaction of Hydrogen and Plutonium

    SciTech Connect

    Noone, Bailey C

    2012-08-15

    Plutonium hydride has many practical and experimental purposes. The reaction of plutonium and hydrogen has interesting characteristics, which will be explored in the following analysis. Plutonium is a radioactive actinide metal that emits alpha particles. When plutonium metal is exposed to air, the plutonium oxides and hydrides, and the volume increases. PuH{sub 2} and Pu{sub 2}O{sub 3} are the products. Hydrogen is a catalyst for plutonium's corrosion in air. The reaction can take place at room temperature because it is fairly insensitive to temperature. Plutonium hydride, or PuH{sub 2}, is black and metallic. After PuH{sub 2} is formed, it quickly flakes off and burns. The reaction of hydrogen and plutonium is described as pyrophoric because the product will spontaneously ignite when oxygen is present. This tendency must be considered in the storage of metal plutonium. The reaction is characterized as reversible and nonstoichiometric. The reaction goes as such: Pu + H{sub 2} {yields} PuH{sub 2}. When PuH{sub 2} is formed, the hydrogen/plutonium ratio is between 2 and 2.75 (approximately). As more hydrogen is added to the system, the ratio increases. When the ratio exceeds 2.75, PuH{sub 3} begins to form along with PuH{sub 2}. Once the ratio surpasses 2.9, only PuH{sub 3} remains. The volume of the plutonium sample increases because of the added hydrogen and the change in crystal structure which the sample undergoes. As more hydrogen is added to a system of metal plutonium, the crystal structure evolves. Plutonium has a crystal structure classified as monoclinic. A monoclinic crystal structure appears to be a rectangular prism. When plutonium reacts with hydrogen, the product PuH{sub 2}, becomes a fluorite structure. It can also be described as a face centered cubic structure. PuH{sub 3} forms a hexagonal crystal structure. As plutonium evolves from metal plutonium to plutonium hydride to plutonium trihydride, the crystal structure evolves from monoclinic to

  18. Kilogram-scale purification of americium by ion exchange

    SciTech Connect

    Wheelwright, E.J.

    1980-05-01

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

  19. Plutonium speciation, solubilization, and migration in soils. 1998 annual progress report

    SciTech Connect

    Neu, M.; Runde, W.; Haire, R.G.

    1998-06-01

    'The DOE is currently conducting cleanup activities at its nuclear weapons development sites, many of which have accumulated plutonium in soils for 50 years. To properly control Pu migration in soils within Federal sites and onto public lands, better evaluate the public risk, and design effective remediation strategies, a fundamental understanding of Pu speciation, transport, and release mechanisms is needed. Key scientific goals include: determine Pu concentrations and speciation at a contaminated DOE site; study the formation, stability, and structural and spectroscopic features of environmentally relevant Pu (III, IV, and V) species; determine the mechanism of interaction between Pu and Mn/Fe minerals and the potential release of Pu via redox cycling; and model the environmental behavior of plutonium.'

  20. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    SciTech Connect

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

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

    SciTech Connect

    McKibben, J.M.; Chostner, D.F.; Orebaugh, E.G.

    1983-11-01

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

  2. METHOD FOR RECOVERING PLUTONIUM VALUES FROM SOLUTION USING A BISMUTH HYDROXIDE CARRIER PRECIPITATE

    DOEpatents

    Faris, B.F.

    1961-04-25

    Carrier precipitation processes for separating plutonium values from aqueous solutions are described. In accordance with the invention a bismuth hydroxide precipitate is formed in the plutonium-containing solution, thereby carrying plutonium values from the solution.

  3. Process modeling of plutonium conversion and MOX fabrication for plutonium disposition

    SciTech Connect

    Schwartz, K. L.

    1998-10-01

    Two processes are currently under consideration for the disposition of 35 MT of surplus plutonium through its conversion into fuel for power production. These processes are the ARIES process, by which plutonium metal is converted into a powdered oxide form, and MOX fuel fabrication, where the oxide powder is combined with uranium oxide powder to form ceramic fuel. This study was undertaken to determine the optimal size for both facilities, whereby the 35 MT of plutonium metal will be converted into fuel and burned for power. The bounding conditions used were a plutonium concentration of 3-7%, a burnup of 20,000-40,000 MWd/MTHM, a core fraction of 0.1 to 0.4, and the number of reactors ranging from 2-6. Using these boundary conditions, the optimal cost was found with a plutonium concentration of 7%. This resulted in an optimal throughput ranging from 2,000 to 5,000 kg Pu/year. The data showed minimal costs, resulting from throughputs in this range, at 3,840, 2,779, and 3,497 kg Pu/year, which results in a facility lifetime of 9.1, 12.6, and 10.0 years, respectively.

  4. Plutonium disposition via immobilization in ceramic or glass

    SciTech Connect

    Gray, L.W.; Kan, T.; Shaw, H.F.; Armantrout, A.

    1997-03-05

    The management of surplus weapons plutonium is an important and urgent task with profound environmental, national, and international security implications. In the aftermath of the Cold War, Presidential Policy Directive 13, and various analyses by renown scientific, technical, and international policy organizations have brought about a focused effort within the Department of Energy to identify and implement paths for the long term disposition of surplus weapons- usable plutonium. The central goal of this effort is to render surplus weapons plutonium as inaccessible and unattractive for reuse in nuclear weapons as the much larger and growing stock of plutonium contained in spent fuel from civilian reactors. One disposition option being considered for surplus plutonium is immobilization, in which the plutonium would be incorporated into a glass or ceramic material that would ultimately be entombed permanently in a geologic repository for high-level waste.

  5. Plutonium speciation in water from Mono Lake, California

    USGS Publications Warehouse

    Cleveland, J.M.; Rees, T.F.; Nash, K.L.

    1983-01-01

    The solubility of plutonium in Mono Lake water is enhanced by the presence of large concentrations of indigenous carbonate ions and moderate concentrations of fluoride ions. In spite of the complex chemical composition of this water, only a few ions govern the behavior of plutonium, as demonstrated by the fact that it was possible to duplicate plutonium speciation in a synthetic water containing only the principal components of Mono Lake water.

  6. 14. END VIEW OF THE PLUTONIUM STORAGE VAULT FROM THE ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    14. END VIEW OF THE PLUTONIUM STORAGE VAULT FROM THE REMOTE CONTROL STATION. THE STACKER-RETRIEVER, A REMOTELY-OPERATED, MECHANIZED TRANSPORT SYSTEM, RETRIEVES CONTAINERS OF PLUTONIUM FROM SAFE GEOMETRY PALLETS STORED ALONG THE LENGTH OF THE VAULT. THE STACKER-RETRIEVER RUNS ALONG THE AISLE BETWEEN THE PALLETS OF THE STORAGE CHAMBER. (3/2/86) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

  7. SEPARATION OF PLUTONIUM FROM AQUEOUS SOLUTIONS BY ION-EXCHANGE

    DOEpatents

    Schubert, J.

    1958-06-01

    A process is described for the separation of plutonium from an aqueous solution of a plutonium salt, which comprises adding to the solution an acid of the group consisting of sulfuric acid, phosphoric acid, and oxalic acid, and mixtures thereof to provide an acid concentration between 0.0001 and 1 M, contacting the resultant solution with a synthetic organic anion exchange resin, and separating the aqueous phase and the resin which contains the plutonium.

  8. Spectrophotometers for plutonium monitoring in HB-line

    SciTech Connect

    Lascola, R. J.; O'Rourke, P. E.; Kyser, E. A.; Immel, D. M.; Plummer, J. R.; Evans, E. V.

    2016-02-12

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

  9. Plutonium speciation in water from Mono Lake, California

    SciTech Connect

    Cleveland, J.M.; Rees, T.F.; Nash, K.L.

    1983-12-23

    The solubility of plutonium in Mono Lake water is enhanced by the presence of large concentrations of indigenous carbonate ions and moderate concentrations of fluoride ions. In spite of the complex chemical composition of this water, only a few ions govern the behavior of plutonium, as demonstrated by the fact that it was possible to duplicate plutonium speciation in a synthetic water containing only the principal components of Mono Lake water.

  10. Investigations of plutonium immobilization into the vitreous compositions

    SciTech Connect

    Matyunin, Y.I.,

    1998-04-15

    Development and characterizations of phosphate and borosilicate glasses for vitrifying high level waste (HLW) solutions in Russia has been extensive. The technical data generated were for low concentrations (less than 0.05% Pu) of plutonium. Limited studies have been performed with plutonium concentrations one to two orders of magnitude larger. The results of these studies are being used to plan and implement an expanded experimental program to establish the limitations and characteristics of plutonium in similar glass compositions.

  11. Plutonium stabilization and handling (PuSH)

    SciTech Connect

    Weiss, E.V.

    1997-01-23

    This Functional Design Criteria (FDC) addresses construction of a Stabilization and Packaging System (SPS) to oxidize and package for long term storage remaining plutonium-bearing special nuclear materials currently in inventory at the Plutonium Finishing Plant (PFP), and modification of vault equipment to allow storage of resulting packages of stabilized SNM for up to fifty years. The major sections of the project are: site preparation; SPS Procurement, Installation, and Testing; storage vault modification; and characterization equipment additions. The SPS will be procured as part of a Department of Energy nationwide common procurement. Specific design crit1460eria for the SPS have been extracted from that contract and are contained in an appendix to this document.

  12. Recent plutonium metal production experience at Hanford

    SciTech Connect

    Gibson, M.W.; Nyman, D.H.

    1989-10-01

    Plutonium metal is produced at the Hanford Site in the Remote Mechanical C (RMC) line. The line is housed in the Plutonium Finishing Plant (PFP). The RMC line was built in the early 1960s and operated until 1973 when it was shut down. The line was restarted in 1985 and has operated on a campaign basis since that time. The fiscal years (FY) 1988/89 RMC line campaigns have shown improved yields and plant safety performance when compared to previous years. This is attributed to numerous process improvements that have been made in the line and to an enhanced standard of disciplined operations. This report discusses the improvements made to the RMC line.

  13. Plutonium fractionation in southern Baltic Sea sediments.

    PubMed

    Strumińska-Parulska, Dagmara I; Skwarzec, Bogdan; Pawlukowska, Magdalena

    2012-01-01

    In this study, different chemical plutonium fractions (dissolved in water, connected to carbonates, connected to oxides, complexed with organic matter, mineral acids soluble and the rest) in sediments from the Vistula River estuary, the Gdańsk Basin and the Bornholm Deep were determined. The distribution of (239+240)Pu in analysed sediments samples was not uniform but dependent on its chemical form, depth and the sediment geomorphology. The highest amount of plutonium exists in middle parts of sediments and comes from the global atmospheric fallout from nuclear tests in 1958-1961. According to all analysed fractions, the biggest amount of (239+240)Pu was in the mobile form, connected to carbonate fractions from the Vistula River estuary, the Gulf of Gdańsk and the Bornholm Deep sediments.

  14. CRITICALITY CURVES FOR PLUTONIUM HYDRAULIC FLUID MIXTURES

    SciTech Connect

    WITTEKIND WD

    2007-10-03

    This Calculation Note performs and documents MCNP criticality calculations for plutonium (100% {sup 239}Pu) hydraulic fluid mixtures. Spherical geometry was used for these generalized criticality safety calculations and three geometries of neutron reflection are: {sm_bullet}bare, {sm_bullet}1 inch of hydraulic fluid, or {sm_bullet}12 inches of hydraulic fluid. This document shows the critical volume and critical mass for various concentrations of plutonium in hydraulic fluid. Between 1 and 2 gallons of hydraulic fluid were discovered in the bottom of HA-23S. This HA-23S hydraulic fluid was reported by engineering to be Fyrquel 220. The hydraulic fluid in GLovebox HA-23S is Fyrquel 220 which contains phosphorus. Critical spherical geometry in air is calculated with 0 in., 1 in., or 12 inches hydraulic fluid reflection.

  15. MEANS FOR PRODUCING PLUTONIUM CHAIN REACTIONS

    DOEpatents

    Wigner, E.P.; Weinberg, A.M.

