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Sample records for pu u-oxide fuel

  1. ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

    SciTech Connect

    Lell, R. M.; Schaefer, R. W.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; Nuclear Engineering Division; Inst. of Physics and Power Engineering

    2007-10-01

    Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of demonstration

  2. ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

    SciTech Connect

    Lell, R. M.; Morman, J. A.; Schaefer, R.W.; McKnight, R.D.; Nuclear Engineering Division

    2009-02-23

    ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide, U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium

  3. Pu-238 fuel form activities, June 1-30, 1980

    SciTech Connect

    Not Available

    1980-07-18

    This monthly report for Pu-238 Fuel Form Activities has two main sections: SRP-PuFF Pu-238 Fuel Form Production Processes and SRL Pu-238 Fuel Form Research and Development. The program status, budget information, and milestone information are discussed in each main section. The Work Breakdown Structures (WBS) for this program is outlined. Only one monthly report per year is processed for EDB.

  4. Pu-238 fuel form activities, January 1-31, 1982

    SciTech Connect

    Not Available

    1982-03-01

    This monthly report for /sup 238/Pu fuel form activities has two main sections: SRP-PuFF facility and SRL fuel form activities. The program status, budget information, and milestone schedules are discussed in each main section. The Work Breakdown Structure (WBS) for this program is shown. Only one monthly report per year is processed for EDB.

  5. Improved MOX fuel calculations using new Pu-239, Am-241 and Pu-240 evaluations

    NASA Astrophysics Data System (ADS)

    Noguere, G.; Bouland, O.; Bernard, D.; Leconte, P.; Blaise, P.; Peneliau, Y.; Vidal, J. F.; De Saint Jean, C.; Leal, L.; Schillebeeckx, P.; Kopecky, S.; Lampoudis, C.

    2013-03-01

    Several studies based on the JEFF-3.1.1 nuclear data library show a systematic overestimation of the critical keff for core configurations of MOX fuel assemblies. The present work investigates possible improvements of the C/E results by using new evaluations for Am-241, Pu-239 and Pu-240.

  6. Performance of Cladding on MOX Fuel with Low 240Pu/239Pu Ratio

    SciTech Connect

    McCoy, Kevin; Blanpain, Patrick; Morris, Robert Noel

    2014-01-01

    The U.S. Department of Energy has decided to dispose of a portion of its surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. As part of fuel qualification, four lead assemblies were manufactured and irradiated to a maximum fuel rod average burnup of 47.3 MWd/kg heavy metal. This was the world s first commercial irradiation of MOX fuel with a 240Pu/239Pu ratio less than 0.10. Five fuel rods with varying burnups and plutonium contents were selected from one of the assemblies and shipped to Oak Ridge National Laboratory for hot cell examination. This paper discusses the results of those examinations with emphasis on cladding performance. Exams relevant to the cladding included visual and eddy current exams, profilometry, microscopy, hydrogen analysis, gallium analysis, and mechanical testing. There was no discernible effect of the type of MOX fuel on the performance of the cladding.

  7. Fuel-Cladding Interaction Between U-Pu-Zr Fuel and Fe

    NASA Astrophysics Data System (ADS)

    Aitkaliyeva, Assel; Madden, James W.; Miller, Brandon D.; Papesch, Cynthia A.; Cole, James I.

    2015-12-01

    This work investigates fuel-cladding chemical interaction (FCCI) between U-25Pu-14Zr (in wt pct) fuel and pure Fe at elevated temperatures, understanding of which is critical for evaluation of the fuel performance. Phases and microstructure formed in the quaternary uranium-plutonium-zirconium-iron (U-Pu-Zr-Fe) system were characterized using the transmission electron microscopy technique. Phases formed within the FCCI layer were identified using selective area electron diffraction (SAED) analysis as Fe2U (Fd- m), Fe2Zr (Fd-3m), α-U (Cmcm), Fe2Pu (Fd-3m), β-Pu (C12/m1), and β-Zr (Im-3m).

  8. Thermal Analysis of ZPPR High Pu Content Stored Fuel

    DOE PAGES

    Solbrig, Charles W.; Pope, Chad L.; Andrus, Jason P.

    2014-09-17

    The Zero Power Physics Reactor (ZPPR) operated from April 18, 1969, until 1990. ZPPR operated at low power for testing nuclear reactor designs. This paper examines the temperature of Pu content ZPPR fuel while it is in storage. Heat is generated in the fuel due to Pu and Am decay and is a concern for possible cladding damage. Damage to the cladding could lead to fuel hydriding and oxidizing. A series of computer simulations were made to determine the range of temperatures potentially occuring in the ZPPR fuel. The maximum calculated fuel temperature is 292°C (558°F). Conservative assumptions in themore » model intentionally overestimate temperatures. The stored fuel temperatures are dependent on the distribution of fuel in the surrounding storage compartments, the heat generation rate of the fuel, and the orientation of fuel. Direct fuel temperatures could not be measured but storage bin doors, storage sleeve doors, and storage canister temperatures were measured. Comparison of these three temperatures to the calculations indicates that the temperatures calculated with conservative assumptions are, as expected, higher than the actual temperatures. The maximum calculated fuel temperature with the most conservative assumptions is significantly below the fuel failure criterion of 600°C (1,112°F).« less

  9. Bulk characterization of (U, Pu) mixed carbide fuel for distribution of plutonium

    SciTech Connect

    Devi, K. V. Vrinda Khan, K. B.; Biju, K.; Kumar, Arun

    2015-06-24

    Homogeneous distribution of plutonium in (U, Pu) mixed fuels is important from fuel performance as well as reprocessing point of view. Radiation imaging and assay techniques are employed for the detection of Pu rich agglomerates in the fuel. A simulation study of radiation transport was carried out to analyse the technique of autoradiography so as to estimate the minimum detectability of Pu agglomerates in MC fuel with nominal PuC content of 70% using Monte Carlo simulations.

  10. Method of removing Pu(IV) polymer from nuclear fuel reclaiming liquid

    DOEpatents

    Tallent, Othar K.; Mailen, James C.; Bell, Jimmy T.; Arwood, Phillip C.

    1982-01-01

    A Pu(IV) polymer not extractable from a nuclear fuel reclaiming solution by conventional processes is electrolytically converted to Pu.sup.3+ and PuO.sub.2.sup.2+ ions which are subsequently converted to Pu.sup.4+ ions extractable by the conventional processes.

  11. Cracking mechanism of PuO/sub 2/ fuel hot pressed in graphite dies

    SciTech Connect

    Taylor, D.H.

    1981-01-01

    Internal cracking in PuO/sub 2/ fuel is caused by gas pressure during hot pressing. Surface cracks are caused by tensile stresses arising from the phase change of ..cap alpha..-PuO/sub 2/ to ..gamma..-PuO/sub 2/ on reoxidation. To control cracking, process variables were chosen to minimize fuel reduction and to give large intergranular porosity. 19 figures.

  12. Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

    NASA Astrophysics Data System (ADS)

    Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

    2016-04-01

    This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.

  13. Oxidation and reduction behaviors of a prototypic MgO-PuO2-x inert matrix fuel

    NASA Astrophysics Data System (ADS)

    Miwa, Shuhei; Osaka, Masahiko

    2017-04-01

    Oxidation and reduction behaviors of prototypic MgO-based inert matrix fuels (IMFs) containing PuO2-x were experimentally investigated by means of thermogravimetry. The oxidation and reduction kinetics of the MgO-PuO2-x specimen were determined. The oxidation and reduction rates of the MgO-PuO2-x were found to be low compared with those of PuO2-x. It is note that the changes in O/Pu ratios of MgO-PuO2-x from stoichiometry were smaller than those of PuO2-x at high oxygen partial pressure.

  14. TEM identification of subsurface phases in ternary U-Pu-Zr fuel

    NASA Astrophysics Data System (ADS)

    Aitkaliyeva, Assel; Madden, James W.; Papesch, Cynthia A.; Cole, James I.

    2016-05-01

    Phases and microstructure in as-cast, annealed at 850 °C, and subsequently cooled U-23Pu-9Zr fuel were characterized using scanning and transmission electron microscopy techniques. SEM examination shows formation of three phases in the alloy that were identified in TEM using selective area diffraction pattern analysis: α-Zr globular and elongated δ-UZr2 inclusions and a thick oxide layer formed on top of β-Pu phase, which has been initially assumed to be ζ-(U, Pu). However, further examination of the cross-sectional TEM specimens identified the matrix phases as δ-UZr2, β-Pu, and (U, Zr)ht. Two types of inclusions were observed in the immediate vicinity of the specimen surface and they were consistent with α-Zr and ζ-(U, Pu).

  15. Characterization of phases formed between U-Pu-Mo fuels and Fe-12Cr cladding

    NASA Astrophysics Data System (ADS)

    Aitkaliyeva, Assel; Madden, James W.; Miller, Brandon D.; Papesch, Cynthia A.; Cole, James I.

    2015-09-01

    Exposure to high temperatures and irradiation can lead to interaction between fuel and cladding constituents, inter-diffusion, and formation of brittle or low-melting phases. Therefore, understanding of fuel-cladding interaction (FCCI) is critical for evaluation of fuel performance in a reactor environment. In this contribution, phases formed between U-22Pu-4Mo and U-25Pu-15Mo (in wt%) fuel alloys and Fe-12Cr cladding were characterized using scanning and transmission electron microscopy (SEM/TEM) techniques. Phases formed within FCCI layers in both alloys were identified by implementing selective area diffraction pattern analysis as Cr0.3Mo0.7 (Im-3m), Fe2U (Fd-3m), UCrFe (Fd-3m), and Fe2Pu (Fd-3m). Phases formed at the end of the FCCI layer in the U-22Pu-4Mo alloy included UCrFe (Fd-3m), Fe2U (Fd-3m), and Cr2FeO4 (Fd-3m) while in the U-25Pu-15Mo alloy the phases were consistent with Cr0.49Fe0.51 (P42/mnm), Cr0.8Fe0.2 (Im-3m), and UCrFe (Fd-3m).

  16. TEM examination of phases formed between U-Pu-Zr fuel and Fe

    NASA Astrophysics Data System (ADS)

    Aitkaliyeva, Assel; Madden, James W.; Miller, Brandon D.; Papesch, Cynthia A.; Cole, James I.

    2015-12-01

    Exposure to high temperatures and irradiation results in interaction and interdiffusion between fuel and cladding constituents that can lead to formation of undesirable brittle or low-melting point phases. A diffusion couple study has been conducted to understand fuel-cladding interaction occurring between U-22Pu-4Zr (in wt%) fuel and pure Fe at elevated temperatures. The phases formed within fuel cladding chemical interaction (FCCI) layer have been characterized in the transmission electron microscope (TEM). The phases formed within FCCI layer have been identified as Fe2U (Fd-3m), FeU6 (I4/mcm), Fe2Zr (Fd-3m), FeZr2 (I4/mcm), Fe2Pu (Fd-3m), UZr2 (P6/mmm), β-Zr (Im-3m), and ZrO2 (Fm-3m).

  17. Interdiffusion between U-Pu-Zr fuel and HT9 cladding

    SciTech Connect

    Keiser, D.D. Jr.; Petri, M.C.

    1994-06-01

    As part of systematic interdiffusion studies of fuel-cladding compatibility in the integral Fast Reactor, a solid-solid diffusion couple was assembled with U-22Pu-23{sup 1} Zr fuel and HT9{sup 2} cladding and annealed at 650{degrees}C for 100 hours. The couple was examined for diffusion structure development using a scanning electron microscope equipped with an energy dispersive x-ray analyzer (SEM-EDX). Point-by-point and linescan analysis was used to generate composition profiles and diffusion paths. From the composition profiles, average effective interdiffusion coefficients were calculated for individual components on both sides of the Matano plane. Results from this investigation indicate that the same types of phases as would be expected from binary U-Fe, Pu-Fe, and Zr-Fe phase diagrams develop in this couple; and U and Pu are the fastest diffusing fuel components and Fe is the fastest diffusing cladding component. Compared with diffusion couples with binary (U-Zr) fuel, the addition of Pu greatly enhanced the extent of diffusion and affected the types of phases observed.

  18. Irradiation performance of U-Pu-Zr metal fuels for liquid-metal-cooled reactors

    SciTech Connect

    Tsai, H.; Cohen, A.B.; Billone, M.C.; Neimark, L.A.

    1994-10-01

    This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result`s from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of {approx}75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of {approx}1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of {approx}11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction.

  19. Progress on High Energy Delayed Gamma Spectroscopy for Direct Assay of Pu in Spent Fuel

    SciTech Connect

    Campbell, Luke W.; Smith, Leon E.

    2010-08-11

    The direct, nondestructive measurement of fissile and fissionable isotopes in spent fuel is not yet possible. Current methods which infer plutonium content through proxy measurements and confirmatory burnup calculations have relatively large uncertainty and do not satisfy the desire for a measurement that is independent of operator declarations. We are currently exploring the High Energy Delayed Gamma Spectroscopy (HEDGS) technique for direct, independent Pu measurement in light-water reactor fuels. HEDGS exploits the unique distribution of fission-product nuclei from each of the fissile isotopes. Fission is stimulated in the sample with a source of interrogating neutrons, and delayed gamma rays from the decay of the short-lived fission-product nuclei are measured. The measured gamma spectrum from the unknown sample is then fit with a linear combination of gamma spectra from pure U-235, Pu-239, and Pu-241, as deduced from the known fission-product yield curves and decay properties of the fission-product nuclei, to determine the original proportions of these fissile isotopes. In previous work, we performed preliminary modeling studies of HEDGS on idealized single fuel pins of various burnups. Here, we report progress on extending our GEANT-based modeling tools to efficiently model full pressurized water reactor (PWR) fuel assemblies using variance reduction techniques specific to the background emissions and induced signal, as appropriate. Predicted performance for a nominal HEDGS instrument design, is reported for the assay of U-235, Pu-239 and Pu-241 in spent fuel assemblies ranging from fresh to 60 GWd/MTU in burnup.

  20. Stress and Diffusion in Stored Pu ZPPR Fuel from Alpha Generation

    SciTech Connect

    Charles W. Solbrig; Chad L. Pope; Jason P. Andrus

    2014-07-01

    ZPPR (Zero Power Physics Reactor) is a research reactor that has been used to investigate breeder reactor fuel designs. The reactor has been dismantled but its fuel is still stored there. Of concern are its plutonium containing metal fuel elements which are enclosed in stainless steel cladding with gas space filled with helium–argon gas and welded air tight. The fuel elements which are 5.08 cm by 0.508 cm up to 20.32 cm long (2 in × 0.2 in × 8 in) were manufactured in 1968. A few of these fuel elements have failed releasing contamination raising concern about the general state of the large number of other fuel elements. Inspection of the large number of fuel elements could lead to contamination release so analytical studies have been conducted to estimate the probability of failed fuel elements. This paper investigates the possible fuel failures due to generation of helium in the metal fuel from the decay of Pu and its possible damage to the fuel cladding from metal fuel expansion or from diffusion of helium into the fuel gas space. This paper (1) calculates the initial gas loading in a fuel element and its internal free volume after it has been brought into the atmosphere at ZPPR, (2) shows that the amount of helium generated by decay of Pu over 46 years since manufacture is significantly greater than this initial loading, (3) determines the amount of fuel swelling if the helium stays fixed in the fuel plate and estimates the amount of helium which diffuses out of the fuel plate into the fuel plenum assuming the helium does not remain fixed in the fuel plate but can diffuse to the plenum and possibly through the cladding. Since the literature is not clear as to which possibility occurs, as with Schroedinger’s cat, both possibilities are analyzed. The paper concludes that (1) if the gas generated is fixed in the fuel, then the fuel swelling it can cause would not cause any fuel failure and (2) if the helium does diffuse out of the fuel (in accordance

  1. TRISO-Fuel Element Performance Modeling for the Hybrid LIFE Engine with Pu Fuel Blanket

    SciTech Connect

    DeMange, P; Marian, J; Caro, M; Caro, A

    2010-02-18

    A TRISO-coated fuel thermo-mechanical performance study is performed for the hybrid LIFE engine to test the viability of TRISO particles to achieve ultra-high burnup of a weapons-grade Pu blanket. Our methodology includes full elastic anisotropy, time and temperature varying material properties for all TRISO layers, and a procedure to remap the elastic solutions in order to achieve fast fluences up to 30 x 10{sup 25} n {center_dot} m{sup -2} (E > 0.18 MeV). In order to model fast fluences in the range of {approx} 7 {approx} 30 x 10{sup 25} n {center_dot} m{sup -2}, for which no data exist, careful scalings and extrapolations of the known TRISO material properties are carried out under a number of potential scenarios. A number of findings can be extracted from our study. First, failure of the internal pyrolytic carbon (PyC) layer occurs within the first two months of operation. Then, the particles behave as BISO-coated particles, with the internal pressure being withstood directly by the SiC layer. Later, after 1.6 years, the remaining PyC crumbles due to void swelling and the fuel particle becomes a single-SiC-layer particle. Unrestrained by the PyC layers, and at the temperatures and fluences in the LIFE engine, the SiC layer maintains reasonably-low tensile stresses until the end-of-life. Second, the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Obtaining more reliable measurements, especially at higher fluences, is an imperative for the fidelity of our models. Finally, varying the geometry of the TRISO-coated fuel particles results in little differences in the scope of fuel performance. The mechanical integrity of 2-cm graphite pebbles that act as fuel matrix has also been studied and it is concluded that they can reliable serve the entire LIFE burnup cycle without failure.

  2. /sup 238/Pu fuel form processes. Bimonthly report, November-December 1979

    SciTech Connect

    Folger, R. L.

    1980-11-01

    Progress in the Savannah River Laboratory's /sup 238/Pu Fuel Form Program is summarized. Full-scale fabrication tests continued in the Plutonium Experimental Facility (PEF) with the successful fabrication of seven additional GPHS pellets. Three pellets (GPHS Pellets 14, 15, and 16) were fabricated at off-centerline conditions to help define process limits for production of GPHS fuel pellets in the Plutonium Fuel Fabrication (PuFF) Facility. Two additional limit-test pellets (GPHS Pellets 12 and 13) previously hot pressed underwent final heat treatment. Two pellets (GPHS Pellets 17 and 18) were fabricated at centerline conditions as part of the effort to have Savannah River Laboratory (SRL) GPHS pellets impact tested at LASL. All seven pellets remained integral and demonstrated excellent dimensional stability during final heat treatment. However, the quality of those pellets fabricated at centerline conditions was superior to those that were fabricated as part of the limit tests.

  3. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    SciTech Connect

    Syarifah, Ratna Dewi Suud, Zaki

    2015-09-30

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

  4. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Syarifah, Ratna Dewi; Suud, Zaki

    2015-09-01

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

  5. Compatibility of U-Pu-Zr alloy fuel with stainless steel cladding

    SciTech Connect

    Leibowitz, L.; Hins, A. G.; Strain, R. V.; Tsai, H.; Hofman, G. L.

    1989-07-01

    An important issue in the development of the Integral Fast Reactor is the chemical interaction between fuel and cladding. With a U--Pu--Zr ternary alloy fuel clad in stainless steel, the existence of low-melting eutectics in the U--Fe and Pu--Fe systems must be taken into account. There is at least the possibility that cladding integrity could be compromised by formation of liquid phases during irradiation. Data from a variety of sources have been brought to bear on this issue, and our current state of understanding is summarized in this report. The phase behavior of the relevant systems has been studied to provide a framework for understanding fuel/cladding interactions. Several experimental methods have been used. Differential thermal analysis experiments were performed, and the results compared with those of the phase studies. In addition, experiments were performed in which unirradiated fuel was heated in contact with cladding alloys. The results of these tests were strongly dependent on the state of the contacting surfaces. Other tests were performed in which sections of irradiated fuel with cladding were subjected to carefully controlled heating cycles. In general, there were no inconsistencies in these experimental and analytical studies although the data are still fragmentary. Differences that were observed may be reflections of kinetic factors. 9 refs., 12 figs., 6 tabs.