    1961-01-24

    A neutronic reactor is described with an active portion capable of operating at an energy level of 0.5 to 1000 ev comprising discrete bodies of Pu/ sup 239/ disposed in a body of water which contains not more than 5 molecules of water to one atom of plutonium, the total amount of Pu/sup 239/ being sufficient to sustain a chain reaction. (auth)

  16. Surplus Plutonium Disposition (SPD) Environmental Data Summary

    SciTech Connect

    Fledderman, P.D.

    2000-08-24

    This document provides an overview of existing environmental and ecological information at areas identified as potential locations of the Savannah River Site's (SRS) Surplus Plutonium Disposition (SPD) facilities. This information is required to document existing environmental and baseline conditions from which SPD construction and operation impacts can be defined. It will be used in developing the required preoperational monitoring plan to be used at specific SPD facilities construction sites.

  17. Derivation of plutonium-239 materials disposition categories

    SciTech Connect

    Brough, W.G.

    1995-04-27

    At this time, the Office of Fissile Materials Disposition within the DOE, is assessing alternatives for the disposition of excess fissile materials. To facilitate the assessment, the Plutonium-Bearing Materials Feed Report for the DOE Fissile Materials Disposition Program Alternatives report was written. The development of the material categories and the derivation of the inventory quantities associated with those categories is documented in this report.

  18. Neptunium as a Tool for Reducing Proliferation Risks with Plutonium: A Technical Analysis of its Efficiency and its Drawbacks

    SciTech Connect

    Greneche, Dominique; Ng, Selena; Guesdon, Bernard; Vinoche, Richard; Delpech, Marc; Golfier, Herve; Dolci, Florence; Poinot-Salanon, Christine

    2006-07-01

    Introducing neptunium into the nuclear fuel cycle has been proposed in the past as a way to impede the diversion or the direct use of plutonium to fabricate a nuclear explosive device. This paper aims to technically analyze the industrial consequences should this proposal be implemented. Two scenarios are considered: 1) adding neptunium to fresh uranium oxide (UOX) fuel before irradiation in a light water reactor; 2) separating neptunium together with plutonium from used UOX fuel and using this combined oxide to fabricate mixed oxide (MOX) fuel before subsequent irradiation in a light water reactor. In both cases, assembly calculations for a pressurized water reactor using fresh fuel doped with neptunium are presented for a wide range of neptunium proportions. Consequences on core and fuel performances and the fuel cycle are analyzed. These are weighed against the potential proliferation resistance benefits of adding neptunium due to the increased quantity of the plutonium isotope {sup 238}Pu in the discharged fuel, or due to the potentially increased detectability through gamma ray emissions of a plutonium-neptunium oxide mixture. Finally, the proliferation risk presented by neptunium itself is discussed. (authors)

  19. PLUTONIUM RECOVERY FROM NEUTRON-BOMBARDED URANIUM FUEL

    DOEpatents

    Moore, R.H.

    1962-04-10

    A process of recovering plutonium from neutronbombarded uranium fuel by dissolving the fuel in equimolar aluminum chloride-potassium chloride; heating the mass to above 700 deg C for decomposition of plutonium tetrachloride to the trichloride; extracting the plutonium trichloride into a molten salt containing from 40 to 60 mole % of lithium chloride, from 15 to 40 mole % of sodium chloride, and from 0 to 40 mole % of potassium chloride or calcium chloride; and separating the layer of equimolar chlorides containing the uranium from the layer formed of the plutonium-containing salt is described. (AEC)

  20. Magnetic separation as a plutonium residue enrichment process

    SciTech Connect

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

    1989-01-01

    We have subjected several plutonium contaminated residues to Open Gradient Magnetic Separation (OGMS) on an experimental scale. Separation of graphite, bomb reduction sand, and bomb reduction sand, and bomb reduction sand, slag, and crucible, resulted in a plutonium rich fraction and a plutonium lean fraction. 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 direct oxide reduction and electrorefining pyrochemical salts gave less favorable results. While a fraction very rich in plutonium could be obtained, the plutonium content of the lean fraction was to 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. The detailed results of these experiments and the implications for OGMS use in recycle plutonium processing are discussed. 4 refs., 3 figs., 9 tabs.

  1. Technical considerations and policy requirements for plutonium management

    SciTech Connect

    Christensen, D.C.; Dinehart, S.M.; Yarbro, S.L.

    1995-12-31

    The goals for plutonium management have changed dramatically over the past few years. Today, the challenge is focused on isolating plutonium from the environment and preparing it for permanent disposition. In parallel, the requirements for managing plutonium are rapidly changing. For example, there is a significant increase in public awareness on how facilities operate, increased attention to environmental safety and health (ES and H) concerns, greater interest in minimizing waste, more emphasis on protecting material from theft, providing materials for international inspection, and a resurgence of interest in using plutonium as an energy source. Of highest concern, in the immediate future, is protecting plutonium from theft or diversion, while the national policy on disposition is debated. These expanded requirements are causing a broadening of responsibilities within the Department of Energy (DOE) to include at least seven organizations. An unavoidable consequence is the divergence in approach and short-term goals for managing similar materials within each organization. The technology base does exist, properly, safely, and cost effectively to extract plutonium from excess weapons, residues, waste, and contaminated equipment and facilities, and to properly stabilize it. Extracting the plutonium enables it to be easily inventoried, packaged, and managed to minimize the risk of theft and diversion. Discarding excess plutonium does not sufficiently reduce the risk of diversion, and as a result, long-term containment of plutonium from the environment may not be able to be proven to the satisfaction of the public.

  2. Geomorphology of plutonium in the Northern Rio Grande

    SciTech Connect

    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.

  3. 30. VIEW OF A GLOVEBOX LINE USED IN PLUTONIUM OPERATIONS. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    30. VIEW OF A GLOVEBOX LINE USED IN PLUTONIUM OPERATIONS. SAFETY AND HEALTH CONCERNS WERE OF MAJOR IMPORTANCE AT THE PLANT, BECAUSE OF THE RADIOACTIVE NATURE OF THE MATERIALS USED. PLUTONIUM GIVES OFF ALPHA AND BETA PARTICLES, GAMMA PROTONS, NEUTRONS, AND IS ALSO PYROPHORIC. AS A RESULT, PLUTONIUM OPERATIONS ARE PERFORMED UNDER CONTROLLED CONDITIONS THAT INCLUDE CONTAINMENT, FILTERING, SHIELDING, AND CREATING AN INERT ATMOSPHERE. PLUTONIUM WAS HANDLED WITHIN GLOVEBOXES THAT WERE INTERCONNECTED AND RAN SEVERAL HUNDRED FEET IN LENGTH (5/5/70). - Rocky Flats Plant, Bounded by Indiana Street & Routes 93, 128 & 72, Golden, Jefferson County, CO

  4. Design of the improved plutonium canister assay system (IPCAS)

    SciTech Connect

    Abhold, M. E.; Baker, M. C.; Bourret, S. C.; Polk, P. J.; Vo, Duc T.

    2001-01-01

    The improved Plutonium Canister Assay System (iPCAS) is designed to detect gross and partial defects in the declared plutonium content of plutonium and MOX storage canisters during transfer to storage and process areas of the MOX fuel fabrication facility in Kokkasho, Japan. In addition, an associated Gamma Isotopics System (GIS) will be used to confirm facility-declared plutonium isotopics with accuracy sufficient to reduce the amount of destructive isotopic analysis needed. The design of the iPCAS instrument and its associated GIS is described and the expected performance of the instrument is discussed.

  5. A Plutonium-Contaminated Wound, 1985, USA

    SciTech Connect

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

    2012-02-02

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

  6. Plutonium immobilization in glass and ceramics

    SciTech Connect

    Knecht, D.A.; Murphy, W.M.

    1996-05-01

    The Materials Research Society Nineteenth Annual Symposium on the Scientific Basis for Nuclear Waste Management was held in Boston on November 27 to December 1, 1995. Over 150 papers were presented at the Symposium dealing with all aspects of nuclear waste management and disposal. Fourteen oral sessions and on poster session included a Plenary session on surplus plutonium dispositioning and waste forms. The proceedings, to be published in April, 1996, will provide a highly respected, referred compilation of the state of scientific development in the field of nuclear waste management. This paper provides a brief overview of the selected Symposium papers that are applicable to plutonium immobilization and plutonium waste form performance. Waste forms that were described at the Symposium cover most of the candidate Pu immobilization options under consideration, including borosilicate glass with a melting temperature of 1150 {degrees}C, a higher temperature (1450 {degrees}C) lanthanide glass, single phase ceramics, multi-phase ceramics, and multi-phase crystal-glass composites (glass-ceramics or slags). These Symposium papers selected for this overview provide the current status of the technology in these areas and give references to the relevant literature.

  7. Proposed Modification to the Plutonium Systemic Model.

    PubMed

    Konzen, Kevin; Miller, Scott; Brey, Richard

    2015-10-01

    The currently accepted biokinetic model for plutonium distribution within the human body was recommended by the International Commission on Radiological Protection in publication 67. This model was developed from human and animal studies and behavioral knowledge acquired from other known bone-seeking radionuclides. The biokinetic model provides a mathematical means of predicting the distribution, retention, and clearance of plutonium within the human body that may be used in deriving organ, tissue, and whole body dose. This work proposed a modification to the ICRP 67 systemic model for plutonium that incorporated the latest knowledge acquired from recent human injection studies with physiologically based improvements. In summary, the changes included a separation of the liver compartments, removed the intermediate soft tissue-to-bladder pathway, and added pathways from the blood compartment to both the cortical and trabecular bone volumes. The proposed model provided improved predictions for several bioassay indicators compared to the ICRP 67 model while also maintaining its basic structure. Additionally, the proposed model incorporated physiologically based improvements for the liver and skeleton and continued to ensure efficient coupling with intake biokinetic models.

  8. Atomistic modeling of thermodynamic equilibrium of plutonium

    NASA Astrophysics Data System (ADS)

    Lee, Tongsik; Valone, Steve; Baskes, Mike; Chen, Shao-Ping; Lawson, Andrew

    2012-02-01

    Plutonium metal has complex thermodynamic properties. Among its six allotropes at ambient pressure, the fcc delta-phase exhibits a wide range of anomalous behavior: extraordinarily high elastic anisotropy, largest atomic volume despite the close-packed structure, negative thermal expansion, strong elastic softening at elevated temperature, and extreme sensitivity to dilute alloying. An accurate description of these thermodynamic properties goes far beyond the current capability of first-principle calculations. An elaborate modeling strategy at the atomic level is hence an urgent need. We propose a novel atomistic scheme to model elemental plutonium, in particular, to reproduce the anomalous characteristics of the delta-phase. A modified embedded atom method potential is fitted to two energy-volume curves that represent the distinct electronic states of plutonium in order to embody the mechanism of the two-state model of Weiss, in line with the insight originally proposed by Lawson et al. [Philos. Mag. 86, 2713 (2006)]. By the use of various techniques in Monte Carlo simulations, we are able to provide a unified perspective of diverse phenomenological aspects among thermal expansion, elasticity, and phase stability.

  9. Characterizing Surplus US Plutonium for Disposition - 13199

    SciTech Connect

    Allender, Jeffrey S.; Moore, Edwin N.

    2013-07-01

    The United States (US) has identified 61.5 metric tons (MT) of plutonium that is permanently excess to use in nuclear weapons programs, including 47.2 MT of weapons-grade plutonium. Surplus inventories will be stored safely by the Department of Energy (DOE) and then transferred to facilities that will prepare the plutonium for permanent disposition. The Savannah River National Laboratory (SRNL) operates a Feed Characterization program for the Office of Fissile Materials Disposition (OFMD) of the National Nuclear Security Administration (NNSA) and the DOE Office of Environmental Management (DOE-EM). SRNL manages a broad program of item tracking through process history, laboratory analysis, and non-destructive assay. A combination of analytical techniques allows SRNL to predict the isotopic and chemical properties that qualify materials for disposition through the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). The research also defines properties that are important for other disposition paths, including disposal to the Waste Isolation Pilot Plant (WIPP) as transuranic waste (TRUW) or to high-level waste (HLW) systems. (authors)

  10. Characterizing surplus US plutonium for disposition

    SciTech Connect

    Allender, Jeffrey S.; Moore, Edwin N.