  6. Characterization of self-damaged (U,Pu)N fuel used in the NIMPHE program

    NASA Astrophysics Data System (ADS)

    Carvajal-Nunez, Ursula; Prieur, Damien; Janssen, Arne; Wiss, Thierry; Cambriani, Andrea; Vermorel, Emmanuel; Scheinost, Andreas; Somers, Joseph

    2013-11-01

    During in-pile, irradiation-induced damage occurs in nuclear fuel and results in a deterioration of its properties, which can affect the margin to melt of the fuel. Damage also occurs in fresh fuel through the self-irradiation process, and thus provides a convenient means to investigate changes in the material. A uranium-plutonium mixed nitride fuel made over 25 years ago, and stored in the ITU archives has been retrieved. Coupling EXAFS and TEM has shown that this material was still well-crystallized. However, an increase of 0.3% of the lattice parameter was found. As shown by the EXAFS, the U-N and Pu-N as well as the metal-metal distances are similarly affected. However, no significant modification of both anion and cation sublattice was found. Although no defect clustering was found by EXAFS, the presence of nanometric helium bubbles was demonstrated by TEM as well as nanometric disordered domains.

  7. Performance of U-Pu-Zr fuel cast into zirconium molds

    SciTech Connect

    Crawford, D.C.; Lahm, C.E. ); Tsai, H. )

    1992-10-01

    U-3Zr and U-20.5Pu-3Zr were injection cast into Zr tubes, or sheaths, rather than into quartz molds and clad in 316SS. These elements and standard-cast U-l0Zr and U-IgPu-l0Zr elements were irradiated in EBR-II to 2 at.% and removed for interim examination. Measurements of axial growth at indicate that the Zr-sheathed elements exhibited significantly less axial elongation than the standard-cast elements (1.3 to 1.8% versus 4.9 to 8.1%). Fuel material extruded through the ends of the Zr sheaths. allowing the low-Zr fuel to contact the cladding in some cases. Transverse metallographic sections reveal cracks in the Zr sheath through which fuel extruded and contacted cladding. The sheath is not a sufficient barrier between fuel and cladding to reduce FCCI. and any adverse effects due to increased FCCI will be evident as the elements attain higher burnup.

  8. Preparation, characterisation and out-of-pile property evaluation of (U,Pu)N fuel pellets

    NASA Astrophysics Data System (ADS)

    Ganguly, C.; Hegde, P. V.; Sengupta, A. K.

    1991-02-01

    (U 0.45Pu 0.55)N and (U 0.8Pu 0.2)N are being considered in India as advanced alternative fuels for the operating fast breeder test reactor (FBTR) and the forthcoming prototype fast breeder reactor (PFBR). Mixed nitride fuel pellets containing <0.1 wt% each of oxygen and carbon impurities were fabricated by the conventional "powder-pellet" (POP) and the advanced "sol-gel microsphere pelletisation" (SGMP) processes, involving two major steps. First, carbothermic reduction of an oxide-graphite powder mixture (in the form of tablets) or gel-microspheres at 1773-1823 K in N 2 followed by N2 + H2 and Ar+ H2 atmospheres. The nitride microspheres could be directly pelletised and sintered to pellets of relatively low density (≤ 85% TD) with an "open" pore structure desirable for LMFBR application. Thermal conductivity and hot hardness of nitride pellets were evaluated up to 1800 and 1500 K respectively. The out-of-pile chemical compatibility experiments of mixed nitride fuel pellets for FBTR with SS 316 cladding at 973 K for 1000 h did not reveal any significant fuel-cladding chemical interaction.

  9. Postirradiation examinations of U-Pu-Zr fuel elements from subassemblies X419 and X419A

    SciTech Connect

    Pahl, R G; Beck, W N; Hofman, G L; Lahm, C E; Villarreal, R

    1986-10-01

    Initial postirradiation examination of IFR type U-Pu-Zr fuel elements from X419 and X419A are reported. Characterization of the fuel at three levels of burnup, 0.8 at.%, 1.9 at.%, and 2.7 at.% is presented. Fuel swelling, microstructure, chemical redistribution, and fission gas behavior is discussed. No evidence was found for any performance-limiting damage to the fuel elements at these burnups.

  10. Validation studies based on critical experiments performed with fuel pin arrays moderated by Pu + U solutions

    SciTech Connect

    Smolen, G.R.; Matsumoto, T. )

    1989-01-01

    This paper outlines the results of a calculational study that was performed to validate the SCALE computer code system using data from critical experiments performed with fuel pin arrays moderated by mixed Pu + U aqueous solutions. A companion paper describes the experiments and discusses the criticality data that were obtained. These experimental activities are part of a joint exchange program between the US Department of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation of Japan in the area of criticality data development. The Consolidated fuel Reprocessing Program (CFRP) at Oak Ridge National Laboratory (ORNL) manages the program for the DOE. The experiments were conducted at the Battelle Pacific Northwest Laboratories-Critical Mass Laboratory (PNL-CML).

  11. Analysis of Pu Isotopes in Melted Fuel by Neutron Resonance Transmission: Examination by Linear Absorption Model

    NASA Astrophysics Data System (ADS)

    Kitatani, F.; Harada, H.; Takamine, J.; Kureta, M.; Seya, M.

    2014-04-01

    We have been studying the feasibility of neutron resonance transmission analysis (NRTA) for quantifying nuclear materials (Pu/U isotopes) in particle-like debris of melted fuel for nuclear material accountability and safeguards. The achievable measurement accuracy of NRTA was examined using a linear absorption model for the sample which contain substances other than nuclear fuel materials, such as boron and iron. The impurities (boron and iron etc.) in melted fuel are from the support structure and criticality control materials of the reactor core, and should be included to study the feasibility of NRTA for actual application. Neutron transmission spectra were calculated using the total neutron cross-sections in JENDL-4.0. The transmission spectra together with their uncertainties were evaluated. The study showed quantitatively that the statistical uncertainty in the determination of atomic number density of each isotope depends on the impurity density in the sample. The optimal thickness of the sample was determined for various impurity densities.

  12. /sup 238/Pu fuel form processes bimonthly report, May-June 1979

    SciTech Connect

    Folger, R. L.

    1980-02-01

    Progress in the Savannah River /sup 238/Pu Fuel Form Program is summarized. Full-scale fabrication tests of General-Purpose Heat Source (GPHS) fuel forms continued in the SRL Plutonium Experimental Facility (PEF) as four additional pellets (GPHS Pellets 5-8) were hot pressed. GPHS pellets fabricated by the reference process were dimensionally and structurally stable during and after final heat treatment. Microstructural studies confirmed that centerline GPHS process conditions produce pellets with a homogeneous microstructure and a uniform density. Because of the potential for excessive metal creep in existing furnace racks, the racks were considered unacceptable for GPHS fuel production in the PuFF. To eliminate metal creep, racks containing some ceramic components were designed to operate at 1600/sup 0/C in an oxygen atmosphere for more than 100 h. The four-key variables previously identified (shard sintering temperature, hot press load, hot press temperature, and load ramp) were found to correlate with production sphere fracture tendency and bulk density.

  13. EQ6 Calculation for Chemical Degradation of Shippingport LWBR (TH/U Oxide) Spent Nuclear Fuel Waste Packages

    SciTech Connect

    S. Arthur

    2000-09-14

    The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Shippingport Light Water Breeder Reactor (LWBR) (Ref. 1). The Shippingport LWBR SNF has been considered for disposal at the potential Yucca Mountain site. Because of the high content of fissile material in the SNF, the waste package (WP) design requires special consideration of the amount and placement of neutron absorbers and the possible loss of absorbers and SNF materials over geologic time. For some WPs, the outer shell corrosion-resistant material (CRM) and the corrosion-allowance inner shell may breach (Refs. 2 and 3), allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components and neutron absorbers from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing a Shippingport LWBR SNF seed assembly, and high-level waste (HLW) glass canisters arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which criticality control material, suggested for this WP design, will remain in the WP after corrosion/dissolution of the initial WP configuration (such that it can be effective in preventing criticality); (2) The extent to which fissile uranium and fertile thorium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this

  14. Assessing the oxygen stoichiometry during the sintering of (U, Pu)O2 fuel

    NASA Astrophysics Data System (ADS)

    Vaudez, Stéphane; Léchelle, Jacques; Berzati, Ségolène; Heintz, Jean-Marc

    2015-05-01

    Diffusion phenomena occurring in ceramics such as (U, Pu)O2 during sintering are affected by the oxygen content in the atmosphere. The latter sets the nature and the concentration of point defects which govern diffusion mechanisms in the bulk of the material. The oxygen partial pressure, pO2, of the sintering gas in equilibrium with mixed oxide (MOX) pellets needs to be precisely controlled; otherwise it may induce a large dispersion in the critical parameters for fuel manufacturing (Gauche, 2013; Matzke, 1987). It is crucial to understand the relation between the sintering atmosphere and the fuel throughout the thermal cycle. In this study, the oxygen potential of the sintering gas was monitored by measuring the oxygen partial pressure (pO2) at the outlet of a dilatometer by means of a zirconia probe. Coupling the thermal cycle with an outlet gas pO2 measurement makes it possible to identify different redox phenomena. Variations in the oxygen stoichiometry can be determined during the sintering of (U, Pu)O2, as well as can its final O/M. Our results make it possible to recommend a sintering atmosphere and sintering thermal cycle in order to obtain an O/M ratio that is as close as possible to the target value.

  15. Moisture content of PuO/sub 2/ fuel used for the milliwatt generator heat source

    SciTech Connect

    Zanotelli, W.A.

    1980-01-31

    The determination of the moisture content of /sup 238/Pu dioxide fuel for use in Milliwatt Generator heat sources was studied in an attempt to more clearly define the production fuel preloading procedures. The study indicated that water was not present or being adsorbed at various steps of the process (or during storage) that could lead to compatibility problems during pretreatment or long-term storage. The moisture content of the plutonium dioxide was analyzed by a commercial moisture analyzer. The moisture content at all steps of the process including storage averaged from 0.002% to 0.005%. The moisture content of the plutonium dioxide exposed to moist atmosphere for 7 days was 0.001%. These values indicated that no significant amount of moisture was adsorbed by the plutonium dioxide fuel charges. The only significant moisture content found was an average of 3.47%, after self-calcination. This was expected since no additional steps, other than self-heating of the fuel, are taken to remove the water.

  16. Thermochemical Analysis of Gas-Cooled Reactor Fuels Containing Am and Pu Oxides

    SciTech Connect

    Lindemer, T.B.

    2002-09-05

    Literature values and estimated data for the thermodynamics of the actinide oxides and fission products are applied to explain the chemical behavior in gas-cooled-reactor fuels. Emphasis is placed on the Am-O-C and Pu-O-C systems and the data are used to plot the oxygen chemical potential versus temperature of solid-solid and solid-gas equilibria. These results help explain observations of vaporization in Am oxides, nitrides, and carbides and provide guidance for the ceramic processing of the fuels. The thermodynamic analysis is then extended to the fission product systems and the Si-C-O system. Existing data on oxygen release (primarily as CO) as a function of burnup in the thoria-urania fuel system is reviewed and compared to values calculated from thermodynamic data. The calculations of oxygen release are then extended to the plutonia and americia fuels. Use of ZrC not only as a particle coating that may be more resistant to corrosion by Pd and other noble-metal fission products, but also as a means to getter oxygen released by fission is discussed.

  17. /sup 238/Pu fuel form processes. Quarterly report, April-June 1982

    SciTech Connect

    j

    1982-11-01

    Progress in studies of /sup 238/Pu fuel form processes is reported. Analytical studies of weld-quench cracking in DOP-26 iridium alloy-clad vent sets in General Purpose Heat Sources (GPHS) showed that weld-quench cracking is much more severe in MER alloy than in LR and NR alloys. Spark source mass spectrometry indicated that areas in DOP-26 alloy with severe weld-quench cracking have high thorium inhomogeneity. Secondary ion mass spectrometry revealed differences in LR and MR alloys that may be related to their dissimilar susceptibilities for weld-quench cracking. Impact ductility tests showed that welds in DOP-26 alloy clad vent sets made using parameters similar to PuFF production welding had high elongations. Decontamination of encapsulated GPHS pellets in PuFF was demonstrated using a solution of 3.5 M HNO/sub 3/ + 6.4 M HF which is capable of reducing transferable contamination below the specified 10/sup 3/ dpm upper limit in <30 minutes at a bath temperature of 80/sup 0/C using ultrasonic cleaning. Decontamination vessels were constructed to trap and condense acid vapors during decontamination. Impact and metallographic data showed that although the micro and macrostructures between LANL and SRL pellets have large differences, the difference in impact response between these two types of pellets is not correspondingly large. Both types of pellets have impacted successfully. The micro and macrostructures of SRP pellets made with either low fired shards sintered in Ar/5%O/sub 2/ or Ar are intermediate between those of the LANL and SRL pellets. Therefore, either type of SRP pellet should impact successfully.

  18. Improving the Assay of 239Pu in Spent and Melted Fuel Using the Nuclear Resonance Fluorescence Integral Resonance Transmission Method

    NASA Astrophysics Data System (ADS)

    Angell, C. T.; Hayakawa, T.; Shizuma, T.; Hajima, R.; Quiter, B. J.; Ludewigt, B. A.; Karwowski, H.; Rich, G.

    2015-10-01

    Non-destructive assay (NDA) of 239Pu in spent nuclear fuel is possible using the isotope-specific nuclear resonance fluorescence (NRF) integral resonance transmission (IRT) method. The IRT method measures the absorption of photons from a quasi-monoenergetic γ-ray beam due to all resonances in the energy width of the beam. According to calculations the IRT method could greatly improve assay times for 239Pu in nuclear fuel. To demonstrate and verify the IRT method, the IRT signature was first measured in 181Ta, whose nuclear resonant properties are similar to those of 239Pu, and then measured in 239Pu. These measurements were done using the quasi-monoenergetic beam at the High Intensity γ-ray Source (HIγS) in Durham, NC, USA. The IRT signature was observed as a decrease in scattering strength when the same isotope material was placed upstream of the scattering target. The results confirm the validity of the IRT method in both 181Ta and 239Pu.

  19. Full-length U-xPu-10Zr (x=0, 8, 19 wt%) Fast Reactor Fuel Test in FFTF

    SciTech Connect

    D. L. Porter; H.C. Tsai

    2012-08-01

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt%) metallic fast reactor test with commercial-length (91.4 cm active fuel column length) conducted to date. With few remaining test reactors there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning of life (BOL) peak cladding temperature of the hottest pin was 608?C, cooling to 522?C at end of life (EOL). Selected fuel pins were examined non destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3 cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ~0.7 X/L axial location along the fuel column. This resulted from a lower production of rare earth fission products higher in the fuel column as well as a much smaller delta-T between fuel center and cladding, and therefore less FCCI, despite the higher cladding temperature. This behavior could

  20. /sup 238/Pu fuel form processes. Final report, January-September 1983

    SciTech Connect

    Mosley, W.C.; Taylor, D.H.

    1983-01-01

    Progress is reported on the following: analytical studies of weld-quench cracking in DOP-26 iridium alloy, iridium//sup 238/PuO/sub 2/ compatibility test, surface area measurements of /sup 238/PuO/sub 2/ using the Blaine air permeability apparatus, and helium release from /sup 238/PuO/sub 2/. (MHR)

  1. Phase Characteristics of a Number of U-Pu-Am-Np-Zr Metallic Alloys for Use as Fast Reactor Fuels

    SciTech Connect

    Douglas E. Burkes; J. Rory Kennedy; Thomas Hartmann; Cynthia A. Papesch; Denis D. Keiser, Jr.

    2010-01-01

    Metallic fuel alloys consisting of uranium, plutonium, and zirconium with minor additions of americium and neptunium are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. A series of test designs for the Advanced Fuel Cycle Initiative (AFCI) have been irradiated in the Advanced Test Reactor (ATR), designated as the AFC-1 and AFC-2 designs. Metal fuel compositions in these designs have included varying amounts of U, Pu, Zr, and minor actinides (Am, Np). Investigations into the phase behavior and relationships based on the alloy constituents have been conducted using x-ray diffraction and differential thermal analysis. Results of these investigations, along with proposed relationships between observed behavior and alloy composition, are provided. In general, observed behaviors can be predicted by a ternary U-Pu-Zr phase diagram, with transition temperatures being most dependent on U content. Furthermore, the enthalpy associated with transitions is strongly dependent on the as-cast microstructural characteristics.

  2. Phase characteristics of a number of U-Pu-Am-Np-Zr metallic alloys for use as fast reactor fuels

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Kennedy, J. Rory; Hartmann, Thomas; Papesch, Cynthia A.; Keiser, Dennis D., Jr.

    2010-01-01

    Metallic fuel alloys consisting of uranium, plutonium, and zirconium with minor additions of americium and neptunium are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. A series of test designs for the Advanced Fuel Cycle Initiative (AFCI) have been irradiated in the Advanced Test Reactor (ATR), designated as the AFC-1 and AFC-2 designs. Metal fuel compositions in these designs have included varying amounts of U, Pu, Zr, and minor actinides (Am, Np). Investigations into the phase behavior and relationships based on the alloy constituents have been conducted using X-ray diffraction and differential thermal analysis. Results of these investigations, along with proposed relationships between observed behavior and alloy composition, are provided. In general, observed behaviors can be predicted by a ternary U-Pu-Zr phase diagram, with transition temperatures being most dependent on U content. Furthermore, the enthalpy associated with transitions is strongly dependent on the as-cast microstructural characteristics.

  3. Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

    NASA Astrophysics Data System (ADS)

    Porter, D. L.; Tsai, Hanchung

    2012-08-01

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 °C, cooling to 522 °C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ˜0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher ΔT between fuel center and cladding than at core center, together providing more rare earths at the cladding and more FCCI. This behavior could

  4. Space and Time Distribution of Pu Isotopes inside The First Experimental Fuel Pin Designed for PWR and Manufactured in Indonesia

    NASA Astrophysics Data System (ADS)

    Suwardi; Setiawan, J.; Susilo, J.

    2017-01-01

    The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared and planned to be tested in power ramp irradiation. An irradiation test should be designed to allow an experiment can be performed safely and giving maximum results of many performance aspects of design and manufacturing. Performance analysis to the fuel specimen shows that the specimen is not match to be used for power ramp testing. Enlargement by 0.20 mm of pellet diameter has been proposed. The present work is evaluation of modified design for important aspect of isotopic Pu distribution during irradiation test, because generated Pu isotopes in natural UO2 fuel, contribute more power relative to the contribution by enriched UO2 fuel. The axial profile of neutrons flux have been chosen from both experimental measurement and model calculation. The parameters of ramp power has been obtained from statistical experiment data. A simplified and typical base-load commercial PHWR profile of LHR history has been chosen, to determine the minimum irradiation time before ramp test can be performed. The data design and Mat pro XI materials properties models have been chosen. The axial profile of neutrons flux has been accommodated by 5 slices of discrete pin. The Pu distribution of slice-4 with highest power rate has been chosen to be evaluated. The radial discretion of pellet and cladding and numerical parameter have been used the default best practice of TU. The results shows that Pu 239 increased rapidly. The maximum burn up of slice 4 at upper the median slice, it reached nearly 90% of maximum value at about 6000 h with peak of 0.8%a Pu/HM at 22000 h, which is higher than initial U 235. Each 240, 241 and 240 Pu grows slower and ends up to 0.4, 0.2 and 0.18 % respectively. This results can be used for verification of other aspect of fuel behavior in the modeling results and also can be used as guide and comparison to the future post irradiation examination for Pu isotopes distribution.

  5. Modeling of Selected Ceramic Processing Parameters Employed in the Fabrication of 238PuO 2 Fuel Pellets

    NASA Astrophysics Data System (ADS)

    Brockman, R. A.; Kramer, D. P.; Barklay, C. D.; Cairns-Gallimore, D.; Brown, J. L.; Huling, J. C.; Van Pelt, C. E.

    Recent deep space missions utilize the thermal output of the radioisotope plutonium-238 as the fuel in the thermal to electrical power system. Since the application of plutonium in its elemental state has several disadvantages, the fuel employed in these deep space power systems is typically in the oxide form such as plutonium-238 dioxide (238PuO2). As an oxide, the processing of the plutonium dioxide into fuel pellets is performed via "classical" ceramic processing unit operations such as sieving of the powder, pressing, sintering, etc. Modeling of these unit operations can be beneficial in the understanding and control of processing parameters with the goal of further enhancing the desired characteristics of the 238PuO2 fuel pellets. A finite element model has been used to help identify the time-temperature-stress profile within a pellet during a furnace operation taking into account that 238PuO2 itself has a significant thermal output. Results of the modeling efforts will be discussed.