    2013-02-26

    The United States (US) has identified 61.5 metric tons (MT) of plutonium that is permanently excess to use in nuclear weapons programs, including 47.2 MT of weapons-grade plutonium. Surplus inventories will be stored safely by the Department of Energy (DOE) and then transferred to facilities that will prepare the plutonium for permanent disposition. The Savannah River National Laboratory (SRNL) operates a Feed Characterization program for the Office of Fissile Materials Disposition (OFMD) of the National Nuclear Security Administration (NNSA) and the DOE Office of Environmental Management (DOE-EM). SRNL manages a broad program of item tracking through process history, laboratory analysis, and non-destructive assay. A combination of analytical techniques allows SRNL to predict the isotopic and chemical properties that qualify materials for disposition through the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). The research also defines properties that are important for other disposition paths, including disposal to the Waste Isolation Pilot Plant (WIPP) as transuranic waste (TRUW) or to high-level waste (HLW) systems.

  11. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    SciTech Connect

    STAN, MARIUS; HECKER, SIEGFRIED S.

    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.

  12. TRACKING SURPLUS PLUTONIUM FROM WEAPONS TO DISPOSITION

    SciTech Connect

    Allender, J.; Beams, J.; Sanders, K.; Myers, L.

    2013-07-16

    Supporting nuclear nonproliferation and global security principles, beginning in 1994 the United States has withdrawn more than 50 metric tons (MT) of government-controlled plutonium from potential use in nuclear weapons. The Department of Energy (DOE), including the National Nuclear Security Administration, established protocols for the tracking of this "excess" and "surplus" plutonium, and for reconciling the current storage and utilization of the plutonium to show that its management is consistent with the withdrawal policies. Programs are underway to ensure the safe and secure disposition of the materials that formed a major part of the weapons stockpile during the Cold War, and growing quantities have been disposed as waste, after which they are not included in traditional nuclear material control and accountability (NMC&A) data systems. A combination of resources is used to perform the reconciliations that form the basis for annual reporting to DOE, to U.S. Department of State, and to international partners including the International Atomic Energy Agency.

  13. Microdosimetry of plutonium in beagle dog lung

    SciTech Connect

    Fisher, D.R.; Roesch, W.C.

    1980-08-01

    A better understanding of the microdosimetry of internally-deposited radionuclides should provide new clues to the complex relationships between organ dose distribution and early or late biological effects. Our current interest is the microdosimetry of plutonium and other alpha emitters in the lung. Since the lung is an inhomogeneous tissue, it was necessary to characterize the microscopic distributions of alveolar tissue, air space, and epithelial cell nuclei to define source-target parameters. A statistical representation of the microstructure of beagle dog lung was developed from automated image analysis of specimens from three healthy adult male dogs. The statistical distributions obtained constituted a data base from which it was possible to calculate both the energy dissipation of an alpha particle as it traversed a straight line path through pulmonary tissue, and the probability of intersecting a potentially sensitive biological site in the cell. Computer methods were modified to accomodate tissues with air space regions such as one finds in lung tissue. With the lung model description, these methods were used to determine probability density curves in specific energy for inhaled plutonium aerosols. It was assumed that the activity was randomly distributed on alveolar walls. Calculated examples are given for various activities of inhaled plutonium point sources deposited in lung tissue.

  14. Plutonium Immobilization Project -- Robotic canister loading

    SciTech Connect

    Hamilton, R.L.

    2000-01-04

    The Plutonium Immobilization Program (PIP) is a joint venture between the Savannah River Site (SRS), Lawrence Livermore National Laboratory (LLNL), Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL). When operational in 2008, the PIP will fulfill the nation's nonproliferation commitment by placing surplus weapons-grade plutonium in a permanently stable ceramic form and making it unattractive for reuse. Since there are significant radiation and security concerns, the program team is developing novel and unique technology to remotely perform plutonium immobilization tasks. The remote task covered in this paper employs a jointed arm robot to load seven 3.5 inch diameter, 135-pound cylinders (magazines) through the 4 inch diameter neck of a stainless steel canister. Working through the narrow canister neck, the robot secures the magazines into a specially designed rack pre-installed in the canister. To provide the deterrent effect, the canisters are filled with a mixture of high-level waste and glass at the Defense Waste Processing Facility (DWPF).

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

    DOEpatents

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

    2015-09-01

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

  16. Plutonium release from pressed plutonium oxide fuel pellets in aquatic environments

    SciTech Connect

    Patterson, J.H.; Steinkruger, F.J.; Matlack, G.M.; Heaton, R.C.; Coffelt, K.P.; Herrera, B.

    1983-12-01

    Plutonium oxide pellets (80% /sup 238/Pu, 40 g each) were exposed to fresh water and sea water at two temperatures for 3 y in enclosed glass chambers. The concentrations of plutonium observed in the waters increased linearly with time throughout the experiment. However, the observed release rates were inversely dependent on temperature and salinity, ranging from 160 ..mu..Ci/day for cold fresh water to 1.4 ..mu..Ci/day for warm sea water. The total releases, including the chamber residues, showed similar dependencies. A major portion (typically greater than 50%) of the released plutonium passed through a 0.1-..mu..m filter, with even larger fractions (greater than 80%) for the fresh water systems.

  17. New nuclear safe plutonium ceramic compositions with neutron poisons for plutonium storage

    NASA Astrophysics Data System (ADS)

    Nadykto, B. A.; Timofeeva, L. F.

    2000-07-01

    A complex of works is conducted to study the possibility of reprocessing surplus weapon-grade plutonium to a critical-mass-free composition with neutron poison. Nuclear safe ceramic compositions of PuO2 with four most efficient neutron poisons, Hf, Gd, Li, and B, are fabricated in the laboratory. Various methods for fabrication of the compositions with PuO2 depending on neutron poison element are used and studied: a — by sintering initial component powders; b — by impregnation of a porous skeleton made of neutron poison oxide with plutonium sol-gel; c — by sintering microspheres made of plutonium oxide with neutron poison (B4C), with the microspheres having a coating completely absorbing alpha particles.

  18. SEPARATION OF PLUTONIUM IONS FROM SOLUTION BY ADSORPTION ON ZIRCONIUM PYROPHOSPHATE

    DOEpatents

    Stoughton, R.W.

    1961-01-31

    A method is given for separating plutonium in its reduced, phosphate- insoluble state from other substances. It involves contacting a solution containing the plutonium with granular zirconium pyrophosphate.

  19. Performance of Thorium-Based Mixed Oxide Fuels for the Consumption of Plutonium in Current and Advanced Reactors

    SciTech Connect

    Weaver, Kevan Dean; Herring, James Stephen

    2003-07-01

    A renewed interest in thorium-based fuels has arisen lately based on the need for proliferation resistance, longer fuel cycles, higher burnup, and improved waste form characteristics. Recent studies have been directed toward homogeneously mixed, heterogeneously mixed, and seed-and-blanket thorium-uranium oxide fuel cycles that rely on "in situ" use of the bred-in 233U. However, due to the higher initial enrichment required to achieve acceptable burnups, these fuels are encountering economic constraints. Thorium can nevertheless play a large role in the nuclear fuel cycle, particularly in the reduction of plutonium inventories. While uranium-based mixed-oxide (MOX) fuel will decrease the amount of plutonium in discharged fuel, the reduction is limited due to the breeding of more plutonium (and higher actinides) from the 238U. Here, we present calculational results and a comparison of the potential burnup of a thorium-based and uranium-based mixed-oxide fuel in a light water reactor. Although the uranium-based fuels outperformed the thorium-based fuels in achievable burnup, a depletion comparison of the initially charged plutonium (both reactor and weapons grade) showed that the thorium-based fuels outperformed the uranium-based fuels by more that a factor of 2, where >70% of the total plutonium in the thorium-based fuel is consumed during the cycle. This is significant considering that the achievable burnups of the thorium-based fuels were 1.4 to 4.6 times less than the uranium-based fuels for similar plutonium enrichments. For equal specific burnups of ~60 MWd/kg (i.e., using variable plutonium weight percentages to give the desired burnup), the thorium-based fuels still outperform the uranium-based fuels by more than a factor of 2, where the total plutonium consumption in a three-batch, 18-month cycle was 60 to 70%. This is fairly significant considering that 10 to 15% (by weight) more plutonium is needed in the thorium-based fuels as compared to the uranium

  20. Plutonium and Cesium Colloid Mediated Transport

    NASA Astrophysics Data System (ADS)

    Boukhalfa, H.; Dittrich, T.; Reimus, P. W.; Ware, D.; Erdmann, B.; Wasserman, N. L.; Abdel-Fattah, A. I.

    2013-12-01

    Plutonium and cesium have been released to the environment at many different locations worldwide and are present in spent fuel at significant levels. Accurate understanding of the mechanisms that control their fate and transport in the environment is important for the management of contaminated sites, for forensic applications, and for the development of robust repositories for the disposal of spent nuclear fuel and nuclear waste. Plutonium, which can be present in the environment in multiple oxidations states and various chemical forms including amorphous oxy(hydr)oxide phases, adsorbs/adheres very strongly to geological materials and is usually immobile in all its chemical forms. However, when associated with natural colloids, it has the potential to migrate significant distances from its point of release. Like plutonium, cesium is not very mobile and tends to remain adhered to geological materials near its release point, although its transport can be enhanced by natural colloids. However, the reactivity of plutonium and cesium are very different, so their colloid-mediated transport might be significantly different in subsurface environments. In this study, we performed controlled experiments in two identically-prepared columns; one dedicated to Pu and natural colloid transport experiments, and the other to Cs and colloid experiments. Multiple flow-through experiments were conducted in each column, with the effluent solutions being collected and re-injected into the same column two times to examine the persistence and scaling behavior of the natural colloids, Pu and Cs. The data show that that a significant fraction of colloids were retained in the first elution through each column, but the eluted colloids collected from the first run transported almost conservatively in subsequent runs. Plutonium transport tracked natural colloids in the first run but deviated from the transport of natural colloids in the second and third runs. Cesium transport tracked natural

  1. Chemical Disposition of Plutonium in Hanford Site Tank Wastes

    SciTech Connect

    Delegard, Calvin H.; Jones, Susan A.

    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

  2. Technical and Engineering Feasibility Study of the Vitrification of Plutonium-Bearing Sludges at the Krasnoyarsk Mining and Chemical Combine by Means of Microwave Heating

    SciTech Connect

    Revenko, Y.A.; Kudinov, K.G.; Tretyakov, A.A.; Vassilyev, A.V.; Borisov, G.B.; Nazarov, A.V.; Aloy, A.S.; Shvedov, A.A.; Gusakov, B.V.; Jardine, L.J.

    2000-03-03

    This engineering feasibility study compared three possible technical options and their economic viability of processing plutonium-bearing sludges containing 0.6 MT of weapons-grade Pu accumulated at the Mining and Chemical Combine (MCC) at Krasnoyarsk. In Option 1, the baseline, the sludges are processed by extraction and purification of plutonium for storage using existing technologies, and the non-soluble radioactive residues generated in these processes undergo subsequent solidification by cementation. Options 2 and 3 involve the direct immobilization of plutonium-bearing sludges into a solid matrix (without any Pu extraction) using a microwave solidification process in a metal crucible to produce a glass, which is boron-silicate in Option 2 and phosphate glass in Option 3. In all three options, the solid radioactive waste end products will be placed in storage for eventual geologic disposal. Immobilization of residual plutonium into glass-like matrices provides both safer storage over the lifetime of the radionuclides and greater security against unauthorized access to stored materials than does the extraction and concentration of PuO{sub 2}, supporting our efforts toward non-proliferation of fissile materials. Although immobilization in boron-silicate glass appears now to be marginally preferable compared to the phosphate glass option, a number of technical issues remain to be assessed by further study to determine the preferable immobilization option.

  3. Chemical species of plutonium in Hanford radioactive tank waste

    SciTech Connect

    Barney, G.S.