  6. A new fabrication route for SFR fuel using (U, Pu)O2 powder obtained by oxalic co-conversion

    NASA Astrophysics Data System (ADS)

    Vaudez, Stéphane; Belin, Renaud C.; Aufore, Laurence; Sornay, Philippe; Grandjean, Stéphane

    2013-11-01

    The standard powder metallurgy preparation of SFR (Sodium Fast Reactor) oxide fuel involves UO2 and PuO2 co-milling. An alternative route, using a solid-solution of mixed oxide obtained by oxalic co-conversion as the starting material, is presented. It was used to manufacture nuclear fuels for the "COPIX" irradiation conducted in the Phenix SFR. Two processes using co-converted powders were tested to elaborate fuel pellets: (1) the Direct Process that consists in pressing and sintering the mixed oxide with the final Pu content and (2) the Dilution Process, which involves the dilution of a high Pu content mixed oxide with UO2. After studying the structural and microstructural evolution with temperature of these innovative raw materials, the elaboration parameters were adjusted to obtain final pellets in accordance with the Phenix fuel specifications. This study demonstrates the feasibility of such new fabrication route at laboratory scale and, from a more fundamental prospect, allows a better understanding of the underlying phenomena involved during sintering.

  7. Modeling of selected ceramic processing parameters employed in the fabrication of 238PuO2 fuel pellets

    DOE PAGES

    Brockman, R. A.; Kramer, D. P.; Barklay, C. D.; ...

    2011-10-01

    Recent deep space missions utilize the thermal output of the radioisotope plutonium-238 as the fuel in the thermal to electrical power system. Since the application of plutonium in its elemental state has several disadvantages, the fuel employed in these deep space power systems is typically in the oxide form such as plutonium-238 dioxide (238PuO2). As an oxide, the processing of the plutonium dioxide into fuel pellets is performed via ''classical'' ceramic processing unit operations such as sieving of the powder, pressing, sintering, etc. Modeling of these unit operations can be beneficial in the understanding and control of processing parameters withmore » the goal of further enhancing the desired characteristics of the 238PuO2 fuel pellets. A finite element model has been used to help identify the time-temperature-stress profile within a pellet during a furnace operation taking into account that 238PuO2 itself has a significant thermal output. The results of the modeling efforts will be discussed.« less

  8. Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

    SciTech Connect

    Pahl, R.G.; Porter, D.L.; Lahm, C.E. ); Hofman, G.L. )

    1990-07-01

    Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to {gt}15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel.

  9. Evaluation of high plutonia (44% PuO 2) MOX as a fuel for fast breeder test reactor

    NASA Astrophysics Data System (ADS)

    Sengupta, A. K.; Khan, K. B.; Panakkal, Jose; Kamath, H. S.; Banerjee, S.

    2009-03-01

    Uranium plutonium mixed oxide (MOX) containing up to 30% plutonia is the conventional fuel for liquid metal cooled fast breeder reactor (LMFBR). Use of high plutonia (>30%) MOX fuel in LMFBR had been of interest but not pursued. Of late, it has regained importance for faster disposition of plutonium and also for making compact fast reactors. Some of the issues of high plutonia MOX fuels which are of concern are its chemical compatibility with liquid sodium coolant, dimensional stability and low thermal conductivity. Available literature information for MOX fuel is limited to a plutonium content of 30%. Thermodynamic assessment of mixed oxide fuels indicate that with increasing plutonia oxygen potential of the fuel increases and the fuel become more prone to chemical attack by liquid sodium coolant in case of a clad breach. In the present investigation, some of these issues of MOX fuel have been studied to evaluate this fuel for its use in fast reactor. Extensive work on the out-of-pile thermo-physical properties and fuel-coolant chemical compatibility under different simulated reactor conditions has been carried out. Results of these studies were compared with the available literature information on low plutonia MOX fuel and critically analyzed to predict in reactor behaviour of this fuel containing 44% PuO 2. The results of these out-of-pile studies have been very encouraging and helped in arriving at a suitable and achievable fuel specification for utilization of this fuel in fast breeder test reactor (FBTR). As a first step of test pin irradiation programme in FBTR, eight subassemblies of the MOX fuel are undergoing irradiation in FBTR.

  10. Fuel/cladding compatibility in high-burnup U-19 Pu-10 Zr/HT-9 clad fuel at elevated temperatures

    SciTech Connect

    Cohen, A.B.; Tsai, H.; Sanecki, J.E.; Neimark, L.A. )

    1992-01-01

    The U-Pu-Zr metallic fuel in the integral fast reactor may interact chemically with the steel cladding at elevated temperatures, leading to a thinning of the cladding and eventual pin failure. Also, as a result of the fuel/cladding chemical interaction (FCCI), iron may diffuse into the fuel and form a lower melting phase with uranium and plutonium. If the temperature is raised above the solidus temperature of this phase, the fuel can undergo liquefaction, i.e., the formation of a mixture of liquid and solid phases, that may promote further cladding interaction. Fuel/cladding chemical interaction, therefore, is a complex phenomenon on both sides of the fuel/cladding interface that depends on fuel and cladding compositions, linear power rating, burnup, and cladding temperature. The purpose of this study was to determine the temperature at which the fuel/cladding interaction region forms solid-plus-liquid phases above the normal in-reactor operating temperatures of high-burnup (11 at.%) Mark V-type fuel for the Experimental Breeder Reactor II (EBR-II). The Mark V fuel is being developed as a future driver fuel for the reactor. The effect of this solid-plus-liquid mixture on the kinetics and mechanism of FCCI was also investigated. This paper updates results previously reported for lower-burnup Mark V-type fuel elements.

  11. Results of irradiation of (U0.55Pu0.45)N and (U0.4Pu0.6)N fuels in BOR-60 up to ˜12 at.% burn-up

    NASA Astrophysics Data System (ADS)

    Rogozkin, B. D.; Stepennova, N. M.; Fedorov, Yu. Ye.; Shishkov, M. G.; Kryukov, F. N.; Kuzmin, S. V.; Nikitin, O. N.; Belyaeva, A. V.; Zabudko, L. M.

    2013-09-01

    In the article presented are the results of post-irradiation tests of helium bonded fuel pins with mixed mononitride fuel (U0.55Pu0.45)N and (U0.4Pu0.6)N having 85% density irradiated in BOR-60 reactor. Achieved maximum burn-up was, respectively, equal to 9.4 and 12.1 at.% with max linear heat rates 41.9 and 54.5 kW/m. Maximum irradiation dose was 43 dpa. No damage of claddings made of ChS-68 steel (20% cold worked) was observed, and ductility margin existed. Maximum depth of cladding corrosion was within 15 μm. Swelling rates of (U0.4Pu0.6)N and (U0.55Pu0.45)N were, respectively, ˜1.1% and ˜0.68% per 1 at.%. Gas release rate did not exceed 19.3% and 19%. Pattern of porosity distribution in the fuel influenced fuel swelling and gas release rates. Plutonium and uranium are uniformly distributed in the fuel, local minimum values of their content being caused by pores and cracks in the pellets. The observable peaks in content distribution are probably connected with the local formation of isolated phases (e.g. Mo, Pd) while the minimum values refer to fuel pores and cracks. Xenon and cesium tend to migrate from the hot sections of fuel, and therefore their min content is observed in the central section of the fuel pellets. Phase composition of the fuel was determined with X-ray diffractometer. The X-ray patterns of metallographic specimens were obtained by the scanning method (the step was 0.02°, the step exposition was equal to 2 s). From the X-ray diffraction analysis data, it follows that the nitrides of both fuel types have the single-phase structure with an FCC lattice (see Table 6).

  12. Physical characteristics of LWRs and SCLWRs loaded by ({sup 233}U-Th-{sup 238}U) oxide fuel with small additions of {sup 231}Pa

    SciTech Connect

    Kulikov, E.G.; Shmelev, A.N.; Apse, V.A.; Kulikov, G.G.

    2007-07-01

    The paper investigates the possibility and attractiveness of using (U-Th) fuel in light-water reactors (LWRs) and in light-water reactors with super-critical coolant parameters (SCLWRs). It is proposed to dilute {sup 233}U with {sup 238}U to enhance the proliferation resistance of this fissionable isotope. If is noteworthy that she idea was put forward for the first time by she well known American physicist and participant of the Manhattan Project Dr. T. Taylor. Various fuel compositions are analyzed and compared on fuel breeding, achievable values of fuel burn-up and cross-sections of parasitic neutron absorption. It is also demonstrated that small {sup 231}Pa additions (several percent) into the fuel allows: to increase fuel burn-up, to achieve more negative temperature reactivity coefficient of coolant and to enhance nonproliferation of the fuel. (authors)

  13. U and Pu Gamma-Ray Measurements of Spent Fuel Using a Gamma-Ray Mirror Band-Pass Filter

    SciTech Connect

    Ziock, Klaus-Peter; Alameda, J.B.; Brejnholt, N.F.; Decker, T.A.; Descalle, M.A.; Fernandez-Perea, M.; Hill, R.M.; Kisner, R.A.; Melin, A.M.; Patton, B.W.; Ruz, J.; Soufli, R.; Pivovaroff, M.J.

    2014-01-01

    Abstract. We report on the use of grazing incidence gamma-ray mirrors to serve as a narrow band-pass filter for advanced non-destructive analysis (NDA) of spent nuclear fuel. The purpose of the mirrors is to limit the radiation reaching a HPGe detector to narrow spectral bands around characteristic emission lines from fissile isotopes in the fuel. This overcomes the normal rate issues when performing gamma-ray NDA measurements. In a proof-of-concept experiment, a set of simple flat gamma-ray mirrors were used to directly observe the atomic florescence lines from U and Pu from spent fuel pins with the detector located in a shirt-sleeve environment. The mirrors, consisting of highly polished silicon substrates deposited with WC/SiC multilayer coatings, successfully deflected the lines of interest while the intense primary radiation beam from the fuel was blocked by a lead beam stop. The gamma-ray multilayer coatings that make the mirrors work at the gamma-ray energies used here (~ 100 keV) have been experimentally tested at energies as high as 645 keV, indicating that direct observation of nuclear emission lines from 239Pu should be possible with an appropriately designed optic and shielding configuration.

  14. Evaluation of Aqueous and Powder Processing Techniques for Production of Pu-238-Fueled General Purpose Heat Sources

    SciTech Connect

    Not Available

    2008-06-01

    This report evaluates alternative processes that could be used to produce Pu-238 fueled General Purpose Heat Sources (GPHS) for radioisotope thermoelectric generators (RTG). Fabricating GPHSs with the current process has remained essentially unchanged since its development in the 1970s. Meanwhile, 30 years of technological advancements have been made in the fields of chemistry, manufacturing, ceramics, and control systems. At the Department of Energy’s request, alternate manufacturing methods were compared to current methods to determine if alternative fabrication processes could reduce the hazards, especially the production of respirable fines, while producing an equivalent GPHS product. An expert committee performed the evaluation with input from four national laboratories experienced in Pu-238 handling.

  15. Development of spent fuel reprocessing process based on selective sulfurization: Study on the Pu, Np and Am sulfurization

    NASA Astrophysics Data System (ADS)

    Kirishima, Akira; Amano, Yuuki; Nihei, Toshifumi; Mitsugashira, Toshiaki; Sato, Nobuaki

    2010-03-01

    For the recovery of fissile materials from spent nuclear fuel, we have proposed a novel reprocessing process based on selective sulfurization of fission products (FPs). The key concept of this process is utilization of unique chemical property of carbon disulfide (CS2), i.e., it works as a reductant for U3O8 but works as a sulfurizing agent for minor actinides and lanthanides. Sulfurized FPs and minor actinides (MA) are highly soluble to dilute nitric acid while UO2 and PuO2 are hardly soluble, therefore, FPs and MA can be removed from Uranium and Plutonium matrix by selective dissolution. As a feasibility study of this new concept, the sulfurization behaviours of U, Pu, Np, Am and Eu are investigated in this paper by the thermodynamical calculation, phase analysis of chemical analogue elements and tracer experiments.

  16. Feedback components of a U20Pu10Zr-fueled compared to a U10Zr-fueled EBR-II

    SciTech Connect

    Meneghetti, D.; Kucera, D.A.

    1988-12-31

    Calculated feedback components of the regional contributions of the power reactivity decrements (PRDs) and of the temperature coefficients of reactivity of a U20Pu10Zr-fueled and of a U10Zr-fueled Experimental Breeder Reactor II (EBR-II) are compared. The PRD components are also separated into power-to-flow dependent and solely power dependent parts. The effects of these values upon quantities useful for indicating the comparative potential inherent safety characteristics of these EBR-II loadings are presented.

  17. Feedback components of a U20Pu10Zr-fueled compared to a U10Zr-fueled EBR-II

    SciTech Connect

    Meneghetti, D.; Kucera, D.A.

    1988-01-01

    Calculated feedback components of the regional contributions of the power reactivity decrements (PRDs) and of the temperature coefficients of reactivity of a U20Pu10Zr-fueled and of a U10Zr-fueled Experimental Breeder Reactor II (EBR-II) are compared. The PRD components are also separated into power-to-flow dependent and solely power dependent parts. The effects of these values upon quantities useful for indicating the comparative potential inherent safety characteristics of these EBR-II loadings are presented.

  18. Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II (Experimental Breeder Reactor)

    SciTech Connect

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1988-01-01

    The Integral Fast Reactor (IFR) is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the Experimental Breeder Reactor II (EBR-II) on the Department of Energy site near Idaho Falls, Idaho. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing. The purpose of this paper will be to summarize our latest results of irradiation testing uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. 10 refs., 13 figs., 2 tabs.

  19. High-silicon {sup 238}PuO{sub 2} fuel characterization study: Half module impact tests

    SciTech Connect

    Reimus, M.A.H.

    1997-01-01

    The General-Purpose Heat Source (GPHS) provides power for space missions by transmitting the heat of [sup 238]Pu decay to an array of thermoelectric elements. The modular GPHS design was developed to address both survivability during launch abort and return from orbit. Previous testing conducted in support of the Galileo and Ulysses missions documented the response of GPHSs to a variety of fragment- impact, aging, atmospheric reentry, and Earth-impact conditions. The evaluations documented in this report are part of an ongoing program to determine the effect of fuel impurities on the response of the heat source to conditions baselined during the Galileo/Ulysses test program. In the first two tests in this series, encapsulated GPHS fuel pellets containing high levels of silicon were aged, loaded into GPHS module halves, and impacted against steel plates. The results show no significant differences between the response of these capsules and the behavior of relatively low-silicon fuel pellets tested previously.

  20. Conceptual design for a receiving station for the nondestructive assay of PuO/sub 2/ at the fuels and materials examination facility

    SciTech Connect

    Sampson, T.E.; Speir, L.G.; Ensslin, N.; Hsue, S.T.; Johnson, S.S.; Bourret, S.; Parker, J.L.

    1981-11-01

    We propose a conceptual design for a receiving station for input accountability measurements on PuO/sub 2/ received at the Fuels and Materials Examination Facility at the Hanford Engineering Development Laboratory. Nondestructive assay techniques are proposed, including neutron coincidence counting, calorimetry, and isotopic determination by gamma-ray spectroscopy, in a versatile data acquisition system to perform input accountability measurements with precisions better than 1% at throughputs of up to 2 M.T./yr of PuO/sub 2/.

  1. 238PuO2 Fuel and Dosimetry

    SciTech Connect

    Mayo, Douglas R.; Rawool-Sullivan, Mohini; Garner, Scott Edward; Wenz, Tracy R.; Karpius, Peter Joseph

    2016-06-01

    238Pu is an ideal material for use as a heat source with its half-life of 87.7 years and copious particle emissions. 238Pu radioisotope thermoelectric generators (RTGs) have found use for pacemakers, Apollo Space missions, Mars rovers, and Voyager spacecraft. In evaluating the dose to personnel and components near a 238Pu-based RTG, a number of additional nuclides and their daughter products must be considered to get an accurate estimate for γ-dose, and the amount of 17O and 18O for the neutron-dose must be considered. This paper looks at the contributing nuclides and their daughter products that add the most to the dose rates.

  2. Direct Measurement of U235 and Pu239 in Spent Fuel Rods with Gamma-Ray Mirrors

    SciTech Connect

    Ziock, Klaus-Peter; Alameda, J. B.; Brejnholt, N. F.; Decker, T. A.; Descalle, M. A.; Fernandez-Perea, M.; Hill, R. M.; Kisner, R. A.; Melin, A. M.; Patton, B. W.; Ruz, J.; Soufli, R.

    2013-09-30

    The amounts of fissile Pu and U in spent nuclear fuel are of primary concern to the safeguards community. In particular, there are issues when safeguards transitions from an item accountancy basis (such as fuel bundles) to a fissile material mass basis as occurs when spent fuel enters a reprocessing plant. Discrepancies occur because item accountancy requires estimating the content of fissile material using indirect techniques such as the fuel burn-up and item-level measurements of radiation emissions from fission by-products. Direct measurement of the fissile content by monitoring line emissions from fissile species themselves is impossible because the lines are much weaker than those emitted by shorter-lived isotopes in the fuel. The goal of this project is to develop a technique to directly measure these weaker lines despite the presence of overwhelming radiation from other isotopes. This is achieved by using gamma-ray mirrors as a narrow band-pass filter. The mirrors reflect only energies of interest toward a HPGe detector that is shielded from direct view of the spent fuel and its fierce emissions. This can significantly improve the reliability with which the mass of fissile material is tracked.

  3. Oxidative dissolution of unirradiated Mimas MOX fuel (U/Pu oxides) in carbonated water under oxic and anoxic conditions

    NASA Astrophysics Data System (ADS)

    Odorowski, Mélina; Jégou, Christophe; De Windt, Laurent; Broudic, Véronique; Peuget, Sylvain; Magnin, Magali; Tribet, Magaly; Martin, Christelle

    2016-01-01

    Few studies exist concerning the alteration of Mimas Mixed-OXide (MOX) fuel, a mixed plutonium and uranium oxide, and data is needed to better understand its behavior under leaching, especially for radioactive waste disposal. In this study, two leaching experiments were conducted on unirradiated MOX fuel with a strong alpha activity (1.3 × 109 Bq.gMOX-1 reproducing the alpha activity of spent MOX fuel with a burnup of 47 GWd·tHM-1 after 60 years of decay), one under air (oxic conditions) for 5 months and the other under argon (anoxic conditions with [O2] < 1 ppm) for one year in carbonated water (10-2 mol L-1). For each experiment, solution samples were taken over time and Eh and pH were monitored. The uranium in solution was assayed using a kinetic phosphorescence analyzer (KPA), plutonium and americium were analyzed by a radiochemical route, and H2O2 generated by the water radiolysis was quantified by chemiluminescence. Surface characterizations were performed before and after leaching using Scanning Electron Microscopy (SEM), Electron Probe Microanalyzer (EPMA) and Raman spectroscopy. Solubility diagrams were calculated to support data discussion. The uranium releases from MOX pellets under both oxic and anoxic conditions were similar, demonstrating the predominant effect of alpha radiolysis on the oxidative dissolution of the pellets. The uranium released was found to be mostly in solution as carbonate species according to modeling, whereas the Am and Pu released were significantly sorbed or precipitated onto the TiO2 reactor. An intermediate fraction of Am (12%) was also present as colloids. SEM and EPMA results indicated a preferential dissolution of the UO2 matrix compared to the Pu-enriched agglomerates, and Raman spectroscopy showed the Pu-enriched agglomerates were slightly oxidized during leaching. Unlike Pu-enriched zones, the UO2 grains were much more sensitive to oxidative dissolution, but the presence of carbonates did not enable observation of an

  4. Microwave based oxidation process for recycling the off-specification (U,Pu)O2 fuel pellets

    NASA Astrophysics Data System (ADS)

    Singh, G.; Khot, P. M.; Kumar, Pradeep; Bhatt, R. B.; Behere, P. G.; Afzal, Mohd

    2017-02-01

    This paper reports development of a process named MicroWave Direct Oxidation (MWDO) for recycling the off-specification (U,Pu)O2 mixed oxide (MOX) fuel pellets generated during fabrication of typical fast reactor fuels. MWDO is a two-stage, single-cycle process based on oxidative pulverisation of pellets using 2450 MHz microwave. The powder sinterability was evaluated by bulk density and BET specific surface area. The oxidised powders were analyzed for phases using XRD and stoichiometry by thermogravimetry. The sinterability was significantly enhanced by carrying out oxidation in higher oxygen partial pressure and by subjecting MOX to multiple micronisation-oxidation cycles. After three cycles, the recycled powder from (U,28%Pu)O2 resulted surface area >3 m2/g and 100% re-used for MOX fabrication. The flow sheet was developed for maximum utilization of recycled powder describable by a parameter called Scrap Recycling Ratio (SRR). The process demonstrates smaller processing cycle, better powder properties and higher oxidative pulverisation over conventional method.

  5. Review of the phase equilibria and thermodynamics of binary and ternary systems composed of IFR fuel component metals (U, Pu, and Zr) and nitrogen

    SciTech Connect

    Leibowitz, L.; Veleckis, E.; Blomquist, R. A.

    1987-09-01

    Studies of the interactions of IFR fuel alloys with stainless steel cladding materials have shown evidence for segregation of metallic alloy components at the fuel-cladding interface. The phenomenon seems to be related to the availability of nitrogen and other reactive elements. For the interpretation of surface segregation results it will be necessary to establish (1) the composition of the phases formed in the ternary alloys under the action of nitrogen, (2) diffusion characteristics of various components in these phases and those of nitrogen itself, and (3) activity gradients established in the component metals providing driving forces for material migration. In this technical memorandum we present a literature survey and evaluation of the binary and ternary systems formed among uranium, plutonium, zirconium, and nitrogen. Both the phase diagram data and available thermodynamics are included. The following systems were surveyed: U-N, Pu-N, Zr-N, U-Pu, U-Zr, Pu-Zr, U-Pu-Zr, U-Pu-N, U-Zr-N, and Pu-Zr-N.

  6. Fully Coupled Modeling of Burnup-Dependent (U1- y , Pu y )O2- x Mixed Oxide Fast Reactor Fuel Performance

    NASA Astrophysics Data System (ADS)

    Liu, Rong; Zhou, Wenzhong; Zhou, Wei

    2016-03-01

    During the fast reactor nuclear fuel fission reaction, fission gases accumulate and form pores with the increase of fuel burnup, which decreases the fuel thermal conductivity, leading to overheating of the fuel element. The diffusion of plutonium and oxygen with high temperature gradient is also one of the important fuel performance concerns as it will affect the fuel material properties, power distribution, and overall performance of the fuel pin. In order to investigate these important issues, the (U1- y Pu y )O2- x fuel pellet is studied by fully coupling thermal transport, deformation, oxygen diffusion, fission gas release and swelling, and plutonium redistribution to evaluate the effects on each other with burnup-dependent models, accounting for the evolution of fuel porosity. The approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with fast reactor mixed oxide fuel performance analysis. The modeling results showed a consistent fuel performance comparable with the previous results. Burnup degrades the fuel thermal conductivity, resulting in a significant fuel temperature increase. The fission gas release increased rapidly first and then steadily with the burnup increase. The fuel porosity increased dramatically at the beginning of the burnup and then kept constant as the fission gas released to the fuel free volume, causing the fuel temperature to increase. Another important finding is that the deviation from stoichiometry of oxygen affects greatly not only the fuel properties, for example, thermal conductivity, but also the fuel performance, for example, temperature distribution, porosity evolution, grain size growth, fission gas release, deformation, and plutonium redistribution. Special attention needs to be paid to the deviation from stoichiometry of oxygen in fuel fabrication. Plutonium content will also affect the fuel material properties and performance

  7. Use of fission track analysis technique for the determination of MicroBequerel level of 239Pu in urine samples from radiation workers handling MOX fuel.

    PubMed

    Yadav, J R; Rao, D D; Kumar, Ranjeet; Aggarwal, S K

    2011-07-01

    Fission track analysis (FTA) technique for the determination of (239)Pu excreted through urine has been standardized using blank samples, tracer and (239)Pu spikes. Double stage anion exchange separation protocol has been applied and an average radiochemical recovery of (239)Pu of 18% was obtained. An average track registration efficiency of 11 tracks per μBq of (239)Pu, irradiated to 0.35×10(17) neutron fluence was established. Reagent blank urine samples from 11 controlled subjects were analyzed by FTA and an average of 149±14 tracks was obtained. Minimum detectable activity of 34μBqL(-1) of urine sample was obtained and will be useful for monitoring chronic exposure cases handling MOX fuel.

  8. The Geochemical Behaviour of Tc, Np, and Pu in Spent Nuclear Fuel in an Oxidizing Environment

    SciTech Connect

    Buck, Edgar C.; Hanson, Brady D.; McNamara, Bruce K.; R. Giere and P. Stille

    2004-10-01

    Studies at the Nopal and Shinkolowbwe uranium deposits show that the primary uraninite (UO2) altered to a suite of secondary uranyl minerals similar to those observed in corrosion tests with uranium oxide . Although the Nopal I deposit tells us something about the possible fate of uranium, it tells us little about the likely fate of the important long-lived radionuclides; iodine (129I), cesium (135Cs), technetium (99Tc), neptunium (237Np), and plutonium (239Pu). Most performance assessment (PA) models, assume conservatively, that as the UO2 matrix corrodes, the key radionuclides (129I, 99Tc, 237Np, and 239Pu) will be released congruently. In so doing, these PA models force increased reliance on human engineered barriers.

  9. Modeling Constituent Redistribution in U-Pu-Zr Metallic Fuel Using the Advanced Fuel Performance Code BISON

    SciTech Connect

    Douglas Porter; Steve Hayes; Various

    2014-06-01

    The Advanced Fuels Campaign (AFC) metallic fuels currently being tested have higher zirconium and plutonium concentrations than those tested in the past in EBR reactors. Current metal fuel performance codes have limitations and deficiencies in predicting AFC fuel performance, particularly in the modeling of constituent distribution. No fully validated code exists due to sparse data and unknown modeling parameters. Our primary objective is to develop an initial analysis tool by incorporating state-of-the-art knowledge, constitutive models and properties of AFC metal fuels into the MOOSE/BISON (1) framework in order to analyze AFC metallic fuel tests.

  10. Imminent: Irradiation Testing of (Th,Pu)O{sub 2} Fuel - 13560

    SciTech Connect

    Kelly, Julian F.; Franceschini, Fausto

    2013-07-01

    Commercial-prototype thorium-plutonium oxide (Th-MOX) fuel pellets have been loaded into the material test reactor in Halden, Norway. The fuel is being operated at full power - with instrumentation - in simulated LWR / PHWR conditions and its behaviour is measured 'on-line' as it operates to high burn-up. This is a vital test on the commercialization pathway for this robust new thoria-based fuel. The performance data that is collected will support a fuel modeling effort to support its safety qualification. Several different samples of Th-MOX fuel will be tested, thereby collecting information on ceramic behaviours and their microstructure dependency. The fuel-cycle reasoning underpinning the test campaign is that commercial Th- MOX fuels are an achievable intermediate / near-term SNF management strategy that integrates well with a fast reactor future. (authors)

  11. Steady-state irradiation testing of U-Pu-Zr fuel to >18% burnup

    SciTech Connect

    Pahl, R.G.; Wisner, R.S. ); Billone, M.C.; Hofman, G.L. )

    1990-01-01

    Tests of austenitic stainless steel clad U-xP-10Zr fuel (x=o, 8, 19 wt. %) to peak burnups as high as 18.4 at. % have been completed in the EBR-II. Fuel swelling and fractional fission gas release are slowly increasing functions of burnup beyond 2 at. % burnup. Increasing plutonium content in the fuel reduces swelling and decreases the amount of fission gas which diffuses from fuel to plenum. LIFE-METAL code modelling of cladding strains is consistent with creep by fission gas loading and irradiation-induced swelling mechanisms. Fuel/cladding chemical interaction involves the ingress of rare-earth fission products. Constituent redistribution in the fuel had not limited steady-state performance. Cladding breach behavior at closure welds, in the gas plenum, and in the fuel column region have been benign events. 3 refs., 5 figs.

  12. Chemical states of fission products in irradiated (U 0.3Pu 0.7)C 1+ x fuel at high burn-ups

    NASA Astrophysics Data System (ADS)

    Agarwal, Renu; Venugopal, V.

    2006-12-01

    The chemical states of fission products have been theoretically determined for the irradiated carbide fuel of Fast Breeder Test Reactor (FBTR) at Kalpakkam, India, at different burn-ups. The SOLGASMIX-PV computer code was used to determine the equilibrium chemical composition of the fuel. The system was assumed to be composed of a gaseous phase at one atmosphere pressure, and various solid phases. The distribution of elements in these phases and their chemical states at different temperatures were calculated as a function of burn-up. The FBTR fuel, (U 0.3Pu 0.7)C 1+ x, was loaded with C/M values in the range, 1.03-1.06. The present calculations indicated that even for the lowest starting C/M of 1.03 in the FBTR fuel, the liquid metal phase of (U, Pu), should not appear at a burn-up as high as 150 GWd/t.

  13. High-silicon {sup 238}PuO{sub 2} fuel characterization study: Half module impact tests

    SciTech Connect

    Reimus, M.A.

    1997-01-01

    The General-Purpose Heat Source (GPHS) provides power for space missions by transmitting the heat of {sup 238}Pu decay to an array of thermoelectric elements. The modular GPHS design was developed to address both survivability during launch abort and return from orbit. Previous testing conducted in support of the Galileo and Ulysses missions documented the response of GPHSs to a variety of fragment-impact, aging, atmospheric reentry, and Earth-impact conditions. The evaluations documented in this report are part of an ongoing program to determine the effect of fuel impurities on the response of the heat source to conditions baselined during the Galileo/Ulysses test program. In the first two tests in this series, encapsulated GPHS fuel pellets containing high levels of silicon were aged, loaded into GPHS module halves, and impacted against steel plates. The results show no significant differences between the response of these capsules and the behavior of relatively low-silicon fuel pellets tested previously. {copyright} {ital 1997 American Institute of Physics.}

  14. Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

    DOEpatents

    Lloyd, Milton H.

    1983-01-01

    Method for direct coprocessing of nuclear fuels derived from a product stream of a fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

  15. Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

    DOEpatents

    Lloyd, M.H.

    1981-01-09

    Method for direct coprocessing of nuclear fuels derived from a product stream of fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

  16. The thermal conductivity of mixed fuel UxPu1-xO2: molecular dynamics simulations

    SciTech Connect

    Liu, Xiang-Yang; Cooper, Michael William Donald; Stanek, Christopher Richard; Andersson, Anders David Ragnar

    2015-10-16

    Mixed oxides (MOX), in the context of nuclear fuels, are a mixture of the oxides of heavy actinide elements such as uranium, plutonium and thorium. The interest in the UO2-PuO2 system arises from the fact that these oxides are used both in fast breeder reactors (FBRs) as well as in pressurized water reactors (PWRs). The thermal conductivity of UO2 fuel is an important material property that affects fuel performance since it is the key parameter determining the temperature distribution in the fuel, thus governing, e.g., dimensional changes due to thermal expansion, fission gas release rates, etc. For this reason it is important to understand the thermal conductivity of MOX fuel and how it differs from UO2. Here, molecular dynamics (MD) simulations are carried out to determine quantitatively, the effect of mixing on the thermal conductivity of UxPu1-xO2, as a function of PuO2 concentrations, for a range of temperatures, 300 – 1500 K. The results will be used to develop enhanced continuum thermal conductivity models for MARMOT and BISON by INL. These models express the thermal conductivity as a function of microstructure state-variables, thus enabling thermal conductivity models with closer connection to the physical state of the fuel.

  17. LWR spent fuel reduction by the removal of U and the compact storage of Pu with FP for long-term nuclear sustainability

    SciTech Connect

    Fukasawa, T.; Hoshino, K.; Takano, M.; Sato, S.; Shimazu, Y.

    2013-07-01

    Fast breeder reactors (FBR) nuclear fuel cycle is needed for long-term nuclear sustainability while preventing global warming and maximum utilizing the limited uranium (U) resources. The 'Framework for Nuclear Energy Policy' by the Japanese government on October 2005 stated that commercial FBR deployment will start around 2050 under its suitable conditions by the successive replacement of light water reactors (LWR) to FBR. Even after Fukushima Daiichi Nuclear Power Plant accident which made Japanese tendency slow down the nuclear power generation activities, Japan should have various options for energy resources including nuclear, and also consider the delay of FBR deployment and increase of LWR spent fuel (LWR-SF) storage amounts. As plutonium (Pu) for FBR deployment will be supplied from LWR-SF reprocessing and Japan will not possess surplus Pu, the authors have developed the flexible fuel cycle initiative (FFCI) for the transition from LWR to FBR. The FFCI system is based on the possibility to stored recycled materials (U, Pu)temporarily for a suitable period according to the FBR deployment rate to control the Pu demand/supply balance. This FFCI system is also effective after the Fukushima accident for the reduction of LWR-SF and future LWR-to-FBR transition. (authors)

  18. Separation of actinides from rare earth elements by means of molten salt electrorefining with anodic dissolution of U Pu Zr alloy fuel

    NASA Astrophysics Data System (ADS)

    Kinoshita, Kensuke; Koyama, Tadafumi; Inoue, Tadashi; Ougier, Michel; Glatz, Jean-Paul

    2005-02-01

    Electrorefining is the main process for pyro-reprocessing of the fuel of a metallic fuel FBR. To obtain a basic knowledge of electrorefining technology, a series of experiments was carried out with unirradiated fuel alloy. The alloy, 71U 19Pu 10Zr (wt.%), was dissolved anodically into a molten LiCl KCl bath at 753 K. Simultaneously, Pu and U were recovered into the Cd cathode with small amounts of minor actinides, Zr and rare earth elements (REs). The separation factors of U, Np, Am, Cm and Ce against Pu, derived from the composition of recovered deposits and of the salt bath, were about 2.04, 0.949, 0.597, 0.534 and 0.0393, respectively, which are similar to the equilibrium values observed in a distribution experiment in a LiCl KCl/Cd system. This demonstrates that electrorefining achieves the separation of actinides from REs. The anodic dissolution of the alloy was found to progress from the outside, leaving a dense layer containing salt and Zr metal around the alloy surface. It was found that more than 99.9% of both U and Pu could be dissolved from the alloy and about 55% of Zr remained in this layer.

  19. Quantification of Uncertainties due to 235,238U, 239,240,241Pu and Fission Products Nuclear Data Uncertainties for a PWR Fuel Assembly

    NASA Astrophysics Data System (ADS)

    da Cruz, D. F.; Rochman, D.; Koning, A. J.

    2014-04-01

    Uncertainty analysis on reactivity and discharged inventory for a typical PWR fuel element as a result of uncertainties in 235,238U, 239,240,241Pu, and fission products nuclear data was performed. The Total Monte-Carlo (TMC) method was applied using the deterministic transport code DRAGON. The nuclear data used in this study is from the JEFF-3.1 evaluations, with the exception of the nuclear data files for U, Pu and fission products isotopes, which are taken from the nuclear data library TENDL-2012. Results show that the calculated total uncertainty in keff (as result of uncertainties in nuclear data of the considered isotopes) is virtually independent on fuel burnp and amounts to 700 pcm. The uncertainties in inventory of the discharged fuel is dependent on the element considered and lies in the range 1-15% for most fission products, and is below 5% for the most important actinides.

  20. Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

    NASA Astrophysics Data System (ADS)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-01-01

    Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

  1. Pu-Zr alloy for high-temperature foil-type fuel

    DOEpatents

    McCuaig, Franklin D.

    1977-01-01

    A nuclear reactor fuel alloy consists essentially of from slightly greater than 7 to about 4 w/o zirconium, balance plutonium, and is characterized in that the alloy is castable and is rollable to thin foils. A preferred embodiment of about 7 w/o zirconium, balance plutonium, has a melting point substantially above the melting point of plutonium, is rollable to foils as thin as 0.0005 inch thick, and is compatible with cladding material when repeatedly cycled to temperatures above 650.degree. C. Neutron reflux densities across a reactor core can be determined with a high-temperature activation-measurement foil which consists of a fuel alloy foil core sandwiched and sealed between two cladding material jackets, the fuel alloy foil core being a 7 w/o zirconium, plutonium foil which is from 0.005 to 0.0005 inch thick.

  2. Atmospheric deposition, resuspension, and root uptake of Pu in corn and other grain-producing agroecosystems near a nuclear fuel facility.

    PubMed

    Pinder, J E; McLeod, K W; Adriano, D C; Corey, J C; Boni, A L

    1990-12-01

    Plutonium released to the environment may contribute to dose to humans through inhalation or ingestion of contaminated foodstuffs. Plutonium contamination of agricultural plants may result from interception and retention of atmospheric deposition, resuspension of Pu-bearing soil particles to plant surfaces, and root uptake. Plutonium on vegetation surfaces may be transferred to grain surfaces during mechanical harvesting. Data obtained from corn grown near the U.S. Department of Energy's H-Area nuclear fuel chemical separations facility on the Savannah River Site were used to estimate parameters of a simple model of Pu transport in agroecosystems. The parameter estimates for corn were compared to those previously obtained for wheat and soybeans. Despite some differences in parameter estimates among crops, the relative importances of atmospheric deposition, resuspension, and root uptake were similar among crops. For even small deposition rates, the relative importances of processes for Pu contamination of corn grain should be: transfer of atmospheric deposition from vegetation surfaces to grain surfaces during combining greater than resuspension of soil to grain surfaces greater than root uptake. Approximately 3.9 X 10(-5) of a year's atmospheric deposition is transferred to grain. Approximately 6.2 X 10(-9) of the Pu inventory in the soil is resuspended to corn grain, and a further 7.3 X 10(-10) of the soil Pu inventory is absorbed and translocated to grains.

  3. Modeling of selected ceramic processing parameters employed in the fabrication of 238PuO2 fuel pellets

    SciTech Connect

    Brockman, R. A.; Kramer, D. P.; Barklay, C. D.; Cairns-Gallimore, D.; Brown, J. L.; Huling, J. C.; Van Pelt, C. E.

    2011-10-01

    Recent deep space missions utilize the thermal output of the radioisotope plutonium-238 as the fuel in the thermal to electrical power system. Since the application of plutonium in its elemental state has several disadvantages, the fuel employed in these deep space power systems is typically in the oxide form such as plutonium-238 dioxide (238PuO2). As an oxide, the processing of the plutonium dioxide into fuel pellets is performed via ''classical'' ceramic processing unit operations such as sieving of the powder, pressing, sintering, etc. Modeling of these unit operations can be beneficial in the understanding and control of processing parameters with the goal of further enhancing the desired characteristics of the 238PuO2 fuel pellets. A finite element model has been used to help identify the time-temperature-stress profile within a pellet during a furnace operation taking into account that 238PuO2 itself has a significant thermal output. The results of the modeling efforts will be discussed.

  4. New measurement of the 242Pu(n,γ) cross section at n_TOF-EAR1 for MOX fuels: Preliminary results in the RRR

    NASA Astrophysics Data System (ADS)

    Lerendegui-Marco, J.; Guerrero, C.; Cortés-Giraldo, M. A.; Quesada, J. M.; Mendoza, E.; Cano-Ott, D.; Eberhardt, K.; Junghans, A.; Aberle, O.; Andrzejewski, J.; Audouin, L.; Bacak, M.; Balibrea, J.; Barbagallo, M.; Bečvář, F.; Berthoumieux, E.; Billowes, J.; Bosnar, D.; Brown, A.; Caamaño, M.; Calviño, F.; Calviani, M.; Cardella, R.; Casanovas, A.; Cerutti, F.; Chen, Y. H.; Chiaveri, E.; Colonna, N.; Cortés, G.; Cosentino, L.; Damone, L. A.; Diakaki, M.; Domingo-Pardo, C.; Dressler, R.; Dupont, E.; Durán, I.; Fernández-Domínguez, B.; Ferrari, A.; Ferreira, P.; Finocchiaro, P.; Göbel, K.; Gómez-Hornillos, M. B.; García, A. R.; Gawlik, A.; Gilardoni, S.; Glodariu, T.; Gonçalves, I. F.; González, E.; Griesmayer, E.; Gunsing, F.; Harada, H.; Heinitz, S.; Heyse, J.; Jenkins, D. G.; Jericha, E.; Käppeler, F.; Kadi, Y.; Kalamara, A.; Kavrigin, P.; Kimura, A.; Kivel, N.; Kokkoris, M.; Krtička, M.; Kurtulgil, D.; Leal-Cidoncha, E.; Lederer, C.; Leeb, H.; Meo, S. Lo; Lonsdale, S. J.; Macina, D.; Marganiec, J.; Martínez, T.; Masi, A.; Massimi, C.; Mastinu, P.; Mastromarco, M.; Maugeri, E. A.; Mazzone, A.; Mengoni, A.; Milazzo, P. M.; Mingrone, F.; Musumarra, A.; Negret, A.; Nolte, R.; Oprea, A.; Patronis, N.; Pavlik, A.; Perkowski, J.; Porras, I.; Praena, J.; Radeck, D.; Rauscher, T.; Reifarth, R.; Rout, P. C.; Rubbia, C.; Ryan, J. A.; Sabaté-Gilarte, M.; Saxena, A.; Schillebeeckx, P.; Schumann, D.; Smith, A. G.; Sosnin, N. V.; Stamatopoulos, A.; Tagliente, G.; Tain, J. L.; Tarifeño-Saldivia, A.; Tassan-Got, L.; Valenta, S.; Vannini, G.; Variale, V.; Vaz, P.; Ventura, A.; Vlachoudis, V.; Vlastou, R.; Wallner, A.; Warren, S.; Woods, P. J.; Wright, T.; Žugec, P.