    1997-10-22

    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

  4. The Borexino purification system

    NASA Astrophysics Data System (ADS)

    Benziger, Jay

    2014-05-01

    Purification of 278 tons of liquid scintillator and 889 tons of buffer shielding for the Borexino solar neutrino detector is performed with a system of combined distillation, water extraction, gas stripping and filtration. The purification system removed K, U and Th by distillation of the pseudocumene solvent and the PPO fluor. Noble gases, Rn, Kr and Ar were removed by gas stripping. Distillation was also employed to remove optical impurities and reduce the attenuation of scintillation light. The success of the purification system has facilitated the first time real time detection of low energy solar neutrinos.

  5. Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

    NASA Astrophysics Data System (ADS)

    Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

    2016-08-01

    The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel

  6. Systematic method for optimizing plutonium transmutation in LWRs

    NASA Astrophysics Data System (ADS)

    Sorensen, Reuben T.

    We have developed the Systematic Reactor Optimization in 2-Dimensions (SRO2D) code to maximize the transmutation of plutonium in light water reactors (LWRs). The necessary conditions for optimal fuel and burnable absorber loadings are obtained with Pontryagin's maximum principle and a direct adjoining approach to explicitly account for a power peaking inequality constraint. The resulting set of coupled system, Euler-Lagrange (E-L), and optimality equations are solved iteratively with the method of conjugate gradients until no further improvement is achieved in the objective function. To satisfy the power peaking inequality constraint throughout the operating cycle we have employed a backwards diffusion theory (BDT) technique as part of the conjugate gradient optimization package. The BDT approach establishes a relationship between the burnable absorber loading and the power distribution during the cycle, such that constraint violations are reduced with each conjugate gradient iteration and eventually eliminated. Our in-core optimization methodology has been implemented in the SRO2D code, assuming two-group, two-dimensional neutron diffusion theory. The system equations are solved in a quasi-static fashion forward in time from beginning-of-cycle (BOC) to end-of-cycle (EOC), while the E-L equations are solved backwards in time from EOC to BOC to reflect the adjoint nature of the Lagrange multipliers. Cycle length extension calculations of a first cycle AP600 plant verify our implementation effort, yielding a nearly identical loading pattern to that issued by Westinghouse in the AP600 Safety Analysis Report. Utilizing a self-generated Pu recycling mode, our in-core optimization methodology is coupled with an equilibrium cycle methodology to arrive at an optimized asymptotic Pu inventory and composition. Beginning with a poor loading pattern, our LWR optimization package improves the core performance by reducing the maximum power peaking factor from 2.0 to 1

  7. Fifty years of plutonium exposure to the Manhattan Project plutonium workers: an update.

    PubMed

    Voelz, G L; Lawrence, J N; Johnson, E R

    1997-10-01

    Twenty-six white male workers who did the original plutonium research and development work at Los Alamos have been examined periodically over the past 50 y to identify possible health effects from internal plutonium depositions. Their effective doses range from 0.1 to 7.2 Sv with a median value of 1.25 Sv. As of the end of 1994, 7 individuals have died compared with an expected 16 deaths based on mortality rates of U.S. white males in the general population. The standardized mortality ratio (SMR) is 0.43. When compared with 876 unexposed Los Alamos workers of the same period, the plutonium worker's mortality rate was also not elevated (SMR = 0.77). The 19 living persons have diseases and physical changes characteristic of a male population with a median age of 72 y (range = 69 to 86 y). Eight of the twenty-six workers have been diagnosed as having one or more cancers, which is within the expected range. The underlying cause of death in three of the seven deceased persons was from cancer, namely cancer of prostate, lung, and bone. Mortality from all cancers was not statistically elevated. The effective doses from plutonium to these individuals are compared with current radiation protection guidelines.

  8. Fifty years of plutonium exposure to the Mahattan Project plutonium workers: An update

    SciTech Connect

    Voelz, G.L.; Lawrence, J.N.P.; Johnson, E.R.

    1997-10-01

    Twenty-six white male workers who did the original plutonium research and development work at Los Alamos have been examined periodically over the past 50 y to identify possible health effects from internal plutonium depositions. Their effective doses range from 0.1 to 7.2 Sv with a median value of 1.25 Sv. As of the end of 1994, 7 individuals have died compared with an expected 16 deaths based on mortality rates of U.S. white males in the general population. The standardized mortality ratio (SMR) is 0.43. When compared with 876 unexposed Los Alamos workers of the same period, the plutonium worker`s mortality rate was also not elevated (SMR = 0.77). The 19 living persons have diseases and physical changes characteristic of a male population with a median age of 72 y (range = 69 to 86 y). Eight of the twenty-six workers have been diagnosed as having one or more cancers, which is within the expected range. The underlying cause of death in three of the seven deceased persons was from cancer, namely cancer of prostate, lung, and bone. Mortality from all cancers was not statistically elevated. The effective doses from plutonium to these individuals are compared with current radiation protection guidelines. 28 refs., 5 tabs.

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

    PubMed

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

    2014-08-11

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

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

    SciTech Connect

    Kogan, V.; Schumacher, P.M.

    1993-12-01

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

  11. 10. VIEW OF THE INSTALLATION OF PLUTONIUM FABRICATION ROLLING MILL. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    10. VIEW OF THE INSTALLATION OF PLUTONIUM FABRICATION ROLLING MILL. THE MILL ROLLED INGOTS INTO SHEETS THAT WERE THEN CUT INTO CIRCLE BLANKS TO BE PASSED THROUGH THE CENTER LINE FOR PRESSING. (2/19/63) - Rocky Flats Plant, Plutonium Fabrication, Central section of Plant, Golden, Jefferson County, CO

  12. 26. Plutonium Recovery From Contaminated Materials, Architectural Elevations, Sections & ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    26. Plutonium Recovery From Contaminated Materials, Architectural Elevations, Sections & Dets., Building 232-Z, U.S. Atomic Energy Commission, Hanford Atomic Products Operation, General Electric Company, Dwg. No. H-2-23106, 1959. - Plutonium Finishing Plant, Waste Incinerator Facility, 200 West Area, Richland, Benton County, WA

  13. 25. Plutonium Recovery From Contaminated Materials, Architectural Plans & Details, ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    25. Plutonium Recovery From Contaminated Materials, Architectural Plans & Details, Building 232-Z, U.S. Atomic Energy Commission, Hanford Atomic Products Operation, General Electric Company, Dwg. No. H-2-23105, 1959. - Plutonium Finishing Plant, Waste Incinerator Facility, 200 West Area, Richland, Benton County, WA

  14. Processing of Non-PFP Plutonium Oxide in Hanford Plants

    SciTech Connect

    Jones, Susan A.; Delegard, Calvin H.

    2011-03-10

    Processing of non-irradiated plutonium oxide, PuO2, scrap for recovery of plutonium values occurred routinely at Hanford’s Plutonium Finishing Plant (PFP) in glovebox line operations. Plutonium oxide is difficult to dissolve, particularly if it has been high-fired; i.e., calcined to temperatures above about 400°C and much of it was. Dissolution of the PuO2 in the scrap typically was performed in PFP’s Miscellaneous Treatment line using nitric acid (HNO3) containing some source of fluoride ion, F-, such as hydrofluoric acid (HF), sodium fluoride (NaF), or calcium fluoride (CaF2). The HNO3 concentration generally was 6 M or higher whereas the fluoride concentration was ~0.5 M or lower. At higher fluoride concentrations, plutonium fluoride (PuF4) would precipitate, thus limiting the plutonium dissolution. Some plutonium-bearing scrap also contained PuF4 and thus required no added fluoride. Once the plutonium scrap was dissolved, the excess fluoride was complexed with aluminum ion, Al3+, added as aluminum nitrate, Al(NO3)3•9H2O, to limit collateral damage to the process equipment by the corrosive fluoride. Aluminum nitrate also was added in low quantities in processing PuF4.

  15. A plutonium-based single-molecule magnet.

    PubMed

    Magnani, N; Colineau, E; Griveau, J-C; Apostolidis, C; Walter, O; Caciuffo, R

    2014-08-04

    The magnetic properties of the 5f(5) [tris-(tri-1-pyrazolylborato)-plutonium(III)] complex have been investigated by ac susceptibility measurements, showing it to be the first plutonium single-molecule magnet; its magnetic relaxation slows down with decreasing temperature through a thermally activated mechanism followed by a quantum tunnelling regime below 5 K.

  16. SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS BY ADSORPTION

    DOEpatents

    Seaborg, G.T.; Willard, J.E.

    1958-01-01

    A method is presented for the separation of plutonium from solutions containing that element in a valence state not higher than 41 together with uranium ions and fission products. This separation is accomplished by contacting the solutions with diatomaceous earth which preferentially adsorbs the plutonium present. Also mentioned as effective for this adsorbtive separation are silica gel, filler's earth and alumina.

  17. METHOD OF SEPARATION OF PLUTONIUM FROM CARRIER PRECIPITATES

    DOEpatents

    Dawson, I.R.

    1959-09-22

    The recovery of plutonium from fluoride carrier precipitates is described. The precipitate is dissolved in zirconyl nitrate, ferric nitrate, aluminum nitrate, or a mixture of these complexing agents, and the plutonium is then extracted from the aqueous solution formed with a water-immiscible organic solvent.

  18. Fuel bundle design for enhanced usage of plutonium fuel

    DOEpatents

    Reese, Anthony P.; Stachowski, Russell E.

    1995-01-01

    A nuclear fuel bundle includes a square array of fuel rods each having a concentration of enriched uranium and plutonium. Each rod of an interior array of the rods also has a concentration of gadolinium. The interior array of rods is surrounded by an exterior array of rods void of gadolinium. By this design, usage of plutonium in the nuclear reactor is enhanced.

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

    SciTech Connect

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

    1996-09-01

    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.

  20. COMPLEX FLUORIDES OF PLUTONIUM AND AN ALKALI METAL

    DOEpatents

    Seaborg, G.T.

    1960-08-01

    A method is given for precipitating alkali metal plutonium fluorides. such as KPuF/sub 5/, KPu/sub 2/F/sub 9/, NaPuF/sub 5/, and RbPuF/sub 5/, from an aqueous plutonium(IV) solution by adding hydrogen fluoride and alkali-metal- fluoride.

  1. 10 CFR 71.88 - Air transport of plutonium.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Air transport of plutonium. 71.88 Section 71.88 Energy... Controls and Procedures § 71.88 Air transport of plutonium. (a) Notwithstanding the provisions of any..., whether for import, export, or domestic shipment, is not transported by air or delivered to a carrier...

  2. 10 CFR 71.88 - Air transport of plutonium.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Air transport of plutonium. 71.88 Section 71.88 Energy... Controls and Procedures § 71.88 Air transport of plutonium. (a) Notwithstanding the provisions of any..., whether for import, export, or domestic shipment, is not transported by air or delivered to a carrier...

  3. 10 CFR 71.88 - Air transport of plutonium.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Air transport of plutonium. 71.88 Section 71.88 Energy... Controls and Procedures § 71.88 Air transport of plutonium. (a) Notwithstanding the provisions of any..., whether for import, export, or domestic shipment, is not transported by air or delivered to a carrier...

  4. Plutonium and Cs-137 in autopsy tissues in Great Britain.

    PubMed

    Popplewell, D S; Ham, G J; Dodd, N J; Shuttler, S D

    1988-03-01

    Tissues removed at autopsy from members of the general public contain significantly higher concentrations of plutonium and 137Cs in west Cumbrians than in people from three other regions of Great Britain. Several autopsy cases from Cumbria showed unusually high values of plutonium. Subsequently it was found that the subjects had been former employees of British Nuclear Fuels.

  5. METHOD FOR OBTAINING PLUTONIUM METAL FROM ITS TRICHLORIDE

    DOEpatents

    Reavis, J.G.; Leary, J.A.; Maraman, W.J.