    2017-09-01

    The spent fuel of current nuclear reactors contains fissile plutonium isotopes that can be combined with 238U to make mixed oxide (MOX) fuel. In this way the Pu from spent fuel is used in a new reactor cycle, contributing to the long-term sustainability of nuclear energy. The use of MOX fuels in thermal and fast reactors requires accurate capture and fission cross sections. For the particular case of 242Pu, the previous neutron capture cross section measurements were made in the 70's, providing an uncertainty of about 35% in the keV region. In this context, the Nuclear Energy Agency recommends in its "High Priority Request List" and its report WPEC-26 that the capture cross section of 242Pu should be measured with an accuracy of at least 7-12% in the neutron energy range between 500 eV and 500 keV. This work presents a brief description of the measurement performed at n_TOF-EAR1, the data reduction process and the first ToF capture measurement on this isotope in the last 40 years, providing preliminary individual resonance parameters beyond the current energy limits in the evaluations, as well as a preliminary set of average resonance parameters.

  5. Development of self-interrogation neutron resonance densitometry (SINRD) to measure U-235 and Pu-239 content in a PWR spent fuel assembly

    SciTech Connect

    Lafleur, Adrienne M; Charlton, William S; Menlove, Howard O; Swinhoe, Martyn T

    2009-01-01

    The use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U and {sup 239}Pu content in a PWR spent fuel assembly was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in fission chamber. These simulations utilize the {sup 244}Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be measured using {sup 235}U and {sup 239}Pu fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the sensitivity of the measurements to extraneous material present in fuel. The development of SINRD to measure the fissile content in spent fuel is of great importance to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in FBR spent fuel and heavy metal product from reprocessing methods.

  6. Fabrication experience of Al- sup 233 U and Al-Pu plate fuel for the Purnima III and Kamini research reactors

    SciTech Connect

    Ganguly, C.; Prasad, G.J.; Mahule, N.; Ghosh, J.K.; Assari, K.V.J.; Chandrasekharan, K.N.P.; Muralidhar, S.; Balan, T.S.; Roy, P.R. )

    1991-10-01

    This paper reports on aluminum-clad Al-20 wt% {sup 233}U and Al-23 wt% Pu plate fuel subassemblies that have been fabricated for the Purnima III critical facility and the Kamini research reactor. The fabrication flow sheet consists of preparing the master alloy using aluminum and uranium or plutonium metals as feed materials, remelting and casting the fuel alloy ingots, rolling, picture framing and sandwiching the fuel alloy between aluminum sheets, roll bonding, locating the fuel alloy core outline by x-ray radiography, and trimming and machining to final dimensions. Metallic molds produce better ingots than graphite ones. The addition of zirconium during melting improves the microstructure of the Al-U and Al-Pu castings and facilitates hot rolling of the ingots. In the subassembly the fuel plates are finally locked in aluminum spacer grooves by a novel roll-swaging technique. High-resolution x-ray radiographs and microdensitometric scans are utilized to confirm the homogeneous distribution of the fissile material in the fuel plates. Nonbond areas are detected by blister testing and immersion ultrasonic testing of the roll-bonded fuel plates.

  7. An initial assessment of a mechanistic model, GRASS-SST, in U-Pu-Zr metallic alloy fuel fission-gas behavior simulations

    NASA Astrophysics Data System (ADS)

    Yun, Di; Rest, Jeffrey; Hofman, Gerard L.; Yacout, Abdellatif M.

    2013-04-01

    A mechanistic kinetic rate theory model originally developed for the prediction of fission gas behavior in oxide nuclear fuels under steady-state and transient conditions has been assessed to investigate its applicability to model fission gas behavior in U-Pu-Zr metallic alloy fuel. In order to capture and validate the underlying physics for irradiated U-Pu-Zr fuels, the mechanistic model was applied to evaluate fission gas release, fission gas and fission product induced swelling, and detailed gas bubble size distributions in three different fuel zones: the outer α-U, the intermediate, and the inner γ-U zones. Due to its special microstructural features, the α-U zone in U-Pu-Zr fuels is believed to contribute the largest fraction of fission gas release among the different fuel zones. It is shown that with the use of small effective grain sizes, the mechanistic model can predict fission gas release that is in reasonable consistence with (though slightly lower than) experimentally measured data. These simulation results are comparable to the experimentally measured fission gas release since the mechanism of fission gas transport through the densely distributed laminar porosity in the α-U zone is analogous to the mechanism of fission gas transport through the interconnected gas bubble porosity utilized in the mechanistic model. Detailed gas bubble size distributions predicted with the mechanistic model in both the intermediate zone and the high temperature γ-U zone of U-Pu-Zr fuel are also compared to experimental measurements from available SEM micrographs. These comparisons show good agreement between the simulation results and experimental measurements, and therefore provide crucial guidelines for the selection of key physical parameters required for modeling these two zones. Material properties such as fuel grain size and thermal diffusivity of gas and model parameters such as di-atom nucleation probability and gas bubble re-solution constant are predicted

  8. Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

    NASA Astrophysics Data System (ADS)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-07-01

    Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (DPuN fuel with fissile contain from Plutonium waste LWR for each geometry. The minimum power density is around 72 Watt/cc, and maximum power density 114 Watt/cc. After we calculate with various geometry core, when we use the balance geometry, the k-eff value flattest and more stable than the others.

  9. Preliminary Study on LiF4-ThF4-PuF4 Utilization as Fuel Salt of miniFUJI Molten Salt Reactor

    NASA Astrophysics Data System (ADS)

    Waris, Abdul; Aji, Indarta K.; Pramuditya, Syeilendra; Widayani; Irwanto, Dwi

    2016-08-01

    miniFUJI reactor is molten salt reactor (MSR) which is one type of the Generation IV nuclear energy systems. The original miniFUJI reactor design uses LiF-BeF2-ThF4-233UF4 as a fuel salt. In the present study, the use of LiF4-ThF4-PuF4 as fuel salt instead of LiF-BeF2-ThF4-UF4 will be discussed. The neutronics cell calculation has been performed by using PIJ (collision probability method code) routine of SRAC 2006 code, with the nuclear data library is JENDL-4.0. The results reveal that the reactor can attain the criticality condition with the plutonium concentration in the fuel salt is equal to 9.16% or more. The conversion ratio diminishes with the enlarging of plutonium concentration in the fuel. The neutron spectrum of miniFUJI MSR with plutonium fuel becomes harder compared to that of the 233U fuel.

  10. The prospect of uranium nitride (UN-PuN) fuel for 25- 100MWe gas cooled fast reactor long life without refuelling

    NASA Astrophysics Data System (ADS)

    Syarifah, R. D.; Su'ud, Z.; Basar, K.; Irwanto, D.

    2016-11-01

    The prospect of uranium nitride (UN-PuN) fuel for 25-100MWe Gas Cooled Fast Reactor has been done. This research use helium coolant which has low neutron moderation, chemical inert and single phase. This study use natural uranium and plutonium. Plutonium taken from spent fuel of LWR (Light Water Reactor). So, it can reduced spent fuel in the world. The calculation use SRAC2006 and JENDL 4.0 for the data libraries. First, we calculate PIJ for fuel pin cell calculation and CITATION for core calculation. The reflector radial-axial width is 50 cm. The variation of fuel fraction is 40% until 65%, cladding 10%, and moderator 25% up to 50%. The variation of the power is 75-300 MWth (25-100 MWe). The calculation of survey parameter has been done. The variation of percentage plutonium is 7% up to 13%. We have optimum k-eff value in percentage of plutonium 11%. The high powers cause k-eff value high too. Second, the core configuration divided by three variation fuel (F1, F2, and F3). F1 is located in the central core, F2 middle core and F3 outer core. The variation percentage Plutonium for fuel F1:F2:F3 = 8%:10%:12%. The increasing power level make the burn up level increase. All case can reach burn up time plus than 20 years. The thermal powers increase cause the peak power density increase. The power 150 MWth, 225 MWth, and 300 MWth have excess reactivity (%Ak/k) less than 2%.

  11. Microstructural Changes In Thermally Cycled U-Pu-Zr-Am-Np Metallic Transmutation Fuel With 1.5% Lanthanides

    SciTech Connect

    Dawn E. Janney; J. Rory Kennedy

    2008-06-01

    The United States Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) is developing metallic actinide-zirconium alloy fuels for the transmutation of minor actinides as part of a closed fuel cycle. The molten salt electrochemical process to be used for fuel recycle has the potential to carry over up to 2% fission product lanthanide content into the fuel fabrication process. Within the scope of the fuel irradiation testing program at Idaho National Laboratory (INL), candidate metal alloy transmutation fuels containing quantities of lanthanide elements have been fabricated, characterized, and delivered to the Advanced Test Reactor for irradiation testing.

  12. Application of density functional theory in assessing properties of thoria and recycled fuels

    NASA Astrophysics Data System (ADS)

    Szpunar, B.; Szpunar, J. A.

    2013-08-01

    The application of the Density Functional Theory (DFT) approximation to assess the mechanical and structural properties of recycled urania and thoria fuel is presented. The total energy technique based on the local density approximation plus Hubbard U as implemented in the CASTEP code is used. The calculated values of the lattice constants and mechanical moduli of ThO2 and UO2 agree with the experimental data. However only non-local hybrid functional (B3LYP) leads to a larger band gap (6.9 eV) of thoria, in better agreement with experiment (6 eV) than value (4.7 eV) calculated using the LDA + U (6 eV) scheme. The calculations are further expanded to study lattice constants of (Pu, U) oxides and U3O8 for which we currently do not have experimental data. The elastic moduli of ThO2, UO2 and (Pu, U) oxides are compared.

  13. Large area quantitative X-ray mapping of (U,Pu)O 2 nuclear fuel pellets using wavelength dispersive electron probe microanalysis

    NASA Astrophysics Data System (ADS)

    Brémier, S.; Haas, D.; Somers, J.; Walker, C. T.

    2003-04-01

    The work presented is an example of how large area compositional mapping (≥1 mm 2) can be used to provide quantitative information on element distribution and specimen homogeneity. High-resolution was accomplished by producing a collage of X-ray maps acquired using classical conditions; magnification ×400, spatial resolution 256×256 pixels. The individual images, each measuring roughly 250×250 μm, were converted to quantitative maps using the HIMAX® software package and the XMAS® matrix correction from SAMx. The quantitative gray-level large area X-ray picture was pieced together using the 'Multiple Image Alignment' function of the ANALYSIS® image processing software. This software was also used to convert the gray-level pictures to false color images. The specimens investigated were transverse sections of MOX fuel pellets. Results are presented for the distribution of Pu by area fraction and cumulative area fraction, the size distribution of regions of high Pu concentration and average separation of these regions.

  14. Determination of total Pu content in a Spent Fuel Assembly by Measuring Passive Neutron Count rate and Multiplication with the Differential Die-Away Instrument

    SciTech Connect

    Henzl, Vladimir; Croft, Stephen; Swinhoe, Martyn T.; Tobin, Stephen J.

    2012-07-18

    A key objective of the Next Generation Safeguards Initiative (NGSI) is to evaluate and develop non-destructive assay (NDA) techniques to determine the elemental plutonium content in a commercial-grade nuclear spent fuel assembly (SFA) [1]. Within this framework, we investigate by simulation a novel analytical approach based on combined information from passive measurement of the total neutron count rate of a SFA and its multiplication determined by the active interrogation using an instrument based on a Differential Die-Away technique (DDA). We use detailed MCNPX simulations across an extensive set of SFA characteristics to establish the approach and demonstrate its robustness. It is predicted that Pu content can be determined by the proposed method to a few %.

  15. Thermodynamic assessment of the LiF-CeF3-ThF4 system: Prediction of PuF3 concentration in a molten salt reactor fuel

    NASA Astrophysics Data System (ADS)

    Beneš, O.; Konings, R. J. M.

    2013-04-01

    A thermodynamic description of the LiF-CeF3-ThF4 system is made in this study using a two-sublattice model for the description of the solid solution and a quasi-chemical model based on quadruplet approximation for the liquid phase. New calorimetric experimental data of the binary LiF-CeF3, CeF3-ThF4 and ternary LiF-CeF3-ThF4 systems have been obtained in this work justifying the calculated phase diagrams. Using the obtained thermodynamic assessment the concentration of PuF3 in the LiF-ThF4 melt was estimated based on the similarities with CeF3 and the melting behaviour of the initial molten salt fast reactor fuel was discussed.

  16. A Deterministic Study of the Deficiency of the Wigner-Seitz Approximation for Pu/MOX Fuel Pins

    SciTech Connect

    DeHart, M.D.

    1999-09-27

    The Wigner-Seitz pin-cell approximation has long been applied as a modeling approximation in analysis of UO2 lattice fuel cells. In the past, this approximation has been appropriate for such fuel. However, with increasing attention drawn to mixed-oxide (MOX) fuels with significant plutonium content, it is important to understand the implications of the approximation in a uranium-plutonium matrix. The special geometric capabilities of the deterministic NEWT computer code have been used to assess the adequacy of the Wigner-Seitz cell in such an environment, as part of a larger study of computational aspects of MOX fuel modeling. Results of calculations using various approximations and boundary conditions are presented, and are validated by comparison to results obtained using KENO V.a and XSDRNPM.

  17. Pyrochemical processes for the recovery of weapons grade plutonium either as a metal or as PuO{sub 2} for use in mixed oxide reactor fuel pellets

    SciTech Connect

    Colmenares, C.A.; Ebbinghaus, B.B.; Bronson, M.C.

    1995-11-03

    The authors have developed two processes for the recovery of weapons grade Pu, as either Pu metal or PuO{sub 2}, that are strictly pyrochemical and do not produce any liquid waste. Large amounts of Pu metal (up to 4 kg.), in various geometric shapes, have been recovered by a hydride/dehydride/casting process (HYDEC) to produce metal ingots of any desired shape. The three processing steps are carried out in a single compact apparatus. The experimental technique and results obtained will be described. The authors have prepared PuO{sub 2} powders from weapons grade Pu by a process that hydrides the Pu metal followed by the oxidation of the hydride (HYDOX process). Experimental details of the best way to carry out this process will be presented, as well as the characterization of both hydride and oxide powders produced.

  18. Fuel/cladding compatibility in high-burnup U-19Pu-10Zr/HT9-clad fuel at elevated temperatures

    SciTech Connect

    Cohen, A.B.; Tsai, H.; Neimark, L.A.

    1992-11-01

    This paper summarizes the most recent results of a continuing experimental effort to study compatibility issues of irradiated metallic fuel and cladding at elevated temperatures that may be encountered beyond those of nominal steady-state conditions.

  19. Irradiation of Argentine (U,Pu)O 2 MOX fuels. Post-irradiation results and experimental analysis with the BACO code

    NASA Astrophysics Data System (ADS)

    Marino, Armando Carlos; Pérez, Edmundo; Adelfang, Pablo

    1996-04-01

    The irradiation of the first Argentine prototypes of pressurized heavy water reactor (PHWR) (U,Pu)O 2 MOX fuels began in 1986. These experiments were carried out in the High Flux Reactor (HFR)-Petten, Holland. The rods were prepared and controlled in the CNEA's α Facility. The postirradiation examinations were performed in the Kernforschungszentrum, Karlsruhe, Germany and in the Joint Research Center (JRC), Petten. The first rod has been used for destructive pre-irradiation analysis. The second one as a pathfinder to adjust systems in the HFR. Two additional rods including iodine doped pellets were intended to simulate 15 000 MWd/T(M) burnup. The remaining two rods were irradiated until 15 000 MWd/T(M). One of them underwent a final ramp with the aim of verifying fabrication processes and studying the behaviour under power transients. BACO (BArra COmbustible) code was used to define the power histories and to analyse the experiments. This paper presents a description of the different experiments and a comparison between the results of the postirradiation examinations and the BACO outputs.

  20. Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

    NASA Astrophysics Data System (ADS)

    Muratov, V. G.; Lopatkin, A. V.

    An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

  1. Determination of Pu content in a Spent Fuel Assembly by Measuring Passive Total Neutron count rate and Multiplication with the Differential Die-Away Instrument

    SciTech Connect

    Henzl, Vladimir; Croft, Stephen; Swinhoe, Martyn T.; Tobin, Stephen J.

    2012-07-13

    Inspired by approach of Bignan and Martin-Didier (ESARDA 1991) we introduce novel (instrument independent) approach based on multiplication and passive neutron. Based on simulations of SFL-1 the accuracy of determination of {sup tot}Pu content with new approach is {approx}1.3-1.5%. Method applicable for DDA instrument, since it can measure both multiplication and passive neutron count rate. Comparison of pro's & con's of measuring/determining of {sup 239}Pu{sub eff} and {sup tot}Pu suggests a potential for enhanced diversion detection sensitivity.

  2. Inventories of 239+240Pu, 137Cs, and excess 210Pb in sediments from freshwater and brackish lakes in Rokkasho, Japan, adjacent to a spent nuclear fuel reprocessing plant.

    PubMed

    Ueda, Shinji; Ohtsuka, Yoshihito; Kondo, Kunio; Hisamatsu, Shun'ichi

    2009-10-01

    We investigated the vertical profiles of (239+240)Pu, (137)Cs, and excess (210)Pb ((210)Pb(ex)) in sediment core samples obtained from two freshwater lakes and two brackish lakes situated near the first commercial spent nuclear fuel reprocessing plant in Rokkasho, Japan, before the final test of the plant using actual spent nuclear fuel. The inventory of (239+240)Pu in those lakes was larger than that in soil in Rokkasho, which indicated the inflow of (239+240)Pu from the catchment area in addition to direct deposition on the lake surfaces. The (137)Cs inventory in sediments of the brackish lakes was lower than that in the soil, which showed that part of the (137)Cs was removed from the sediments by the brackish water or that it was not deposited into the sediments, because of the high solubility of Cs in brackish water. The (137)Cs inventory in sediments of the freshwater lakes was higher than that of the brackish lakes, and comparable with that in soil except for one core sample out of four. The (239+240)Pu/(137)Cs ratio in freshwater lake sediments was higher than that in soil, and that indicated that part of the (137)Cs was lost from the sediments. The low inventory of (137)Cs may be attributable to competition for absorption sites in sediments with ammonium ions formed in the reducing environment which occurs from summer to fall in the sediments. Those data will be used as background data on the artificial radionuclides in the lakes to assess the effect of released radionuclides on their concentrations.

  3. Plutonium isotopic analysis system for plutonium samples enriched in sup 238 Pu in EP 60/61 and fuel-clad containers

    SciTech Connect

    Ruhter, W.D.

    1991-07-01

    This two-part manual describes and provides instructions for installing software for Lawrence Livermore National Laboratory's Pu-238 isotopic analysis system built for Westinghouse Hanford's Radioisotope Power Systems Facility. Part 1 contains descriptions of all the subroutines found in the main software program, WHC.ASY238. Also provided in this part are general instructions for modifying a subroutine and specific directions for relinking the WHC.ASY238 program, as well as information on the supporting program PU238.CHNG. Part 2 contains listings of the Pu-238 isotopic analysis system codes. The system uses a large (20% rel. efficiency), coaxial, n-type germanium detector (COAX). Parameter files for the detector have filenames with IS8 extensions. Spectral data files also have WH8 and I01, I02, etc. filename extensions.

  4. Photoelectron Spectroscopy of U Oxide at LLNL

    SciTech Connect

    Tobin, J G; Yu, S; Chung, B W; Waddill, G D

    2010-03-02

    In our laboratory at LLNL, an effort is underway to investigate the underlying complexity of 5f electronic structure with spin-resolved photoelectron spectroscopy using chiral photonic excitation, i.e. Fano Spectroscopy. Our previous Fano measurements with Ce indicate the efficacy of this approach and theoretical calculations and spectral simulations suggest that Fano Spectroscopy may resolve the controversy concerning Pu electronic structure and electron correlation. To this end, we have constructed and commissioned a new Fano Spectrometer, testing it with the relativistic 5d system Pt. Here, our preliminary photoelectron spectra of the UO{sub 2} system are presented. X-ray photoelectron spectroscopy has been used to characterize a sample of UO{sub 2} grown on an underlying substrate of Uranium. Both AlK{alpha} (1487 eV) and MgK{alpha} (1254 eV) emission were utilized as the excitation. Using XPS and comparing to reference spectra, it has been shown that our sample is clearly UO{sub 2}.

  5. Determination of 240Pu/239Pu atom ratio in coastal surface seawaters from the western North Pacific Ocean and Japan Sea.

    PubMed

    Yamada, Masatoshi; Zheng, Jian

    2008-01-01

    Surface seawater samples were collected from a site in the vicinity of the nuclear fuel reprocessing facility at Rokkasho, Japan and sites along the Japan Sea coast. (239+240)Pu activities and (240)Pu/(239)Pu atom ratios were determined by alpha-spectrometry and isotope-dilution sector-field ICP-MS. The (240)Pu/(239)Pu atom ratio with the mean value of 0.227 +/- 0.006 was significantly higher than the mean global fallout ratio of 0.18. The contribution of the Pacific Proving Grounds close-in fallout was estimated to be 33% of the (239+240)Pu.

  6. Redefining design criteria for Pu-238 gloveboxes

    SciTech Connect

    Acosta, S.V.

    1998-12-31

    Enclosures for confinement of special nuclear materials (SNM) have evolved into the design of gloveboxes. During the early stages of glovebox technology, established practices and process operation requirements defined design criteria. Proven boxes that performed and met or exceeded process requirements in one group or area, often could not be duplicated in other areas or processes, and till achieve the same success. Changes in materials, fabrication and installation methods often only met immediate design criteria. Standardization of design criteria took a big step during creation of ``Special-Nuclear Materials R and D Laboratory Project, Glovebox standards``. The standards defined design criteria for every type of process equipment in its most general form. Los Alamos National Laboratory (LANL) then and now has had great success with Pu-238 processing. However with ever changing Environment Safety and Health (ES and H) requirements and Ta-55 Facility Configuration Management, current design criteria are forced to explore alternative methods of glovebox design fabrication and installation. Pu-238 fuel processing operations in the Power Source Technologies Group have pushed the limitations of current design criteria. More than half of Pu-238 gloveboxes are being retrofitted or replaced to perform the specific fuel process operations. Pu-238 glovebox design criteria are headed toward process designed single use glovebox and supporting line gloveboxes. Gloveboxes that will house equipment and processes will support TA-55 Pu-238 fuel processing needs into the next century and extend glovebox expected design life.

  7. Purification of 238Pu Oxide using the Pu Oxalate Process

    SciTech Connect

    Mew, D A; Krikorian, O H; Dodson, K E; Schmitz, J A

    2001-11-28

    The Pu oxalate process is used to remove {sup 234}U from aged {sup 238}Pu-enriched PuO{sub 2} ({sup 234}U grows into the PuO{sub 2} material with time from a-decay of {sup 238}Pu). The Pu oxalate process was first used on a mixture of weapons grade PuO{sub 2} with UO{sub 2} to work out the processing parameters. It was then applied to aged {sup 238}Pu-enriched PuO{sub 2} ({sup 238}PuO{sub 2}). The {sup 234}U content of the {sup 238}PuO{sub 2} was reduced from 13.2 wt% to 0.0254 wt%, and the Pu recovery yield was 78.5%. The process is complex and is complicated by radiolysis problems when working with {sup 238}Pu. Details of the experiments are described.

  8. Plutonium isotopic analysis system for plutonium samples enriched in sup 238 Pu in EP 60/61 and fuel-clad containers

    SciTech Connect

    Ruhter, W.D.

    1991-07-01

    This software manual is addressed to the Westinghouse Hanford's Radioisotope Power Systems Facility personnel (programmers and supervisors) who maintain the software on the Pu-238 isotopic analysis system. The document is divided into two parts. Part 1 describes the computer codes that control the system, analyze the spectral data, and determine the relative plutonium abundances. Part 2 contains the software listing of the analysis codes.

  9. Plutonium isotopic analysis system for plutonium samples enriched in {sup 238}Pu in EP 60/61 and fuel-clad containers. Volume 3, Part 2: Software listings

    SciTech Connect

    Ruhter, W.D.

    1991-07-01

    This software manual is addressed to the Westinghouse Hanford`s Radioisotope Power Systems Facility personnel (programmers and supervisors) who maintain the software on the Pu-238 isotopic analysis system. The document is divided into two parts. Part 1 describes the computer codes that control the system, analyze the spectral data, and determine the relative plutonium abundances. Part 2 contains the software listing of the analysis codes.

  10. Sintering and characterization of (Pu,Zr)N

    NASA Astrophysics Data System (ADS)

    Pukari, Merja; Takano, Masahide; Nishi, Tsuyoshi

    2014-01-01

    Nitride fuel, with the composition of (Pu0.4Zr0.6)N, is fabricated for studying the sinterability of nitride fuel as a function of oxygen concentration in the material. Oxygen concentration of up to 0.6 wt% evidently enhances the densification of the material. Increasing the sintering temperature from 1923 to 1973 K improves the sintered pellet densities by up to 3.8%TD. In addition, the measured thermophysical and electrical properties of (Pu0.4Zr0.6)N reveal that the values are close to those of PuN. Elevated oxygen concentration in the material decreases its thermal conductivity. Oxygen concentration of 0.34 wt% in (Pu,Zr)N is a consequence of the fabrication process, considering the relatively pure ZrN (0.03 wt% O) and PuN (0.08 wt% O) powders initially fabricated.

  11. Adsorption behaviour of PuF6 on UO2F2 by the use of 236Pu

    NASA Astrophysics Data System (ADS)

    Sato, Nobuaki; Matsuda, Minoru; Mitsugashira, Toshiaki; Kirishima, Akira

    2010-03-01

    To know the behavior of plutonium in the fluoride volatility process (FLUOREX PROCESS) for the spent nuclear fuel, both UO2 and PuO2 are fluorinated by fluorine forming volatile UF6 and PuF6, respectively. Then PuF6 is separated and recovered from UF6 by using adsorption materials such as uranyl fluoride UO2F2. In this paper, adsorption behavior of PuF6on UO2F2 was examined by the use of 236Pu tracer. First, the stability of UO2F2 in F2atmosphere was analyzed by TG-DTA method showing that uranium volatilized completely over 350 °C by the formation of UF6 and the adsorption of plutonium by UO2F2 should be done at temperatures lower than 250 °C. The behavior of PtF6 as a chemical analogue of PuF6 was also conducted for comparison and it showed that the deposition of PtF4 on UO2F2 at 200 °C. When the 236Pu doped U3O8 was reacted with 10%F2-He gas, the PuF6 vaporized at ca. 600 °C. Then adsorption of 236Pu on UO2F2 was observed by α ray measurement. The adsorption mechanism of Pu on UO2F2 was discussed with experimental data and thermodynamic consideration.

  12. AC-3-irradiation test of sphere-pac and pellet (U,Pu)C fuel in the US Fast Flux Test Facility

    NASA Astrophysics Data System (ADS)

    Bart, G.; Botta, F. B.; Hoth, C. W.; Ledergerber, G.; Mason, R. E.; Stratton, R. W.

    2008-05-01

    The objective of the AC-3 bundle experiment in the Fast Flux Test Facility (FFTF) was to evaluate a fuel fabrication method by 'direct conversion' of nitrate solutions into spherical uranium-plutonium carbide particles and to compare the irradiation performance of 'sphere-pac' fuel pins prepared at Paul Scherrer Institute (PSI) with standard pellet fuel pins fabricated at Los Alamos National Laboratory (LANL). The irradiation and post test examination results show that mixed carbide pellet fuel produced by powder methods and sphere-pac particle fuel developed by internal gelation techniques are both valuable advanced fuel candidates for liquid metal reactors. The PSI fabrication process with direct conversion of actinide nitrate solutions into various sizes of fuel spheres by internal gelation and direct filling of spheres into cladding tubes is seen as more easily transferable to remote operation, showing a significant reduction of process steps. The process is also adaptable for the fabrication of carbonitrides and nitrides (still based on a uranium matrix), as well as for actinides diluted in a (uranium-free) yttrium stabilized zirconium oxide matrix. The AC-3 fuel bundle was irradiated in the Fast Flux Test Facility (FFTF) during the years 1986-1988 for 630 full power days to a peak burn up of ˜8 at.% fissile material. All of the pins, irradiated at linear powers of up to 84 kW/m, with cladding outer temperatures of 465 °C appeared to be in good condition when removed from the assembly. The rebirth of interest for fast reactor systems motivated the earlier teams to report about the excellent, still perfectly relevant results reached; this paper focusing on the sphere-pac fuel behaviour.

  13. Plutonium isotopic analysis system for plutonium samples enriched in sup 238 Pu in EP 60/61 and fuel-clad containers

    SciTech Connect

    Ruhter, W.D.

    1991-07-01

    This user's manual is addressed to the Hanford Site personnel (routine operators and supervisors) who perform measurements with the Pu-238 isotopic analysis system. Each chapter begins with a table of contents that lists the section titles, illustrations, and tabular data presented in that chapter. The first chapter in this manual is an introduction to the system. Chapter 2 lists required settings for the system's commercial nuclear instrument modules. System operating procedures are given in Chapter 3. Chapter 4 contains routine and supervisorial operator interactions. Chapter 5 describes the system's short- and long-printout output formats. Chapter 6 gives instructions for changing system parameters. Error messages are listed and described in Chapter 7. Chapter 8 contains reference articles on measuring relative plutonium isotopics in solid samples.

  14. Anaerobic Biotransformation and Mobility of Pu and Pu-EDTA

    SciTech Connect

    Bolton, H., Jr.; Bailey, V.L.; Plymale, A.E.; Rai, D.; Xun, L.

    2006-04-05

    The complexation of radionuclides (e.g., plutonium (Pu) and {sup 60}Co) by co-disposed ethylenediaminetetraacetate (EDTA) has enhanced their transport in sediments at DOE sites. Pu(IV)-EDTA is not stable in the presence of relatively soluble Fe(III) compounds. Since most DOE sites have Fe(III) containing sediments, Pu(IV) is likely not the mobile form of Pu-EDTA. The only other Pu-EDTA complex stable in groundwater relevant to DOE sites would be Pu(III)-EDTA, which only forms under anaerobic conditions. Research is therefore needed to investigate the biotransformation of Pu and Pu-EDTA under anaerobic conditions and the anaerobic biodegradation of Pu-EDTA. The biotransformation of Pu and Pu-EDTA under various anaerobic regimes is poorly understood including the reduction kinetics of Pu(IV) to Pu(III) from soluble (Pu(IV)-EDTA) and insoluble Pu(IV), the redox conditions required for this reduction, the strength of the Pu(III)-EDTA, how the Pu(III)-EDTA competes with other dominant anoxic soluble metals (e.g., Fe(II)), and the oxidation kinetics of Pu(III)-EDTA. Finally, soluble Pu(III)-EDTA under anaerobic conditions would require anaerobic degradation of the EDTA to limit Pu(III) transport. Anaerobic EDTA degrading microorganisms have never been isolated. Recent results have shown that Shewanella oneidensis MR-1, a dissimilatory metal reducing bacterium, can reduce Pu(IV) to Pu(III). The Pu(IV) was provided as insoluble PuO2. The highest rate of Pu(IV) reduction was with the addition of AQDS, an electron shuttle. Of the total amount of Pu solubilized (i.e., soluble through a 0.36 nm filter), approximately 70% was Pu(III). The amount of soluble Pu was between 4.8 and 3.2 micromolar at day 1 and 6, respectively, indicating rapid reduction. The micromolar Pu is significant since the drinking water limit for Pu is 10{sup -12} M. On-going experiments are investigating the influence of EDTA on the rate of Pu reduction and the stability of the formed Pu(III). We have also

  15. Anaerobic Biotransformation and Mobility of Pu and Pu-EDTA

    SciTech Connect

    Bolton, H., Jr.; Rai, D.; Xun, L.

    2005-04-18

    The complexation of radionuclides (e.g., plutonium (Pu) and {sup 60}Co) by codisposed ethylenediaminetetraacetate (EDTA) has enhanced their transport in sediments at DOE sites. Our previous NABIR research investigated the aerobic biodegradation and biogeochemistry of Pu(IV)-EDTA. Plutonium(IV) forms stable complexes with EDTA under aerobic conditions and an aerobic EDTA degrading bacterium can degrade EDTA in the presence of Pu and decrease Pu mobility. However, our recent studies indicate that while Pu(IV)-EDTA is stable in simple aqueous systems, it is not stable in the presence of relatively soluble Fe(III) compounds (i.e., Fe(OH){sub 3}(s)--2-line ferrihydrite). Since most DOE sites have Fe(III) containing sediments, Pu(IV) in likely not the mobile form of Pu-EDTA in groundwater. The only other Pu-EDTA complex stable in groundwater relevant to DOE sites would be Pu(III)-EDTA, which only forms under anaerobic conditions. Research is therefore needed in this brand new project to investigate the biotransformation of Pu and Pu-EDTA under anaerobic conditions. The biotransformation of Pu and Pu-EDTA under various anaerobic regimes is poorly understood including the reduction kinetics of Pu(IV) to Pu(III) from soluble (Pu(IV)-EDTA) and insoluble Pu(IV) as PuO2(am) by metal reducing bacteria, the redox conditions required for this reduction, the strength of the Pu(III)-EDTA complex, how the Pu(III)-EDTA complex competes with other dominant anoxic soluble metals (e.g., Fe(II)), and the oxidation kinetics of Pu(III)-EDTA. Finally, the formation of a stable soluble Pu(III)-EDTA complex under anaerobic conditions would require degradation of the EDTA complex to limit Pu(III) transport in geologic environments. Anaerobic EDTA degrading microorganisms have not been isolated. These knowledge gaps preclude the development of a mechanistic understanding of how anaerobic conditions will influence Pu and Pu-EDTA fate and transport to assess, model, and design approaches to stop

  16. Control of Urania Crystallite Size by HMTA-Urea Reactions in the Internal Gelation Process for Preparing (U, Pu)O2Fuel Kernels

    SciTech Connect

    Collins, J.L.

    2005-04-26

    In the development of (U,Pu)O{sub 2} kernels by the internal gelation process for the Direct Press Spheroidized process at Oak Ridge National Laboratory, a novel crystal growth step was discovered that made it possible to prepare calcined porous kernels that could be used as direct-press feed for Fast Breeder Reactor pellet fabrication. High-quality pellets were prepared that were near theoretical density and that (upon examination) revealed no evidence of sphere remnants. The controlled crystal growth step involved using hexamethylenetetramine (HMTA)-urea stock solutions that were boiled for 60 min or less. Before this discovery, all the other crystal growth steps (when utilized) could reduce the tap density to only {approx}1.3 g/cm{sup 3}, which was not sufficiently low for use in ideal pellet pressing. The use of the boiled HMTA-urea solution allowed the tap density to be lowered to 0.93 g/cm{sup 3}, with the ideal density being about 1.0 g/cm{sup 3}. This report describes the development of this technology and its scaleup.

  17. Biotransformation of PuEDTA: Implications to Pu Immobilization

    SciTech Connect

    Bolton, Harvey, Jr.

    2006-06-01

    This project integrates three distinct goals to develop a fundamental understanding of the potential fate and disposition of plutonium in sediments that are co-contaminated with EDTA. The three objectives are: (1) Develop thermodynamic data for Pu-EDTA species and determine the dominant mobile form of Pu under anaerobic conditions. (2) Elucidate the mechanism and rates of Pu(IV) and Pu(IV)-EDTA reduction by metal-reducing bacteria and determine where the Pu is located (in solution, biosorbed, bioaccumulated). (3) Enrich and isolate anaerobic EDTA-degrading microorganisms to investigate the anaerobic biodegradation of Pu-EDTA.

  18. DEVELOPMENT PROGRAM FOR PU-238 AQUEOUS RECOVERY PROCESS

    SciTech Connect

    M. PANSOY-HJELVIK; M. REIMUS; ET AL

    2000-10-01

    Aqueous processing is necessary for the removal of impurities from {sup 238}Pu dioxide ({sup 238}PuO{sub 2}) fuel due to unacceptable levels of {sup 234}U and other non-actinide impurities in the scrap fuel. Impurities at levels above General Purpose Heat Source (GPHS) fuel specifications may impair the performance.of the heat sources. Efforts at Los Alamos have focused on developing the bench scale methodology for the aqueous process steps which includes comminution, dissolution, ion exchange, precipitation, and calcination. Recently, work has been performed to qualify the bench scale methodology, to show that the developed process produces pure {sup 238}PuO{sub 2} meeting GPHS fuel specifications. In addition, this work has enabled us to determine how waste volumes may be minimized during full-scale processing. Results of process qualification for the bench scale aqueous recovery operation and waste minimization efforts are presented.

  19. A Brief Review of Past INL Work Assessing Radionuclide Content in TMI-2 Melted Fuel Debris: The Use of 144Ce as a Surrogate for Pu Accountancy

    SciTech Connect

    D. L. Chichester; S. J. Thompson

    2013-09-01

    This report serves as a literature review of prior work performed at Idaho National Laboratory, and its predecessor organizations Idaho National Engineering Laboratory (INEL) and Idaho National Engineering and Environmental Laboratory (INEEL), studying radionuclide partitioning within the melted fuel debris of the reactor of the Three Mile Island 2 (TMI-2) nuclear power plant. The purpose of this review is to document prior published work that provides supporting evidence of the utility of using 144Ce as a surrogate for plutonium within melted fuel debris. When the TMI-2 accident occurred no quantitative nondestructive analysis (NDA) techniques existed that could assay plutonium in the unconventional wastes from the reactor. However, unpublished work performed at INL by D. W. Akers in the late 1980s through the 1990s demonstrated that passive gamma-ray spectrometry of 144Ce could potentially be used to develop a semi-quantitative correlation for estimating plutonium content in these materials. The fate and transport of radioisotopes in fuel from different regions of the core, including uranium, fission products, and actinides, appear to be well characterized based on the maximum temperature reached by fuel in different parts of the core and the melting point, boiling point, and volatility of those radioisotopes. Also, the chemical interactions between fuel, fuel cladding, control elements, and core structural components appears to have played a large role in determining when and how fuel relocation occurred in the core; perhaps the most important of these reaction appears to be related to the formation of mixed-material alloys, eutectics, in the fuel cladding. Because of its high melting point, low volatility, and similar chemical behavior to plutonium, the element cerium appears to have behaved similarly to plutonium during the evolution of the TMI-2 accident. Anecdotal evidence extrapolated from open-source literature strengthens this logical feasibility for

  20. Pu-239 and Pu-240 inventories and Pu-240/ Pu-239 atom ratios in the water column off Sanriku, Japan.

    NASA Astrophysics Data System (ADS)

    Yamada, Masatoshi; Zheng, Jian; Aono, Tatsuo

    2013-04-01

    A magnitude 9.0 earthquake and subsequent tsunami occurred in the Pacific Ocean off northern Honshu, Japan, on 11 March 2011 which caused severe damage to the Fukushima Dai-ichi Nuclear Power Plant. This accident has resulted in a substantial release of radioactive materials to the atmosphere and ocean, and has caused extensive contamination of the environment. However, no information is available on the amounts of radionuclides such as Pu isotopes released into the ocean at this time. Investigating the background baseline concentration and atom ratio of Pu isotopes in seawater is important for assessment of the possible contamination in the marine environment. Pu-239 (half-life: 24,100 years), Pu-240 (half-life: 6,560 years) and Pu-241 (half-life: 14.325 years) mainly have been released into the environment as the result of atmospheric nuclear weapons testing. The atom ratio of Pu-240/Pu-239 is a powerful fingerprint to identify the sources of Pu in the ocean. The Pu-239 and Pu-240 inventories and Pu-240/Pu-239 atom ratios in seawater samples collected in the western North Pacific off Sanriku before the accident at Fukushima Dai-ichi Nuclear Power Plant will provide useful background baseline data for understanding the process controlling Pu transport and for distinguishing additional Pu sources. Seawater samples were collected with acoustically triggered quadruple PVC sampling bottles during the KH-98-3 cruise of the R/V Hakuho-Maru. The Pu-240/Pu-239 atom ratios were measured with a double-focusing SF-ICP-MS, which was equipped with a guard electrode to eliminate secondary discharge in the plasma and to enhance overall sensitivity. The Pu-239 and Pu-240 concentrations were 2.07 and 1.67 mBq/m3 in the surface water, respectively, and increased with depth; a subsurface maximum was identified at 750 m depth, and the concentrations decreased with depth, then increased at the bottom layer. The total Pu-239+240 inventory in the entire water column (depth interval 0

  1. 238Pu: accumulation, tissue distribution, and excretion in Mayak workers after exposure to plutonium aerosols.