    1962-08-14

    A method was developed for obtaining plutonium metal by direct reduction of plutonium chloride, without the use of a booster, using calcium and lanthamum as a reductant, the said reduction being carried out at temperature in the range of 700 to 850 deg C and at about atmospheric pressure. (AEC)

  6. PRECIPITATION METHOD OF SEPARATING PLUTONIUM FROM CONTAMINATING ELEMENTS

    DOEpatents

    Duffield, R.B.

    1959-02-24

    S>A method is described for separating plutonium, in a valence state of less than five, from an aqueous solution in which it is dissolved. The niethod consists in adding potassium and sulfate ions to such a solution while maintaining the solution at a pH of less than 7.1, and isolating the precipitate of potassium plutonium sulfate thus formed.

  7. Removal of plutonium and americium from alkaline waste solutions

    DOEpatents

    Schulz, Wallace W.

    1979-01-01

    High salt content, alkaline waste solutions containing plutonium and americium are contacted with a sodium titanate compound to effect removal of the plutonium and americium from the alkaline waste solution onto the sodium titanate and provide an effluent having a radiation level of less than 10 nCi per gram alpha emitters.

  8. Density of Plutonium Turnings Generated from Machining Activities

    SciTech Connect

    Gonzales, John Robert; Vigil, Duane M.; Jachimowski, Thomas A.; Archuleta, Alonso; Arellano, Gerald Joseph; Melton, Vince Lee

    2016-10-20

    The purpose of this project was to determine the density of plutonium (Pu) turnings generated from the range of machining activities, using both surrogate material and machined Pu turnings. Verify that 500 grams (g) of plutonium will fit in a one quart container using a surrogate equivalent volume and that 100 grams of Pu will fit in a one quart Savy container.

  9. Plutonium finishing plant safety systems and equipment list

    SciTech Connect

    Bergquist, G.G.

    1995-01-06

    The Safety Equipment List (SEL) supports Analysis Report (FSAR), WHC-SD-CP-SAR-021 and the Plutonium Finishing Plant Operational Safety Requirements (OSRs), WHC-SD-CP-OSR-010. The SEL is a breakdown and classification of all Safety Class 1, 2, and 3 equipment, components, or system at the Plutonium Finishing Plant complex.

  10. SEPARATION OF PLUTONIUM VALUES FROM URANIUM AND FISSION PRODUCT VALUES

    DOEpatents

    Maddock, A.G.; Booth, A.H.

    1960-09-13

    Separation of plutonium present in small amounts from neutron irradiated uranium by making use of the phenomenon of chemisorption is described. Plutonium in the tetravalent state is chemically absorbed on a fluoride in solid form. The steps for the separation comprise dissolving the irradiated uranium in nitric acid, oxidizing the plutonium in the resulting solution to the hexavalent state, adding to the solution a soluble calcium salt which by the common ion effect inhibits dissolution of the fluoride by the solution, passing the solution through a bed or column of subdivided calcium fluoride which has been sintered to about 8OO deg C to remove the chemisorbable fission products, reducing the plutonium in the solution thus obtained to the tetravalent state, and again passing the solution through a similar bed or column of calcium fluoride to selectively absorb the plutonium, which may then be recovered by treating the calcium fluoride with a solution of ammonium oxalate.

  11. BASIC PEROXIDE PRECIPITATION METHOD OF SEPARATING PLUTONIUM FROM CONTAMINANTS

    DOEpatents

    Seaborg, G.T.; Perlman, I.

    1959-02-10

    A process is described for the separation from each other of uranyl values, tetravalent plutonium values and fission products contained in an aqueous acidic solution. First the pH of the solution is adjusted to between 2.5 and 8 and hydrogen peroxide is then added to the solution causing precipitation of uranium peroxide which carries any plutonium values present, while the fission products remain in solution. Separation of the uranium and plutonium values is then effected by dissolving the peroxide precipitate in an acidic solution and incorporating a second carrier precipitate, selective for plutonium. The plutonium values are thus carried from the solution while the uranium remains flissolved. The second carrier precipitate may be selected from among the group consisting of rare earth fluorides, and oxalates, zirconium phosphate, and bismuth lihosphate.

  12. 23. AERIAL VIEW LOOKING SOUTHEAST AT THE PLUTONIUM OPERATION BUILDINGS ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    23. AERIAL VIEW LOOKING SOUTHEAST AT THE PLUTONIUM OPERATION BUILDINGS 771, 776/777, AND 707. BUILDING 771, IN THE FOREGROUND, WAS BUILT IN 1952 TO HOUSE ALL PLUTONIUM OPERATIONS. BY 1956, BUILDING 771 WAS NO LONGER ADEQUATE FOR PRODUCTION DEMANDS. BUILDING 776/777, TO THE SOUTH OF BUILDING 771, WAS CONSTRUCTED TO HOUSE PLUTONIUM FABRICATION AND FOUNDRY OPERATIONS. PLUTONIUM RECOVERY REMAINED IN BUILDING 771. BY 1967, CONSTRUCTION ON BUILDING 707, TO THE SOUTH OF BUILDING 776/777, BEGAN AS PRODUCTION LEVELS CONTINUED TO EXPAND NECESSITATING THE NEED FOR ADDITIONAL PLUTONIUM FABRICATION SPACE (7/1/69). - Rocky Flats Plant, Bounded by Indiana Street & Routes 93, 128 & 72, Golden, Jefferson County, CO

  13. PROCESS FOR PRODUCTION OF PLUTONIUM FROM ITS OXIDES

    DOEpatents

    Weissman, S.I.; Perlman, M.L.; Lipkin, D.

    1959-10-13

    A method is described for obtaining a carbide of plutonium and two methods for obtaining plutonium metal from its oxides. One of the latter involves heating the oxide, in particular PuO/sub 2/, to a temperature of 1200 to 1500 deg C with the stoichiometrical amount of carbon to fornn CO in a hard vacuum (3 to 10 microns Hg), the reduced and vaporized plutonium being collected on a condensing surface above the reaction crucible. When an excess of carbon is used with the PuO/sub 2/, a carbide of plutonium is formed at a crucible temperature of 1400 to 1500 deg C. The process may be halted and the carbide removed, or the reaction temperature can be increased to 1900 to 2100 deg C at the same low pressure to dissociate the carbide, in which case the plutonium is distilled out and collected on the same condensing surface.

  14. Pyrochemical recovery of plutonium from calcium fluoride reduction slag

    DOEpatents

    Christensen, D.C.

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

  15. Preparation of Pure Plutonium Metal Standards for Nondestructive Assay

    SciTech Connect

    S. -T. Hsue; J. E. Stewart; M. S. Krick

    2000-11-01

    To calibrate neutron coincidence and neutron multiplicity counters for passive assay of plutonium, certain detector parameters must be determined. When one is using small plutonium metal samples, biases can be introduced from non-zero multiplication and impurities. This paper describes preparing small, pure plutonium metal standards with well-known geometries to enable accurate multiplication corrections and with acceptably low levels of impurities. To minimize multiplication, these standards are designed as 2-cm-diameter foils with varying thicknesses and masses of 1.4, 3.6, and 7.2 g plutonium. These standards will significantly improve characterization and calibration of neutron coincidence and multiplicity counters. They can also be equally useful for gamma-ray spectrometry and calorimetry. Five sets will be made: four for other US Department of Energy plutonium facilities, and one set to remain at Los Alamos. We will also describe other nondestructive assay standards that are planned for the next few years.

  16. Plutonium immobilization ceramic feed batching component test report

    SciTech Connect

    Erickson, S.A.

    1999-10-04

    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 high level waste glass for permanent storage. Ceramic feed batching (CFB) is one of the first process steps involved with first stage plutonium immobilization. The CFB step will blend plutonium oxide powder before it is combined with other materials to make pucks. This report discusses the Plutonium Immobilization CFB process preliminary concept (including a process block diagram), batch splitting component test results, CFB development areas, and FY 1999 and 2000 CFB program milestones.

  17. Modelling the distribution of plutonium in the Pacific Ocean.

    PubMed

    Nakano, Masanao; Povinec, Pavel P

    2003-01-01

    An Oceanic General Circulation Model (OGCM) including a plutonium scavenging model as well as an advection-diffusion model has been developed for modelling the distribution of plutonium in the Pacific Ocean. Calculated 239, 240Pu water profile concentrations and 239, 240Pu inventories in water and sediment of the Pacific Ocean have showed a reasonable agreement with the experimental results. The presence of local fallout plutonium in central North Pacific waters has been confirmed. The observed 240Pu/239Pu mass ratios confirm that plutonium originating from local fallout from nuclear weapons tests carried out at Bikini and Enewetak Atolls is more rapidly removed from surface waters to deeper waters than plutonium originating from global fallout. The developed OGCM can be used for modelling the dispersion of other non-conservative tracers in the ocean as well.

  18. Uncertainties on lung doses from inhaled plutonium.

    PubMed

    Puncher, Matthew; Birchall, Alan; Bull, Richard K

    2011-10-01

    In a recent epidemiological study, Bayesian uncertainties on lung doses have been calculated to determine lung cancer risk from occupational exposures to plutonium. These calculations used a revised version of the Human Respiratory Tract Model (HRTM) published by the ICRP. In addition to the Bayesian analyses, which give probability distributions of doses, point estimates of doses (single estimates without uncertainty) were also provided for that study using the existing HRTM as it is described in ICRP Publication 66; these are to be used in a preliminary analysis of risk. To infer the differences between the point estimates and Bayesian uncertainty analyses, this paper applies the methodology to former workers of the United Kingdom Atomic Energy Authority (UKAEA), who constituted a subset of the study cohort. The resulting probability distributions of lung doses are compared with the point estimates obtained for each worker. It is shown that mean posterior lung doses are around two- to fourfold higher than point estimates and that uncertainties on doses vary over a wide range, greater than two orders of magnitude for some lung tissues. In addition, we demonstrate that uncertainties on the parameter values, rather than the model structure, are largely responsible for these effects. Of these it appears to be the parameters describing absorption from the lungs to blood that have the greatest impact on estimates of lung doses from urine bioassay. Therefore, accurate determination of the chemical form of inhaled plutonium and the absorption parameter values for these materials is important for obtaining reliable estimates of lung doses and hence risk from occupational exposures to plutonium.

  19. Development of characterization of plutonium storage containers

    SciTech Connect

    James, D.; Stevkovski, S.

    1999-02-01

    As a result of the end of the Cold War, at least 11,000 (possibly 20,000 or more) plutonium pits are projected to be stored at Pantex for up to fifty years. The current pit container, the ALR8 was not designed for this length of storage duration. As a result, Pantex officials have searched for alternative container options. The objective of this research is to develop and validate a model to predict the temperature distribution within the stored components and the internal structure of the proposed ALR8(SI) container, and to consider and analyze the safety features of the ALR8(SI) container as seen from the thermal performance view. Due to the time scale involved with the current simulations, the radioactive decay of the plutonium may be assumed to provide a uniform rate of heat generation. This heat is conducted to the surroundings through the solid structures of the assembly. In addition to conduction, the inert gas that fills the volume within the steel container convects a fraction of the generated heat from the plutonium to the colder steel surfaces. Radiation must also be accounted for as natural convection and limited conduction paths are present within the container. The research efforts in this project have been directed into two paths, numerical and experimental. First, the temperature distribution within the stored components are being determined experimentally as a function of fill gases, energy generation rate, and boundary conditions. Second, a finite element model of the ALR8 container has been developed so that the temperature distribution can be predicted as a function of the same experimental parameters. This paper presents the experimental method and data that have been obtained thus far, as well as the finite element model created using SDRC I-DEAS.

  20. Massive subcritical compact arrays of plutonium metal

    SciTech Connect

    Rothe, R.E.