    PubMed

    Suslova, Klara G; Sokolova, Alexandra B; Khokhryakov, Viktor V; Miller, Scott C

    2012-03-01

    The alpha spectrometry measurements of specific activity of 238Pu and 239Pu in urine from bioassay examinations of 1,013 workers employed at the radiochemical and plutonium production facilities of the Mayak Production Association and in autopsy specimens of lung, liver, and skeleton from 85 former nuclear workers who died between 1974-2009, are summarized.The accumulation fraction of 238Pu in the body and excreta has not changed with time in workers involved in production of weapons-grade plutonium production (e.g., the plutonium production facility and the former radiochemical facility). The accumulation fraction of 238Pu in individuals exposed to plutonium isotopes at the newer Spent Nuclear Fuel Reprocessing Plant ranged from 0.13% up to 27.5% based on the autopsy data. No statistically significant differences between 238Pu and 239Pu in distribution by the main organs of plutonium deposition were found in the Mayak workers. Based on the bioassay data,the fraction of 238Pu activity in urine is on average 38-69% of the total activity of 238Pu and 239Pu, which correlates with the isotopic composition in workplace air sampled at the Spent Nuclear Fuel Reprocessing Plant. In view of the higher specific activity of 238Pu, the contribution of 238Pu to the total internal dose, particularly in the skeleton and liver, might be expected to continue to increase, and continued surveillance is recommended.

  2. Report of an investigation into deterioration of the Plutonium Fuel Form Fabrication Facility (PuFF) at the DOE Savannah River Site

    SciTech Connect

    Not Available

    1991-10-01

    This investigations of the Savannah River Site's Plutonium Fuel Form fabrication facility located in Building 235-F was initiated in April 1991. The purpose of the investigation was to determine whether, as has been alleged, operation of the facility's argon inert gas system was terminated with the knowledge that continued inoperability of the argon system would cause accelerated corrosion damage to the equipment in the plutonium 238 processing cells. The investigation quickly established that the decision to discontinue operation of the argon system, by not repairing it, was merely one of the measures, and not the most important one, which led to the current deteriorated state of the facility. As a result, the scope of the investigation was broadened to more identify and assess those factors which contributed to the facility's current condition. This document discusses the backgrounds, results, and recommendations of this investigation.

  3. Pu(V) transport through Savannah River Site soils - an evaluation of a conceptual model of surface- mediated reduction to Pu (IV).

    PubMed

    Powell, Brian A; Kaplan, Daniel I; Serkiz, Steven M; Coates, John T; Fjeld, Robert A

    2014-05-01

    Over the last fifteen years the Savannah River Site (SRS) in South Carolina, USA, was selected as the site of three new plutonium facilities: the Mixed Oxide Fuel Fabrication Facility, Pit Disassembly and Conversion Facility, and the Pu Immobilization Plant. In order to assess the potential human and environmental risk associated with these recent initiatives, improved understanding of the fate and transport of Pu in the SRS subsurface environment is necessary. The hypothesis of this study was that the more mobile forms of Pu, Pu(V) and Pu(VI), would be reduced to the less mobile Pu(III/IV) oxidation states under ambient SRS subsurface conditions. Laboratory-scale dynamic flow experiments (i.e., column studies) indicated that Pu(V) was very mobile in SRS sediments. At higher pH values the mobility of Pu decreased and the fraction of Pu that became irreversibly sorbed to the sediment increased, albeit, only slightly. Conversely, these column experiments showed that Pu(IV) was essentially immobile and was largely irreversibly sorbed to the sediment. More than 100 batch sorption experiments were also conducted with four end-member sediments, i.e., sediments that include the chemical, textural, and mineralogical properties likely to exist in the SRS. These tests were conducted as a function of initial Pu oxidation state, pH, and contact time and consistently demonstrated that although Pu(V) sorbed initially quite weakly to sediments, it slowly, over the course of <33 days, sorbed very strongly to sediments, to approximately the same degree as Pu(IV). This is consistent with our hypothesis that Pu(V) is reduced to the more strongly sorbing form of Pu, Pu(IV). These studies provide important experimental support for a conceptual geochemical model for dissolved Pu in a highly weathered subsurface environment. That is that, irrespective of the initial oxidation state of the dissolved Pu introduced into a SRS sediment system, Pu(IV) controls the environmental transport

  4. Temporal record of Pu isotopes in inter-tidal sediments from the northeastern Irish Sea.

    PubMed

    Lindahl, Patric; Worsfold, Paul; Keith-Roach, Miranda; Andersen, Morten B; Kershaw, Peter; Leonard, Kins; Choi, Min-Seok; Boust, Dominique; Lesueur, Patrick

    2011-11-01

    A depth profile of (239)Pu and (240)Pu specific activities and isotope ratios was determined in an inter-tidal sediment core from the Esk Estuary in the northeastern Irish Sea. The study site has been impacted with plutonium through routine radionuclide discharges from the Sellafield nuclear reprocessing plant in Cumbria, NW England. A pronounced sub-surface maximum of ~10 k Bq kg(-1) was observed for (239+240)Pu, corresponding to the peak in Pu discharge from Sellafield in 1973, with a decreasing trend with depth down to ~0.04 k Bq kg(-1) in the deeper layers. The depth profile of (239+240)Pu specific activities together with results from gamma-ray spectrometry for (137)Cs and (241)Am was compared with reported releases from the Sellafield plant in order to estimate a reliable sediment chronology. The upper layers (1992 onwards) showed higher (239+240)Pu specific activities than would be expected from the direct input of annual Sellafield discharges, indicating that the main input of Pu is from the time-integrated contaminated mud patch of the northeastern Irish Sea. The (240)Pu/(239)Pu atom ratios ranged from ~0.03 in the deepest layers to >0.20 in the sub-surface layers with an activity-weighted average of 0.181. The decreasing (240)Pu/(239)Pu atom ratio with depth reflects the changing nature of operations at the Sellafield plant from weapons-grade Pu production to reprocessing spent nuclear fuel with higher burn-up times in the late 1950s. In addition, recent annual (240)Pu/(239)Pu atom ratios in winkles collected during 2003-2008 from three stations along the Cumbrian coastline showed no significant spatial or temporal differences with an overall average of 0.204, which supports the hypothesis of diluted Pu input from the contaminated mud patch. Copyright © 2011 Elsevier B.V. All rights reserved.

  5. Biodegradation of PuEDTA and Impacts on Pu Mobility

    SciTech Connect

    Bolton, H., Jr.; Rai, D.; Xun, L.

    2004-03-17

    The contamination of many DOE sites by Pu presents a long-term problem because of its long half-life (240,000 yrs) and the low drinking water standard (<10{sup -12} M). EDTA was co-disposed with radionuclides (e.g., Pu, {sup 60}Co), formed strong complexes, and enhanced radionuclide transport at several DOE sites. Biodegradation of EDTA should decrease Pu mobility. One objective of this project was to determine the biodegradation of EDTA in the presence of PuEDTA complexes. The aqueous system investigated at pH 7 (10{sup -4} M EDTA and 10{sup -6} M Pu) contained predominantly Pu(OH){sub 2}EDTA{sup 2-}. The EDTA was degraded at a faster rate in the presence of Pu. As the total concentration of both EDTA and PuEDTA decreased (i.e., 10{sup -5} M EDTA and 10{sup -7} M PuEDTA), the presence of Pu decreased the biodegradation rate of the EDTA. It is currently unclear why the concentration of Pu directly affects the increase/decrease in rate of EDTA biodegradation. The soluble Pu concentration decreased, in agreement with thermodynamic predictions, as the EDTA was biodegraded, indicating that biodegradation of EDTA will decrease Pu mobility when the Pu is initially present as Pu(IV)EDTA. A second objective was to investigate how the presence of competing metals, commonly encountered in geologic media, will influence the speciation and biodegradation of Pu(IV)-EDTA. Studies on the solubilities of Fe(OH){sub 3}(s) and of Fe(OH){sub 3}(s) plus PuO{sub 2}(am) in the presence of EDTA and as a function of pH showed that Fe(III) out competes the Pu(IV) for the EDTA complex, thereby showing that Pu(IV) will not form stable complexes with EDTA for enhanced transport of Pu in Fe(III) dominated subsurface systems. A third objective is to investigate the genes and enzymes involved in EDTA biodegradation. BNC1 can use EDTA and another synthetic chelating agent nitrilotriacetate (NTA) as sole carbon and nitrogen sources. The same catabolic enzymes are responsible for both EDTA and NTA

  6. The measurement of 240Pu/ 239Pu and 238Pu/ 239Pu isotopic ratios by alpha-particle spectrometry

    NASA Astrophysics Data System (ADS)

    Raab, W.; Parus, J. L.

    1994-01-01

    The measurement of the alpha-activity ratio of {238Pu }/{( 239Pu + 240Pu) } is a routine practice in the determination of the isotopic composition of plutonium. However, measurement of the atomic ratio of 240Pu/ 239Pu by alpha-particle spectrometry is hampered due to insufficient energy resolution for the set of closely spaced peaks of these two isotopes. Passivated and implanted, planar silicon (PIPS) detectors have recently become available with an energy resolution of 10 keV or better, which significantly improves the deconvolution of spectra from plutonium samples. A set of alpha sources was prepared on porcelain disks by ignition, and the spectra were accumulated at a gain of approximately 1 keV per channel. The GRPANL computer program as developed by Lawrence Livermore National Laboratory was used to analyze the spectra. The isotopic ratios were measured in parallel by mass spectrometry. It was found that the agreement on the ratios of 240Pu/ 239Pu and 238Pu/ 239Pu between mass spectrometry and measurements by PIPS detectors was within ±2%. Half-life values were obtained from the literature (M. Lammer and O. Schwerer, Handbook of Nuclear Data for Safeguards, Rep. INDC(NDS)-248, IAEA, Vienna, 1991; ref. [5]). Other factors were also studied to improve the accuracy of the data. The alpha-particle emission probabilities of highly enriched 239Pu and 240Pu have been measured. The alpha-particle energies obtained in the fitting were in agreement with those in ref. [5]. The fitted energy values were used throughout this work.

  7. Designing Pu600 for Authentication

    SciTech Connect

    White, G

    2008-07-10

    Many recent Non-proliferation and Arms Control software projects include an authentication component. Demonstrating assurance that software and hardware performs as expected without hidden 'back-doors' is crucial to a project's success. In this context, 'authentication' is defined as determining that the system performs only its intended purpose and performs that purpose correctly and reliably over many years. Pu600 is a mature software solution for determining the presence of Pu and the ratio of Pu240 to Pu239 by analyzing the gamma ray spectra in the 600 KeV region. The project's goals are to explore hardware and software technologies which can by applied to Pu600 which ease the authentication of a complete, end-to-end solution. We will discuss alternatives and give the current status of our work.

  8. Pu Workshop Letter

    SciTech Connect

    Tobin, J G; Schwartz, A J; Fluss, M

    2006-03-06

    In preparation for the upcoming Pu Workshop in Livermore, CA, USA, during July 14 and 15, 2006, we have begun to give some thought as to how the meeting will be structured and what will be discussed. Below, you will find our first proposal as to the agenda and contents of the meeting. From you, we need your feedback and suggestions concerning the desirability of each aspect of our proposal. Hopefully, we will be able to converge to a format that is acceptable to all parties. First, it now appears that we will be limited to three main sessions, Friday morning (July 14), Friday afternoon (July 14) and Saturday morning (July 15). The Pu Futures Meeting will conclude on Thursday, July 13. Following a social excursion, the Russian participants will be transported from Monterey Bay to their hotel in Livermore. We anticipate that the hotel will be the Residence Inn at 1000 Airway Blvd in Livermore. However, the hotel arrangements still need to be confirmed. We expect that many of our participants will begin their travels homeward in the afternoon of Saturday, July 15 and the morning of Sunday, July 16. Associated with the three main sessions, we propose that there be three main topics. Each session will have an individual focus. Because of the limited time available, we will need to make some judicious choices concerning the focus and the speakers for each session. We will also have a poster session associated with each session, to facilitate discussions, and a rotating set of Lab Tours, to maximize participation in the tour and minimize the disruption of the speaking schedule. Presently, we are planning a tour of the Dynamical Transmission Electron Microscope (DTEM) facilities, but this is still in a preliminary stage. We estimate that for each session and topic, there will be time for five (5) speakers. We propose that, typically, there be three (3) Russian and two (2) American speakers per session. We also propose that each session have a chair (or two chairs), who

  9. Characterization of insoluble residues from the dissolution of irradiated (U,Pu)O{sub 2}

    SciTech Connect

    Goode, J.H.; Arwood, P.C.

    1983-05-01

    Hot-cell tests were conducted using (U,Pu)O{sub 2} fuels that had been irradiated to about 5.2 TJ/kg (U + Pu) [60 MWd/kg (U + Pu)] in an effort to characterize the insoluble residues that remained after the fuel pellets had been dissolved in HNO{sub 3} and in HNO{sub 3}-KF. The composition, particle size range, and density of the material were determined by newer analytical techniques, including spark-source mass spectrometry, neutron activation, scanning electron microscopy, and x-ray fluorescence, combined with older methods such as sedimentation and powder density by water displacement.

  10. Fabrication of a 238Pu target

    SciTech Connect

    Wu, C Y; Chyzh, A; Kwan, E; Henderson, R; Gostic, J; Carter, D

    2010-11-16

    Precision neutron-induced reaction data are important for modeling the network of isotope production and destruction within a given diagnostic chain. This network modeling has many applications such as the design of advanced fuel cycle for reactors and the interpretation of radiochemical data related to the stockpile stewardship and nuclear forensics projects. Our current funded effort is to improve the neutron-induced reaction data on the short-lived actinides and the specific goal is to improve the neutron capture data on {sup 238}Pu with a half-life of 87.7 years. In this report, the fabrication of a {sup 238}Pu target for the proposed measurement using the DANCE array at LANL is described. The {sup 238}Pu target was fabricated from a sample enriched to 99.35%, acquired from ORNL. A total of 395 {micro}g was electroplated onto both sides of a 3 {micro}m thick Ti foil using a custom-made plating cell, shown in Fig 1. The target-material loaded Ti foil is sandwiched between two double-side aluminized mylar foils with a thickness of 1.4 {micro}m. The mylar foil is glued to a polyimide ring. This arrangement is shown partially in Fig. 2. The assembled target is then inserted into an aluminum container with a wall thickness of 0.76 mm, shown in Fig. 3. A derlin ring is used to keep the target assembly in place. The ends of this cylindrical container are vacuum-sealed by two covers with thin Kapton foils as windows for the beam entrance and exit. Shown in Fig. 4 is details of the arrangement. This target is used for phase I of the proposed measurement on {sup 238}Pu scheduled for Nov 2010 together with the DANCE array to address the safety issues raised by LANL. Shown in Fig. 5 is the preliminary results on the yield spectrum as a function of neutron incident energy with a gate on the total {gamma}-ray energy of equivalent Q value. Since no fission PPAC is employed, the distinction between the capture and fission events cannot be made, which is important for the

  11. Pu Anion Exchange Process Intensification

    SciTech Connect

    Taylor-Pashow, K.

    2015-10-08

    This project seeks to improve the efficiency of the plutonium anion-exchange process for purifying Pu through the development of alternate ion-exchange media. The objective of the project in FY15 was to develop and test a porous foam monolith material that could serve as a replacement for the current anion-exchange resin, Reillex® HPQ, used at the Savannah River Site (SRS) for purifying Pu. The new material provides advantages in efficiency over the current resin by the elimination of diffusive mass transport through large granular resin beads. By replacing the large resin beads with a porous foam there is much more efficient contact between the Pu solution and the anion-exchange sites present on the material. Several samples of a polystyrene based foam grafted with poly(4-vinylpyridine) were prepared and the Pu sorption was tested in batch contact tests.

  12. Overview of advanced technologies for stabilization of {sup 238}Pu-contaminated waste

    SciTech Connect

    Ramsey, K.B.; Foltyn, E.M.; Heslop, J.M.

    1998-02-01

    This paper presents an overview of potential technologies for stabilization of {sup 238}Pu-contaminated waste. Los Alamos National Laboratory (LANL) has processed {sup 238}PuO{sub 2} fuel into heat sources for space and terrestrial uses for the past several decades. The 88-year half-life of {sup 238}Pu and thermal power of approximately 0.6 watts/gram make this isotope ideal for missions requiring many years of dependable service in inaccessible locations. However, the same characteristic which makes {sup 238}Pu attractive for heat source applications, the high Curie content (17 Ci/gram versus 0.06 Ci/gram for 239{sup Pu}), makes disposal of {sup 238}Pu-contaminated waste difficult. Specifically, the thermal load limit on drums destined for transport to the Waste Isolation Pilot Plant (WIPP), 0.23 gram per drum for combustible waste, is impossible to meet for nearly all {sup 238}Pu-contaminated glovebox waste. Use of advanced waste treatment technologies including Molten Salt Oxidation (MSO) and aqueous chemical separation will eliminate the combustible matrix from {sup 238}Pu-contaminated waste and recover kilogram quantities of {sup 238}PuO{sub 2} from the waste stream. A conceptual design of these advanced waste treatment technologies will be presented.

  13. Program Pu Futures 2006

    SciTech Connect

    Fluss, M

    2006-06-12

    The coordination chemistry of plutonium remains relatively unexplored. Thus, the fundamental coordination chemistry of plutonium is being studied using simple multi-dentate ligands with the intention that the information gleaned from these studies may be used in the future to develop plutonium-specific sequestering agents. Towards this goal, hard Lewis-base donors are used as model ligands. Maltol, an inexpensive natural product used in the commercial food industry, is an ideal ligand because it is an all-oxygen bidentate donor, has a rigid structure, and is of small enough size to impose little steric strain, allowing the coordination preferences of plutonium to be the deciding geometric factor. Additionally, maltol is the synthetic precursor of 3,4-HOPO, a siderophore-inspired bidentate moiety tested by us previously as a possible sequestering agent for plutonium under acidic conditions. As comparisons to the plutonium structure, Ce(IV) complexes of the same and related ligands were examined as well. Cerium(IV) complexes serve as good models for plutonium(IV) structures because Ce(IV) has the same ionic radius as Pu(IV) (0.94 {angstrom}). Plutonium(IV) maltol crystals were grown out of a methanol/water solution by slow evaporation to afford red crystals that were evaluated at the Advanced Light Source at Lawrence Berkeley National Laboratory using single crystal X-ray diffraction. Cerium(IV) complexes with maltol and bromomaltol were crystallized via slow evaporation of the mother liquor to afford tetragonal, black crystals. All three complexes crystallize in space group I4{sub 1}/a. The Ce(IV) complex is isostructural with the Pu(IV) complex, in which donating oxygens adopt a trigonal dodecahedral geometry around the metal with the maltol rings parallel to the crystallographic S{sub 4} axis and lying in a non-crystallographic mirror plane of D{sub 2d} molecular symmetry (Fig 1). The metal-oxygen bonds in both maltol complexes are equal to within 0.04 {angstrom

  14. In situ high temperature X-Ray diffraction study of the phase equilibria in the UO2-PuO2-Pu2O3 system

    NASA Astrophysics Data System (ADS)

    Belin, Renaud C.; Strach, Michal; Truphémus, Thibaut; Guéneau, Christine; Richaud, Jean-Christophe; Rogez, Jacques

    2015-10-01

    The region of the U-Pu-O phase diagram delimited by the compounds UO2-PuO2-Pu2O3 is known to exhibit a miscibility gap at low temperature. Consequently, MOX fuels with a composition entering this region could decompose into two fluorite phases and thus exhibit chemical heterogeneities. The experimental data on this domain found in the literature are scarce and usually provided using DTA that is not suitable for the investigation of such decomposition phenomena. In the present work, new experimental data, i.e. crystallographic phases, lattice parameters, phase fractions and temperature of phase separation, were measured in the composition range 0.14 < Pu/(U + Pu) < 0.62 and 1.85 < O/(U + Pu) < 2 from 298 to 1750 K using a novel in situ high temperature X-ray diffraction apparatus. A very good agreement is found between the temperature of phase separation determined from our results and using the thermodynamic model of the U-Pu-O system based on the CALPHAD method. Also, the combined use of thermodynamic calculations and XRD results refinement proved helpful in the determination of the O/M ratio of the samples during cooling. The methodology used in the current work might be useful to investigate other oxides systems exhibiting a miscibility gap.