    1998-04-01

    Two experimental critical-approach programs are reported. Both were performed at the Rocky Flats Plant near Denver, Colorado; and both date back to the late 1960s. Both involve very large arrays of massive plutonium ingots. These ingots had been cast in the foundry at the Rocky Flats Plant as part of their routine production operations; they were not specially prepared for either study. Consequently, considerable variation in ingot mass is encountered. This mass varied between approximately 7 kg and a little more than 10 kg. One program, performed in the spring of 1969, involved stacked arrays of ingots contained within cylindrical, disk-shaped, thin, steel cans. This program studied four arrays defined by the pattern of steel cans in a single layer. The four were: 1 x N, 3 x N, 2 x 2 x N, and 3 x 3 x N. The second was a tightly-packed, triangular-pitched patterns; the last two were square-pitched patterns. The other program, performed about a year earlier, involved similar ingots also contained in similar steel cans, but these canned plutonium ingots were placed in commercial steel drums. This study pertained to one-, two-, and three-layered horizontal arrays of drums. All cases proved to be well subcritical. Most would have remained subcritical had the parameters of the array under study been continued infinitely beyond the reciprocal multiplication safety limit. In one case for the drum arrays, an uncertain extrapolation of the data of the earlier program suggests that criticality might have eventually been attained had several thousand additional kilograms of plutonium been available for use.

  1. Thermal Stability Studies of Candidate Decontamination Agents for Hanford’s Plutonium Finishing Plant Plutonium-Contaminated Gloveboxes

    SciTech Connect

    Scheele, Randall D.; Cooper, Thurman D.; Jones, Susan A.; Ewalt, John R.; Compton, James A.; Trent, Donald S.; Edwards, Matthew K.; Kozelisky, Anne E.; Scott, Paul A.; Minette, Michael J.

    2005-09-29

    This report provides the results of PNNL's and Fluor's studies of the thermal stabilities of potential wastes arising from decontamination of Hanford's Plutonium Finishing Plant's plutonium contaminated gloveboxes. The candidate wastes arising from the decontamination technologies ceric nitrate/nitric acid, RadPro, Glygel, and Aspigel.

  2. Plutonium Hexaboride is a Correlated Topological Insulator

    NASA Astrophysics Data System (ADS)

    Deng, Xiaoyu; Haule, Kristjan; Kotliar, Gabriel

    2013-10-01

    We predict that plutonium hexaboride (PuB6) is a strongly correlated topological insulator, with Pu in an intermediate valence state of Pu2.7+. Within the combination of dynamical mean field theory and density functional theory, we show that PuB6 is an insulator in the bulk, with nontrivial Z2 topological invariants. Its metallic surface states have a large Fermi pocket at the X¯ point and the Dirac cones inside the bulk derived electronic states, causing a large surface thermal conductivity. PuB6 has also a very high melting temperature; therefore, it has ideal solid state properties for a nuclear fuel material.

  3. Modeling thermal properties of plutonium mononitride

    NASA Astrophysics Data System (ADS)

    Yu, H. L.; Huang, H.; Li, G.; Li, H. B.; Meng, D. Q.

    2015-06-01

    The thermal properties of plutonium mononitride (PuN) were investigated by molecular dynamics method. The interatomic potentials of PuN were fitted by using Chen-Möbius multiple lattice inversion technique. Based on these interatomic potentials, the lattice constant, bulk modulus, compressibility, cohesive energy and heat capacity of PuN were obtained and the results are well consistent with experimental data and previous reports. It indicates that the potentials we build in this study are effective for studying thermal properties of PuN.

  4. Plutonium hexaboride is a correlated topological insulator.

    PubMed

    Deng, Xiaoyu; Haule, Kristjan; Kotliar, Gabriel

    2013-10-25

    We predict that plutonium hexaboride (PuB(6)) is a strongly correlated topological insulator, with Pu in an intermediate valence state of Pu(2.7+). Within the combination of dynamical mean field theory and density functional theory, we show that PuB(6) is an insulator in the bulk, with nontrivial Z(2) topological invariants. Its metallic surface states have a large Fermi pocket at the X[over ¯] point and the Dirac cones inside the bulk derived electronic states, causing a large surface thermal conductivity. PuB(6) has also a very high melting temperature; therefore, it has ideal solid state properties for a nuclear fuel material.

  5. Plutonium hexaboride is a correlated topological insulator

    NASA Astrophysics Data System (ADS)

    Deng, Xiaoyu; Haule, Kristjan; Kotliar, Gabriel; Department of Physics and Astronomy, Rutgers University Team

    2014-03-01

    We predict that plutonium hexaboride (PuB6) is a strongly correlated topological insulator, with Pu in an intermediate valence state of Pu2 . 7 +. Within the combination of dynamical mean field theory and density functional theory, we show that PuB6 is an insulator in the bulk, with non-trivial Z2 topological invariants. Its metallic surface states have large Fermi pocket at X point and the Dirac cones inside the bulk derived electronic states causing a large surface thermal conductivity. PB6 has also a very high melting temperature therefore it has ideal solid state properties for a nuclear fuel material.

  6. PLUTONIUM-238 PRODUCTION TARGET DESIGN STUDIES

    SciTech Connect

    Hurt, Christopher J; Wham, Robert M; Hobbs, Randall W; Owens, R Steven; Chandler, David; Freels, James D; Maldonado, G Ivan

    2014-01-01

    A new supply chain is planned for plutonium-238 using existing reactors at the Oak Ridge National Laboratory (ORNL) and Idaho National Laboratory (INL) and existing chemical recovery facilities at ORNL. Validation and testing activities for new irradiation target designs have been conducted in three phases over a 2 year period to provide data for scale-up to production. Target design, qualification, target fabrication, and irradiation of fully-loaded targets have been accomplished. Data from post-irradiation examination (PIE) supports safety analysis and irradiation of future target designs.

  7. Multiple quantum phase transitions of plutonium compounds

    SciTech Connect

    Matsumoto, Munehisa; Yin, Quan; Otsuki, Junya; Savrasov, Sergey Y.

    2011-07-22

    We show by quantum Monte Carlo simulations of realistic Kondo lattice models derived from electronicstructure calculations that multiple quantum critical points can be realized in plutonium-based materials. We place representative systems, including PuCoGa5, on a realistic Doniach phase diagram and identify the regions where the magnetically mediated superconductivity could occur. The solution of an inverse problem to restore the quasiparticle renormalization factor for f electrons is shown to be sufficiently good to predict the trends among Sommerfeld coefficients and magnetism. A suggestion on the possible experimental verification for this scenario is given for PuAs.

  8. Plutonium contamination in the environment. September 1977-November 1989 (A Bibliography from the Selected Water Resources Abstracts data base). Report for September 1977-November 1989

    SciTech Connect

    Not Available

    1990-05-01

    This bibliography contains citations concerning the ecological impact of plutonium contamination in the environment. Topics include plutonium contamination in freshwater and marine sediments, plutonium bioaccumulation, plutonium transport in the food chain, plutonium contamination bioindicators, methods of analysis, plutonium genotoxicity, plutonium contamination in soil and groundwater, and plutonium contamination from nuclear fallout and nuclear facilities. Plutonium distribution changes due to stratification in oxic and anoxic environments are described. (Contains 83 citations fully indexed and including a title list.)

  9. Solvent extraction system for plutonium colloids and other oxide nano-particles

    DOEpatents

    Soderholm, Lynda; Wilson, Richard E; Chiarizia, Renato; Skanthakumar, Suntharalingam

    2014-06-03

    The invention provides a method for extracting plutonium from spent nuclear fuel, the method comprising supplying plutonium in a first aqueous phase; contacting the plutonium aqueous phase with a mixture of a dielectric and a moiety having a first acidity so as to allow the plutonium to substantially extract into the mixture; and contacting the extracted plutonium with second a aqueous phase, wherein the second aqueous phase has a second acidity higher than the first acidity, so as to allow the extracted plutonium to extract into the second aqueous phase. The invented method facilitates isolation of plutonium polymer without the formation of crud or unwanted emulsions.

  10. Direct reduction of actinide oxide and carbide to metal: Application to the preparation of plutonium metal

    NASA Astrophysics Data System (ADS)

    Spirlet, J. C.; Müller, W.; Van Audenhove, J.

    1985-06-01

    Three different conversion and refining methods for the preparation of high purity plutonium metal are described: the calciothermic reduction of plutonium oxide followed by electrorefining; the thoriothermic reduction of plutonium oxide followed by selective evaporation; the tantalothermic reduction of plutonium carbide followed by selective evaporation. The calciothermic reduction of plutonium oxide followed by electrorefining is used for the semi-industrial or large scale production of high purity plutonium metal. The rate and yield of preparation and refining is high. With high purity reactants the reduction of the oxide with thorium is well adapted to obtain high purity plutonium metal on the laboratory scale. The tantalothermic reduction of plutonium carbide gives high purity plutonium metal even with impure plutonium starting material (recovered from waste). This results from the high selectivity at the different steps of the process.

  11. Plutonium Detection with Straw Neutron Detectors

    SciTech Connect

    Mukhopadhyay, Sanjoy; Maurer, Richard; Guss, Paul

    2014-03-27

    A kilogram of weapons grade plutonium gives off about 56,000 neutrons per second of which 55,000 neutrons come from spontaneous fission of 240Pu (~6% by weight of the total plutonium). Actually, all even numbered isotopes (238Pu, 240Pu, and 242Pu) produce copious spontaneous fission neutrons. These neutrons induce fission in the surrounding fissile 239Pu with an approximate multiplication of a factor of ~1.9. This multiplication depends on the shape of the fissile materials and the surrounding material. These neutrons (typically of energy 2 MeV and air scattering mean free path >100 meters) can be detected 100 meters away from the source by vehicle-portable neutron detectors. [1] In our current studies on neutron detection techniques, without using 3He gas proportional counters, we designed and developed a portable high-efficiency neutron multiplicity counter using 10B-coated thin tubes called straws. The detector was designed to perform like commercially available fission meters (manufactured by Ortec Corp.) except instead of using 3He gas as a neutron conversion material, we used a thin coating of 10B.

  12. Expected radiation effects in plutonium immobilization ceramic

    SciTech Connect

    Van Konynenburg, R.A., LLNL

    1997-09-01

    The current formulation of the candidate ceramic for plutonium immobilization consists primarily of pyrochlore, with smaller amounts of hafnium-zirconolite, rutile, and brannerite or perovskite. At a plutonium loading of 10.5 weight %, this ceramic would be made metamict (amorphous) by radiation damage resulting from alpha decay in a time much less than 10,000 years, the actual time depending on the repository temperature as a function of time. Based on previous experimental radiation damage work by others, it seems clear that this process would also result in a bulk volume increase (swelling) of about 6% for ceramic that was mechanically unconfined. For the candidate ceramic, which is made by cold pressing and sintering and has porosity amounting to somewhat more than this amount, it seems likely that this swelling would be accommodated by filling in the porosity, if the material were tightly confined mechanically by the waste package. Some ceramics have been observed to undergo microcracking as a result of radiation-induced anisotropic or differential swelling. It is unlikely that the candidate ceramic will microcrack extensively, for three reasons: (1) its phase composition is dominated by a single matrix mineral phase, pyrochlore, which has a cubic crystal structure and is thus not subject to anisotropic swelling; (2) the proportion of minor phases is small, minimizing potential cracking due to differential swelling; and (3) there is some flexibility in sintering process parameters that will allow limitation of the grain size, which can further limit stresses resulting from either cause.

  13. Management of disused plutonium sealed sources

    SciTech Connect

    Whitworth, Julia Rose; Pearson, Michael W; Abeyta, Cristy

    2010-01-01

    The Global Threat Reduction Initiative's (GTRI) Offsite Source Recovery Project (OSRP) has been recovering excess and unwanted radioactive sealed sources since 1999, including more than 2,400 Plutonium (Pu)-238 sealed sources and 653 Pu-239-bearing sources that represent more than 10% of the total sources recovered by GTRI/OSRP to date. These sources have been recovered from hundreds of sites within the United States (US) and around the world. OSRP grew out of early efforts at the Los Alamos National Laboratory (LANL) to recover and disposition excess Plutonium-239 (Pu-239) sealed sources that were distributed in the 1960s and 1970s under the Atoms for Peace Program, a loan-lease program that serviced 31 countries, as well as domestic users. In the conduct of these recovery operations, GTRI/OSRP has been required to solve problems related to knowledge-of-inventory, packaging and transportation of fissile and heat-source materials, transfer of ownership, storage of special nuclear material (SNM) both at US Department of Energy (DOE) facilities and commercially, and disposal. Unique issues associated with repatriation from foreign countries, including end user agreements required by some European countries and denials of shipment, will also be discussed.

  14. AMS of the Minor Plutonium Isotopes

    PubMed Central

    Steier, P.; Hrnecek, E.; Priller, A.; Quinto, F.; Srncik, M.; Wallner, A.; Wallner, G.; Winkler, S.

    2013-01-01

    VERA, the Vienna Environmental Research Accelerator, is especially equipped for the measurement of actinides, and performs a growing number of measurements on environmental samples. While AMS is not the optimum method for each particular plutonium isotope, the possibility to measure 239Pu, 240Pu, 241Pu, 242Pu and 244Pu on the same AMS sputter target is a great simplification. We have obtained a first result on the global fallout value of 244Pu/239Pu = (5.7 ± 1.0) × 10−5 based on soil samples from Salzburg prefecture, Austria. Furthermore, we suggest using the 242Pu/240Pu ratio as an estimate of the initial 241Pu/239Pu ratio, which allows dating of the time of irradiation based solely on Pu isotopes. We have checked the validity of this estimate using literature data, simulations, and environmental samples from soil from the Salzburg prefecture (Austria), from the shut down Garigliano Nuclear Power Plant (Sessa Aurunca, Italy) and from the Irish Sea near the Sellafield nuclear facility. The maximum deviation of the estimated dates from the expected ages is 6 years, while relative dating of material from the same source seems to be possible with a precision of less than 2 years. Additional information carried by the minor plutonium isotopes may allow further improvements of the precision of the method. PMID:23565016

  15. Avoided valence transition in a plutonium superconductor.

    PubMed

    Ramshaw, B J; Shekhter, Arkady; McDonald, Ross D; Betts, Jon B; Mitchell, J N; Tobash, P H; Mielke, C H; Bauer, E D; Migliori, Albert

    2015-03-17

    The d and f electrons in correlated metals are often neither fully localized around their host nuclei nor fully itinerant. This localized/itinerant duality underlies the correlated electronic states of the high-Tc cuprate superconductors and the heavy-fermion intermetallics and is nowhere more apparent than in the 5f valence electrons of plutonium. Here, we report the full set of symmetry-resolved elastic moduli of PuCoGa5--the highest Tc superconductor of the heavy fermions (Tc = 18.5 K)--and find that the bulk modulus softens anomalously over a wide range in temperature above Tc. The elastic symmetry channel in which this softening occurs is characteristic of a valence instability--therefore, we identify the elastic softening with fluctuations of the plutonium 5f mixed-valence state. These valence fluctuations disappear when the superconducting gap opens at Tc, suggesting that electrons near the Fermi surface play an essential role in the mixed-valence physics of this system and that PuCoGa5 avoids a valence transition by entering the superconducting state. The lack of magnetism in PuCoGa5 has made it difficult to reconcile with most other heavy-fermion superconductors, where superconductivity is generally believed to be mediated by magnetic fluctuations. Our observations suggest that valence fluctuations play a critical role in the unusually high Tc of PuCoGa5.

  16. AMS of the Minor Plutonium Isotopes.

    PubMed

    Steier, P; Hrnecek, E; Priller, A; Quinto, F; Srncik, M; Wallner, A; Wallner, G; Winkler, S

    2013-01-01

    VERA, the Vienna Environmental Research Accelerator, is especially equipped for the measurement of actinides, and performs a growing number of measurements on environmental samples. While AMS is not the optimum method for each particular plutonium isotope, the possibility to measure (239)Pu, (240)Pu, (241)Pu, (242)Pu and (244)Pu on the same AMS sputter target is a great simplification. We have obtained a first result on the global fallout value of (244)Pu/(239)Pu = (5.7 ± 1.0) × 10(-5) based on soil samples from Salzburg prefecture, Austria. Furthermore, we suggest using the (242)Pu/(240)Pu ratio as an estimate of the initial (241)Pu/(239)Pu ratio, which allows dating of the time of irradiation based solely on Pu isotopes. We have checked the validity of this estimate using literature data, simulations, and environmental samples from soil from the Salzburg prefecture (Austria), from the shut down Garigliano Nuclear Power Plant (Sessa Aurunca, Italy) and from the Irish Sea near the Sellafield nuclear facility. The maximum deviation of the estimated dates from the expected ages is 6 years, while relative dating of material from the same source seems to be possible with a precision of less than 2 years. Additional information carried by the minor plutonium isotopes may allow further improvements of the precision of the method.

  17. Collector for recovering gallium from weapons plutonium

    SciTech Connect

    Philip, C.V.; Anthony, R.G.; Chokkaram, S.

    1998-09-01

    Currently, the separation of gallium from weapons plutonium involves the use of aqueous processing using either solvent extraction of ion exchange. However, this process generates significant quantities of liquid radioactive wastes. A Thermally Induced Gallium Removal process, or TIGR, developed by researchers at Los Alamos National Laboratories, is a simpler alternative to aqueous processing. This research examined this process, and the behavior of gallium suboxide, a vapor that is swept away by passing hydrogen/argon over gallium trioxide/plutonium oxide heated at 1100 C during the TIGR process. Through experimental procedures, efforts were made to prevent the deposition of corrosive gallium onto furnace and vent surfaces. Experimental procedures included three options for gallium removal and collection: (1) collection of gallium suboxide through use of a cold finger; (2) collection by in situ air oxidation; and (3) collection of gallium on copper. Results conclude all three collection mechanisms are feasible. In addition, gallium trioxide exists in three crystalline forms, and each form was encountered during each experiment, and that each form will have a different reactivity.

  18. Co-Design: Fabrication of Unalloyed Plutonium

    SciTech Connect

    Korzekwa, Deniece R.; Knapp, Cameron M.; Korzekwa, David A.; Gibbs, John W

    2012-07-25

    The successful induction casting of plutonium is a challenge which requires technical expertise in areas including physical metallurgy, surface and corrosion chemistry, materials science, electromagnetic engineering and a host of other technologies all which must be applied in concert. Here at LANL, we are employing a combined experimental and computational approach to design molds and develop process parameters needed to produce desired temperature profiles and improved castings. Computer simulations are performed using the commercial code FLOW-3D and the LANL ASC computer code TRUCHAS to reproduce the entire casting process starting with electromagnetic or radiative heating of the mold and metal and continuing through pouring with coupled fluid flow, heat transfer and non-isothermal solidification. This approach greatly reduces the time required to develop a new casting designs and also increases our understanding of the casting process, leading to a more homogeneous, consistent product and better process control. We will discuss recent casting development results in support of unalloyed plutonium rods for mechanical testing.

  19. Avoided valence transition in a plutonium superconductor

    PubMed Central

    Ramshaw, B. J.; Shekhter, Arkady; McDonald, Ross D.; Betts, Jon B.; Mitchell, J. N.; Tobash, P. H.; Mielke, C. H.; Bauer, E. D.; Migliori, Albert

    2015-01-01

    The d and f electrons in correlated metals are often neither fully localized around their host nuclei nor fully itinerant. This localized/itinerant duality underlies the correlated electronic states of the high-Tc cuprate superconductors and the heavy-fermion intermetallics and is nowhere more apparent than in the 5f valence electrons of plutonium. Here, we report the full set of symmetry-resolved elastic moduli of PuCoGa5—the highest Tc superconductor of the heavy fermions (Tc = 18.5 K)—and find that the bulk modulus softens anomalously over a wide range in temperature above Tc. The elastic symmetry channel in which this softening occurs is characteristic of a valence instability—therefore, we identify the elastic softening with fluctuations of the plutonium 5f mixed-valence state. These valence fluctuations disappear when the superconducting gap opens at Tc, suggesting that electrons near the Fermi surface play an essential role in the mixed-valence physics of this system and that PuCoGa5 avoids a valence transition by entering the superconducting state. The lack of magnetism in PuCoGa5 has made it difficult to reconcile with most other heavy-fermion superconductors, where superconductivity is generally believed to be mediated by magnetic fluctuations. Our observations suggest that valence fluctuations play a critical role in the unusually high Tc of PuCoGa5. PMID:25737548

  20. Measurements of plutonium residues from recovery processes

    SciTech Connect

    Hsue, S.-T.; Langner, D.G.; Longmire, V.L.; Menlove, H.O.; Russo, P.A.; Sprinkle, J.K. Jr.

    1989-01-01

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

  1. Plutonium immobilization plant using glass in existing facilities at the Savannah River Site

    SciTech Connect

    DiSabatino, A., LLNL

    1998-06-01

    The Plutonium Immobilization Plant (PIP) accepts plutonium (Pu) from pit conversion and from non-pit sources and, through a glass immobilization process, converts the plutonium into an immobilized form that can be disposed of in a high level waste (HLW) repository. 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.

  2. Determination of plutonium oxidation states in dilute nitric acid by complementary tristimulus colorimetry.

    PubMed

    Silver, G L

    1967-06-01

    The preparation of reference standards for use in complementary tristimulus colorimetry for plutonium is described. Plutonium(III) and (VI) are prepared by hydrazine reduction and silver(II) oxidation, respectively, of plutonium(IV). Plutonium(V) is prepared by reduction of plutonium(VI) with ascorbic or sulphurous acid. A method for computerizing tristimulus colorimetry is presented, and the technique is extended to three dimensions ("quadristimulus colorimetry").

  3. How much plutonium does North Korea really have?

    SciTech Connect

    Dreicer, J.S.

    1997-11-01

    In a previous study, as part of the Global Nuclear Material Control Model effort, the author estimated the maximum quantity of plutonium that could be produced in thermal research reactors in the potential nuclear weapon states (including North Korea), based on their declared power level. D. Albright has estimated the amount of plutonium the North Koreans may have produced since 1986 in the 5-megawatt-electric power reactor at Yongbon. Albright provided an upper-bound estimate of 53 kilograms of weapon-grade plutonium produced cumulatively if the gas-graphite (magnox) reactor had achieved a load factor of 0.80. This cumulative estimate of 53 kilograms ignores the potential plutonium production in the 8-megawatt-thermal research reactor, IRT-DPRK. To better quantify the cumulative North Korean production, the author conducted a study to estimate the amount of plutonium that could have been produced in the IRT-DPRK research reactor operating at the declared power level during the entire period it has operated, including a period it was not safeguarded. The author estimates that, at most, an additional 6 to 33 kilograms of plutonium could have been produced cumulatively in the research reactor operating at the declared power level during the entire period it has operated, including a 12-year period it was not safeguarded, resulting in a total of 13 to 47 kilograms of plutonium possibly produced in both the research and power reactors.

  4. Colloid-Facilitated Plutonium Transport in Fractured Tuffaceous Rock.

    PubMed

    Wolfsberg, Andrew; Dai, Zhenxue; Zhu, Lin; Reimus, Paul; Xiao, Ting; Ware, Doug

    2017-05-16

    Colloids have the potential to enhance the mobility of strongly sorbing radionuclide contaminants in groundwater at underground nuclear test sites. This study presents an experimental and numerical investigation of colloid-facilitated plutonium transport in fractured porous media to identify plutonium reactive transport processes. The transport parameters for dispersion, diffusion, sorption, and filtration are estimated with inverse modeling by minimizing the least-squares objective function of multicomponent concentration data from multiple transport experiments with the shuffled complex evolution metropolis algorithm. Capitalizing on an unplanned experimental artifact that led to colloid formation, we adopt a stepwise strategy to first interpret the data from each experiment separately and then to incorporate multiple experiments simultaneously to identify a suite of plutonium-colloid transport processes. Nonequilibrium or kinetic attachment and detachment of plutonium-colloid in fractures were clearly demonstrated and captured in the inverted modeling parameters along with estimates of the source plutonium fraction that formed plutonium-colloids. The results from this study provide valuable insights for understanding the transport mechanisms and environmental impacts of plutonium in groundwater aquifers.

  5. Plutonium recovery from spent reactor fuel by uranium displacement

    DOEpatents

    Ackerman, J.P.

    1992-03-17

    A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

  6. Plutonium recovery from spent reactor fuel by uranium displacement

    DOEpatents

    Ackerman, John P.

    1992-01-01

    A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

  7. A comparative assessment of the economics of plutonium disposition

    SciTech Connect

    Williams, K.A.; Miller, J.W.; Reid, R.L.

    1997-04-01

    The US Department of Energy office of Fissile Materials Disposition (DOE/MD) has been evaluating three technologies for the disposition of approximately 50 metric tons of surplus plutonium from defense-related programs: reactors, immobilization, and deep boreholes. As part of the process supporting an early CY 1997 Record of Decision (ROD), a comprehensive assessment of technical viability, cost, and schedule has been conducted by DOE/MD and its national laboratory contractors. Oak Ridge National Laboratory has managed and coordinated the life-cycle cost (LCC) assessment effort for this program. This paper discusses the economic analysis methodology and the results prior to ROD. A secondary intent of the paper is to discuss major technical and economic issues that impact cost and schedule. To evaluate the economics of these technologies on an equitable basis, a set of cost-estimating guidelines and a common cost-estimating format were utilized by all three technology teams. This paper also includes the major economic analysis assumptions and the comparative constant-dollar and discounted-dollar LCCs.

  8. Fallout plutonium in two oxic-anoxic environments

    SciTech Connect

    Sanchez, A.L.; Murray, J.W.; Schell, W.R.; Miller, L.G.

    1986-09-01

    The profiles of soluble fallout plutonium in two partially anoxic waters revealed minimum concentrations at the O/sub 2/-H/sub 2/S interface, indicating Pu removal onto particulate phases of Fe and other oxidized species that form during the redox cycle. In Saanich Inlet, an intermittently anoxic fjord in Vancouver Island, Canada, the concentration of soluble Pu in the anoxic zone was slightly less than in the oxygenated surface layer. In Soap Lake, a saline meromictic lake in eastern Washington State, Pu concentrations i the permanently anoxic zone were at least an order of magnitude higher than at the surface. Differences in the chemical characteristics of these two waters suggest important chemical species that influenced the observed Pu distribution. In the permanently anoxic zone of Soap Lake, high values of total alkalinity ranging from 940 to 1500 meq liter/sup -1/, sulfide species from 38 to 128 ..mu..M, dissolved organic carbon from 163 to 237 mg liter/sup -1/, and total dissolved solids from 80 to 140 ppt, all correlated with the observed high concentration of Pu. In Saanich Inlet, where total alkalinity ranged from 2.1 to 2.4 meq liter/sup -1/ and salinity from 25 to 32 per thousand and H/sub 2/S concentration in May 1981 showed a maximum of 8..mu..M, the observed Pu concentrations were significantly lower than for the Soap Lake monimolimnion.

  9. Plutonium and tritium produced in the Hanford Site production reactors

    SciTech Connect

    Roblyer, S.P.

    1994-09-28

    In a news release on December 7, 1993, the Secretary of Energy announced declassification action that included totals for plutonium and tritium production in the Hanford Site production reactors. This information was reported as being preliminary because it was not fully supported by documentation. Subsequently, production data were made available from the US Department of Energy-Headquarters (DOE-HQ) records that indicated an increase of about one and one-half metric tons in total plutonium production. The Westinghouse Hanford Company was tasked by the US Department of Energy-Richland Operations Office to substantiate production figures and DOE-HQ data and to provide a defensible report of weapons- (6 wt% {sup 240}Pu) and nonweapons- (fuels-)grade (nominally 9 wt% or higher {sup 240}Pu) plutonium and tritium production in the Hanford Site production reactors. The task was divided into three parts. The first part was to determine plutonium and tritium production based on available reported accountability records. The second part was to determine plutonium production independently by calculational checks based on reactor thermal power generation and plutonium conversion factors representing the various reactor fuels. The third part was to resolve differences, if they occurred, in the reported and calculational results. In summary, the DOE-HQ-reported accountability records of plutonium and tritium production were determined to be the most defensible record of Hanford Site reactor production. The DOE-HQ records were consistently supported by the independent calculational checks and the records of operational data. Total production quantities are 67.4 MT total plutonium, which includes 12.9 MT of nonweapons-grade plutonium. The total tritium production was 10.6 kg.

  10. Succinonitrile Purification Facility

    NASA Technical Reports Server (NTRS)

    2003-01-01

    The Succinonitrile (SCN) Purification Facility provides succinonitrile and succinonitrile alloys to several NRA selected investigations for flight and ground research at various levels of purity. The purification process employed includes both distillation and zone refining. Once the appropriate purification process is completed, samples are characterized to determine the liquidus and/or solidus temperature, which is then related to sample purity. The lab has various methods for measuring these temperatures with accuracies in the milliKelvin to tenths of milliKelvin range. The ultra-pure SCN produced in our facility is indistinguishable from the standard material provided by NIST to well within the stated +/- 1.5mK of the NIST triple point cells. In addition to delivering material to various investigations, our current activities include process improvement, characterization of impurities and triple point cell design and development. The purification process is being evaluated for each of the four vendors to determine the efficacy of each purification step. We are also collecting samples of the remainder from distillation and zone refining for analysis of the constituent impurities. The large triple point cells developed will contain SCN with a melting point of 58.0642 C +/- 1.5mK for use as a calibration standard for Standard Platinum Resistance Thermometers (SPRTs).

  11. Succinonitrile Purification Facility

    NASA Technical Reports Server (NTRS)

    2003-01-01

    The Succinonitrile (SCN) Purification Facility provides succinonitrile and succinonitrile alloys to several NRA selected investigations for flight and ground research at various levels of purity. The purification process employed includes both distillation and zone refining. Once the appropriate purification process is completed, samples are characterized to determine the liquidus and/or solidus temperature, which is then related to sample purity. The lab has various methods for measuring these temperatures with accuracies in the milliKelvin to tenths of milliKelvin range. The ultra-pure SCN produced in our facility is indistinguishable from the standard material provided by NIST to well within the stated +/- 1.5mK of the NIST triple point cells. In addition to delivering material to various investigations, our current activities include process improvement, characterization of impurities and triple point cell design and development. The purification process is being evaluated for each of the four vendors to determine the efficacy of each purification step. We are also collecting samples of the remainder from distillation and zone refining for analysis of the constituent impurities. The large triple point cells developed will contain SCN with a melting point of 58.0642 C +/- 1.5mK for use as a calibration standard for Standard Platinum Resistance Thermometers (SPRTs).

  12. Ribonucleic acid purification.

    PubMed

    Martins, R; Queiroz, J A; Sousa, F

    2014-08-15

    Research on RNA has led to many important biological discoveries and improvement of therapeutic technologies. From basic to applied research, many procedures employ pure and intact RNA molecules; however their isolation and purification are critical steps because of the easy degradability of RNA, which can impair chemical stability and biological functionality. The current techniques to isolate and purify RNA molecules still have several limitations and the requirement for new methods able to improve RNA quality to meet regulatory demands is growing. In fact, as basic research improves the understanding of biological roles of RNAs, the biopharmaceutical industry starts to focus on them as a biotherapeutic tools. Chromatographic bioseparation is a high selective unit operation and is the major option in the purification of biological compounds, requiring high purity degree. In addition, its application in biopharmaceutical manufacturing is well established. This paper discusses the importance and the progress of RNA isolation and purification, considering RNA applicability both in research and clinical fields. In particular and in view of the high specificity, affinity chromatography has been recently applied to RNA purification processes. Accordingly, recent chromatographic investigations based on biorecognition phenomena occurring between RNA and amino acids are focused. Histidine and arginine have been used as amino acid ligands, and their ability to isolate different RNA species demonstrated a multipurpose applicability in molecular biology analysis and RNA therapeutics preparation, highlighting the potential contribution of these methods to overcome the challenges of RNA purification.

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

    SciTech Connect

    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/sub 4/-H/sub 2/SO/sub 4/ mixture. The dried residue is dissolved in dilute H/sub 2/SO/sub 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. Purification of genuine multipartite entanglement

    SciTech Connect

    Huber, Marcus; Plesch, Martin

    2011-06-15

    In tasks where multipartite entanglement plays a central role, state purification is, due to inevitable noise, a crucial part of the procedure. We consider a scenario exploiting the multipartite entanglement in a straightforward multipartite purification algorithm and compare it to bipartite purification procedures combined with state teleportation. While complete purification requires an infinite amount of input states in both cases, we show that for an imperfect output fidelity the multipartite procedure exhibits a major advantage in terms of input states used.

  15. Plutonium stabilization and handling quality assurance program plan

    SciTech Connect

    Weiss, E.V.

    1998-04-22

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

  16. METHOD OF SEPARATING PLUTONIUM FROM LANTHANUM FLUORIDE CARRIER

    DOEpatents

    Watt, G.W.; Goeckermann, R.H.

    1958-06-10

    An improvement in oxidation-reduction type methods of separating plutoniunn from elements associated with it in a neutron-irradiated uranium solution is described. The method relates to the separating of plutonium from lanthanum ions in an aqueous 0.5 to 2.5 N nitric acid solution by 'treating the solution, at room temperature, with ammonium sulfite in an amount sufficient to reduce the hexavalent plutonium present to a lower valence state, and then treating the solution with H/sub 2/O/sub 2/ thereby forming a tetravalent plutonium peroxide precipitate.

  17. Amarillo National Resource Center for Plutonium 1999 plan

    SciTech Connect

    1999-01-30

    The purpose of the Amarillo National Resource Center for Plutonium is to serve the Texas Panhandle, the State of Texas and the US Department of Energy by: conducting scientific and technical research; advising decision makers; and providing information on nuclear weapons materials and related environment, safety, health, and nonproliferation issues while building academic excellence in science and technology. This paper describes the electronic resource library which provides the national archives of technical, policy, historical, and educational information on plutonium. Research projects related to the following topics are described: Environmental restoration and protection; Safety and health; Waste management; Education; Training; Instrumentation development; Materials science; Plutonium processing and handling; and Storage.

  18. Development of first ever scanning probe microscopy capabilities for plutonium

    DOE PAGES

    Beaux, Miles F.; Cordoba, Miguel Santiago; Zocco, Adam T.; ...

    2017-04-01

    Scanning probe microscopy capabilities have been developed for plutonium and its derivative compounds. Specifically, a scanning tunneling microscope and an atomic force microscope housed in an ultra-high vacuum system and an inert atmosphere glove box, respectively, were prepared for the introduction of small non-dispersible δ-Pu coupons. Experimental details, procedures, and preliminary imaging of δ-Pu coupons are presented to demonstrate the functionality of these new capabilities. In conclusion, these first of a kind capabilities for plutonium represent a significant step forward in the ability to characterize and understand plutonium surfaces with high spatial resolution.

  19. Status of immobilization of excess weapons plutonium in Russia

    SciTech Connect

    Borisov, G B; Jardine, L; Mansourov, O A

    1999-02-03

    In this paper, we examine the logic and framework for the development of a capability to immobilize excess Russian weapons plutonium by the year 2004. The initial activities underway in Russia, summarized here, include engineering feasibility studies of the immobilization of plutonium-containing materials at the Krasnoyarsk and Mayak industrial sites. In addition, research and development (R&D) studies are underway at Russian institutes to develop glass and ceramic forms suitable for the immobilization of plutonium-containing materials, residues, and wastes and for their geologic disposal.

  20. Accelerator-driven assembly for plutonium transformation (ADAPT)

    NASA Astrophysics Data System (ADS)

    Tuyle, Greorgy J. Van; Todosow, Michael; Powell, James; Schweitzer, Donald

    1995-01-01

    A particle accelerator-driven spallation target and corresponding blanket region are proposed for the ultimate disposition of weapons-grade plutonium being retired from excess nuclear weapons in the U.S. and Russia. The highly fissle plutonium is contained within .25 to .5 cm diameter silicon-carbide coated graphite beads, which are cooled by helium, within the slightly subcritical blanket region. Major advantages include very high one-pass burnup (over 90%), a high integrity waste form (the coated beads), and operation in a subcritical mode, thereby minimizing the vulnerability to the positive reativity feedbacks often associated with plutonium fuel.