  15. Determination of plutonium isotopes (238Pu, 239Pu, 240Pu, 241Pu) in environmental samples using radiochemical separation combined with radiometric and mass spectrometric measurements.

    PubMed

    Xu, Yihong; Qiao, Jixin; Hou, Xiaolin; Pan, Shaoming; Roos, Per

    2014-02-01

    This paper reports an analytical method for the determination of plutonium isotopes ((238)Pu, (239)Pu, (240)Pu, (241)Pu) in environmental samples using anion exchange chromatography in combination with extraction chromatography for chemical separation of Pu. Both radiometric methods (liquid scintillation counting and alpha spectrometry) and inductively coupled plasma mass spectrometry (ICP-MS) were applied for the measurement of plutonium isotopes. The decontamination factors for uranium were significantly improved up to 7.5 × 10(5) for 20 g soil compared to the level reported in the literature, this is critical for the measurement of plutonium isotopes using mass spectrometric technique. Although the chemical yield of Pu in the entire procedure is about 55%, the analytical results of IAEA soil 6 and IAEA-367 in this work are in a good agreement with the values reported in the literature or reference values, revealing that the developed method for plutonium determination in environmental samples is reliable. The measurement results of (239+240)Pu by alpha spectrometry agreed very well with the sum of (239)Pu and (240)Pu measured by ICP-MS. ICP-MS can not only measure (239)Pu and (240)Pu separately but also (241)Pu. However, it is impossible to measure (238)Pu using ICP-MS in environmental samples even a decontamination factor as high as 10(6) for uranium was obtained by chemical separation.

  16. Hematological responses after inhaling {sup 238}PuO{sub 2}: An extrapolation from beagle dogs to humans

    SciTech Connect

    Scott, B.R.; Muggenburg, B.A.; Welsh, C.A.; Angerstein, D.A.

    1994-11-01

    The alpha emitter plutonium-238 ({sup 238}Pu), which is produced in uranium-fueled, light-water reactors, is used as a thermoelectric power source for space applications. Inhalation of a mixed oxide form of Pu is the most likely mode of exposure of workers and the general public. Occupational exposures to {sup 238}PuO{sub 2} have occurred in association with the fabrication of radioisotope thermoelectric generators. Organs and tissue at risk for deterministic and stochastic effects of {sup 238}Pu-alpha irradiation include the lung, liver, skeleton, and lymphatic tissue. Little has been reported about the effects of inhaled {sup 238}PuO{sub 2} on peripheral blood cell counts in humans. The purpose of this study was to investigate hematological responses after a single inhalation exposure of Beagle dogs to alpha-emitting {sup 238}PuO{sub 2} particles and to extrapolate results to humans.

  17. Characterization of Pu-238 Heat Source Granule Containment

    SciTech Connect

    Richardson, Paul Dean II; Sanchez, Joey Leo; Wall, Angelique Dinorah; Chavarria, Rene

    2015-02-11

    The Milliwatt Radioisotopic Themoelectric Generator (RTG) provides power for permissive-action links. Essentially these are nuclear batteries that convert thermal energy to electrical energy using a doped silicon-germanium thermopile. The thermal energy is provided by a heat source made of 238Pu, in the form of 238PuO2 granules. The granules are contained by 3 layers of encapsulation. A thin T-111 liner surrounds the 238PuO2 granules and protects the second layer (strength member) from exposure to the fuel granules. An outer layer of Hastalloy-C protects the T-111 from oxygen embrittlement. The T-111 strength member is considered the critical component in this 238PuO2 containment system. Any compromise in the strength member seen during destructive testing required by the RTG surveillance program is characterized. The T-111 strength member is characterized through Scanning Electron Microscopy (SEM), and Metallography. SEM is used in the Secondary Electron mode to reveal possible grain boundary deformation and/or cracking in the region of the strength member weld. Deformation and cracking uncovered by SEM are further characterized by Metallography. Metallography sections are mounted and polished, observed using optical microscopy, then documented in the form of microphotographs. SEM mat further be used to examine polished Metallography mounts to characterize elements using the SEM mode of Energy Dispersive X-ray spectroscopy (EDS).

  18. Economical Production of Pu-238

    SciTech Connect

    Steven D. Howe; Douglas Crawford; Jorge Navarro; Terry Ring

    2013-02-01

    All space exploration missions traveling beyond Jupiter must use radioisotopic power sources for electrical power. The best isotope to power these sources is plutonium-238. The US supply of Pu-238 is almost exhausted and will be gone within the next decade. The Department of Energy has initiated a production program with a $10M allocation from NASA but the cost is estimated at over $100 M to get to production levels. The Center for Space Nuclear Research has conceived of a potentially better process to produce Pu-238 earlier and for significantly less cost. The new process will also produce dramatically less waste. Potentially, the front end costs could be provided by private industry such that the government only had to pay for the product produced. Under a NASA Phase I NIAC grant, the CSNR has evaluated the feasibility of using a low power, commercially available nuclear reactor to produce at least 1.5 kg of Pu-238 per year. The impact on the neutronics of the reactor have been assessed, the amount of Neptunium target material estimated, and the production rates calculated. In addition, the size of the post-irradiation processing facility has been established. In addition, a new method for fabricating the Pu-238 product into the form used for power sources has been identified to reduce the cost of the final product. In short, the concept appears to be viable, can produce the amount of Pu-238 needed to support the NASA missions, can be available within a few years, and will cost significantly less than the current DOE program.

  19. Controls on soluble Pu concentrations in PuO2/magnetite suspensions.

    PubMed

    Felmy, Andrew R; Moore, Dean A; Pearce, Carolyn I; Conradson, Steven D; Qafoku, Odeta; Buck, Edgar C; Rosso, Kevin M; Ilton, Eugene S

    2012-11-06

    Time-dependent reduction of PuO(2)(am) was studied over a range of pH values in the presence of aqueous Fe(II) and magnetite (Fe(3)O(4)) nanoparticles. At early time frames (up to 56 days) very little aqueous Pu was mobilized from PuO(2)(am), even though measured pH and redox potentials, coupled to equilibrium thermodynamic modeling, indicated the potential for significant reduction of PuO(2)(am) to relatively soluble Pu(III). Introduction of Eu(III) or Nd(III) to the suspensions as competitive cations to displace possible sorbed Pu(III) resulted in the release of significant concentrations of aqueous Pu. However, the similarity of aqueous Pu concentrations that resulted from the introduction of Eu(III)/Nd(III) to suspensions with and without magnetite indicated that the Pu was solubilized from PuO(2)(am), not from magnetite.

  20. Oxygen diffusion model of the mixed (U,Pu)O2 ± x: Assessment and application

    NASA Astrophysics Data System (ADS)

    Moore, Emily; Guéneau, Christine; Crocombette, Jean-Paul

    2017-03-01

    The uranium-plutonium (U,Pu)O2 ± x mixed oxide (MOX) is used as a nuclear fuel in some light water reactors and considered for future reactor generations. To gain insight into fuel restructuring, which occurs during the fuel lifetime as well as possible accident scenarios understanding of the thermodynamic and kinetic behavior is crucial. A comprehensive evaluation of thermo-kinetic properties is incorporated in a computational CALPHAD type model. The present DICTRA based model describes oxygen diffusion across the whole range of plutonium, uranium and oxygen compositions and temperatures by incorporating vacancy and interstitial migration pathways for oxygen. The self and chemical diffusion coefficients are assessed for the binary UO2 ± x and PuO2 - x systems and the description is extended to the ternary mixed oxide (U,Pu)O2 ± x by extrapolation. A simulation to validate the applicability of this model is considered.

  1. Pu isotopes in the western North Pacific Ocean before the accident at Fukushima Dai-ichi Nuclear Power Station

    NASA Astrophysics Data System (ADS)

    Yamada, M.; Zheng, J.; Aono, T.

    2011-12-01

    Anthropogenic radionuclides such as Pu-239 (half-life: 24100 yr), Pu-240 (half-life: 6560 yr) and Pu-241 (half-life: 14.325 yr) mainly have been released into the environment as the result of atmospheric nuclear weapons testing. In the North Pacific Ocean, two distinct sources of Pu isotopes can be identified; i.e., the global stratospheric fallout and close-in tropospheric fallout from nuclear weapons testing at the Pacific Proving Grounds in the Marshall Islands. The atom ratio of Pu-240/Pu-239 is a powerful fingerprint to identify the sources of Pu in the ocean. The Pu-240/Pu-239 atom ratios in seawater and marine sediment samples collected in the western North Pacific before the accident at Fukushima Dai-ichi Nuclear Power Station will provide useful background data for understanding the process controlling Pu transport and for distinguishing future Pu sources. The atom ratios of Pu-240/Pu-239 in water columns from the Yamato and Tsushima Basins in the Japan Sea were significantly higher than the mean global fallout ratio of 0.18; however, there were no temporal variation of atom ratios during the period from 1984 to 1993 in the Japan Sea. The total Pu-239+240 inventories in the whole water columns were approximately doubled during the period from 1984 to 1993 in the two basins. The atom ratio of Pu-240/Pu-239 in surface water from Sagami Bay, western North Pacific Ocean, was 0.224 and showed no notable variation from the surface to the bottom with the mean atom ratio being 0.234. The atom ratios for the Pacific coast, near the Rokkasho nuclear fuel reprocessing plant, were approximately the same as the 0.224 ratio obtained from Sagami Bay, western North Pacific margin. The atom ratios in the surficial sediments from Sagami Bay ranged from 0.229 to 0.247. The mean atom ratio in the sediment columns in the East China Sea ranged from 0.248 for the Changjiang estuary to 0.268 for the shelf edge. The observed atom ratios were significantly higher than the mean

  2. 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.

  3. Oxidation of Pu-bearing solids: A process for Pu recovery from Rocky Flats incinerator ash

    SciTech Connect

    Karraker, D.G.

    1997-07-18

    High-fired PuO{sub 2}, RFP ash heels, and synthetic RFP incinerator ash were easily soluble after oxidation of Pu(IV) to Pu(VI) by heating with Na{sub 2}O{sub 2} or KO{sub 2} to 450{degrees} for two hours. This offers a route to the recovery of Pu from these and similar PuO{sub 2}-bearing solids that can be carried out in present equipment. Evidence for new compounds K{sub 2}PuO{sub 4}, K{sub 4}PuO{sub 5} and K{sub 6}PuO{sub 6} is presented. A process for recovery of Pu from RFP incinerator ash is presented.

  4. Characterization and property evaluation of U-15 wt%Pu alloy for fast reactor

    NASA Astrophysics Data System (ADS)

    Kaity, Santu; Banerjee, Joydipta; Ravi, K.; Keswani, R.; Kutty, T. R. G.; Kumar, Arun; Prasad, G. J.

    2013-02-01

    The characterization and high temperature behaviour of U-15 wt%Pu alloy has been investigated in this study for the first time. U-15 wt%Pu alloy sample for this study was prepared by following melting and casting route. Microstructural characterization of the alloy was carried out by XRD and optical microscopy. The thermophysical properties like phase transition temperatures, coefficient of thermal expansion and hot hardness of the above alloy were determined. Eutectic temperature between T91 and U-15 wt%Pu was established. Apart from that, the fuel-cladding chemical compatibility of U-15 wt%Pu alloy with T91 grade steel was studied by diffusion couple experiment.

  5. Characterization of Pu-238 heat source granule containment

    SciTech Connect

    Richardson Ii, P D; Thronas, D L; Romero, J P; Sandoval, F E; Neuman, A D; Duncan, W S

    2008-01-01

    The Milliwatt Radioisotopic Thermoelectric Generator (RTG) provides power for permissive-action links. These nuclear batteries convert thermal energy to electrical energy using a doped silicon-germanium thermopile. The thermal energy is provided by a heat source made of {sup 238}Pu, in the form of {sup 238}PuO{sub 2} granules. The granules are contained in 3 layers of encapsulation. A thin T-111 liner surrounds the {sup 238}PuO{sub 2} granules and protects the second layer (strength member) from exposure to the fuel granules. The T-111 strength member contains the fuel under impact condition. An outer clad of Hastelloy-C protects the T-111 from oxygen embrittlement. The T-111 strength member is considered the critical component in this {sup 238}PuO{sub 2} containment system. Any compromise in the strength member is something that needs to be characterized. Consequently, the T-111 strength member is characterized upon it's decommissioning through Scanning Electron Microscopy (SEM), and Metallography. SEM is used in Secondary Electron mode to reveal possible grain boundary deformation and/or cracking in the region of the strength member weld. Deformation and cracking uncovered by SEM are further characterized by Metallography. Metallography sections are mounted and polished, observed using optical microscopy, then documented in the form of photomicrographs. SEM may further be used to examine polished Metallography mounts to characterize elements using the SEM mode of Energy Dispersive X-ray Spectroscopy (EDS). This paper describes the characterization of the metallurgical condition of decommissioned RTG heat sources.

  6. Controls on Soluble Pu Concentrations in PuO2/Magnetite Suspensions

    SciTech Connect

    Felmy, Andrew R.; Moore, Dean A.; Pearce, Carolyn I.; Conradson, Steven D.; Qafoku, Odeta; Buck, Edgar C.; Rosso, Kevin M.; Ilton, Eugene S.

    2012-11-06

    Time-dependent reduction of PuO2(am) was studied over a range of pH values in the presence of aqueous Fe(II) and magnetite (Fe3O4) nanoparticles. At early time frames (up to 56 days) very little aqueous Pu was mobilized from PuO2(am), even though measured pH and redox potentials, coupled to equilibrium thermodynamic modeling indicated the potential for significant reduction of PuO2(am) to relatively soluble Pu(III). Introduction of Eu(III) or Nd(III) to the suspensions as competitive cations to displace possible sorbed Pu(III) resulted in the release of significant concentrations of aqueous Pu. However, the similarity of aqueous Pu concentrations that resulted from the introduction of Eu(III)/Nd(III) to suspensions with and without magnetite indicated that the Pu was displaced from the PuO2(am), not from magnetite. The fact that soluble forms of Pu can be displaced from the surface of PuO2(am) represents a potential, but previously unidentified, source of Pu to aqueous solution or subsurface groundwaters.

  7. Pu, 137Cs and excess 210Pb in Russian Arctic sediments

    NASA Astrophysics Data System (ADS)

    Baskaran, M.; Asbill, Shaunna; Santschi, Peter; Brooks, James; Champ, Michael; Adkinson, Dan; Colmer, Matthew R.; Makeyev, Vyacheslav

    1996-05-01

    The activity ratios of Pu and radiocesium isotopes have been used to delineate the major sources (such as global and close-in (debris) fallout, nuclear fuel reprocessing and fabrication plant effluents) in the environment. We have measured 238Pu, 239,240Pu, 137Cs, and excess 210Pb concentrations in 107 surficial sediments as well as in 5 sediment cores collected in the summer months of 1993 and 1994 from the Ob and Yenisey Rivers (Russia) and the Kara sea. A comparison of the sediment core inventories of 239,240Pu and 137Cs, along with the 238Pu/ 239,240Pu activity ratios, with those expected from global fallout allows us to estimate the relative amounts, if any, of reactor-derived 238Pu and 239,240Pu from the dumped reactor sites in the study area. In surficial sediment samples collected in 1993 and 1994, the 239,240Pu concentrations varied between 4.2 and 856 mBq kg -1, with a mean of 239 mBq kg -1. In samples with a measurable 238Pu, the 238Pu/ 239,240Pu activity ratios varied between 0.010 and 0.069, with an average value of 0.035 ± 0.014. This range can be compared to the average 238Pu/ 239,240Pu activity ratio of 0.030 for the year 1993 from nuclear weapons testing and SNAP fallout obtained from soil studies, indicating very little (≤ 5%) additional sources of 238Pu to the sediments in the study area. The inventories of Pu in the 5 sediment cores from the study area varied between 2.67 ± 0.67 and 24.5 ± 2.2 Bq m -2 with a mean value of 8.83 Bq m -2. The 137Cs concentrations in the upper 3 cm of the sediments varied between below detection limit to 71.4 Bq kg -1, with a mean of 14.9 Bq kg -1. The 137Cs inventories in the 5 sediment cores varied between 156.7 ± 28.3 and 1600 ± 153.3 Bq m -2, with a mean value of 583.3 Bq m -2. The mean ratio of inventories of Pu to that of 137Cs, 0.015, is comparable to the values in other places in the Arctic region. There is a significant correlation between total organic carbon and concentrations of 137Cs, 239,240Pu

  8. Adsorption of Atmospheric Gases on Pu Surfaces

    SciTech Connect

    Nelson, A J; Holliday, K S; Stanford, J A; Grant, W K; Erler, R G; Allen, P G; McLean, W; Roussel, P

    2012-03-29

    Surface adsorption represents a competition between collision and scattering processes that depend on surface energy, surface structure and temperature. The surface reactivity of the actinides can add additional complexity due to radiological dissociation of the gas and electronic structure. Here we elucidate the chemical bonding of gas molecules adsorbed on Pu metal and oxide surfaces. Atmospheric gas reactions were studied at 190 and 300 K using x-ray photoelectron spectroscopy. Evolution of the Pu 4f and O 1s core-level states were studied as a function of gas dose rates to generate a set of Langmuir isotherms. Results show that the initial gas dose forms Pu{sub 2}O{sub 3} on the Pu metal surface followed by the formation of PuO{sub 2} resulting in a layered oxide structure. This work represents the first steps in determining the activation energy for adsorption of various atmospheric gases on Pu.

  9. Pu electronic structure and photoelectron spectroscopy

    SciTech Connect

    Joyce, John J; Durakiewicz, Tomasz; Graham, Kevin S; Bauer, Eric D; Moore, David P; Mitchell, Jeremy N; Kennison, John A; Martin, Richard L; Roy, Lindsay E; Scuseria, G. E.

    2010-01-01

    The electronic structure of PuCoGa{sub 5}, Pu metal, and PuO{sub 2} is explored using photoelectron spectroscopy. Ground state electronic properties are inferred from temperature dependent photoemission near the Fermi energy for Pu metal. Angle-resolved photoemission details the energy vs. crystaJ momentum landscape near the Fermi energy for PuCoGa{sub 5} which shows significant dispersion in the quasiparticle peak near the Fermi energy. For the Mott insulators AnO{sub 2}(An = U, Pu) the photoemission results are compared against hybrid functional calculations and the model prediction of a cross over from ionic to covalent bonding is found to be reasonable.

  10. Recovery of 238PuO2 by Molten Salt Oxidation Processing of 238PuO2 Contaminated Combustibles (Part II)

    NASA Astrophysics Data System (ADS)

    Remerowski, Mary Lynn; Dozhier, C.; Krenek, K.; VanPelt, C. E.; Reimus, M. A.; Spengler, D.; Matonic, J.; Garcia, L.; Rios, E.; Sandoval, F.; Herman, D.; Hart, R.; Ewing, B.; Lovato, M.; Romero, J. P.

    2005-02-01

    Pu-238 heat sources are used to fuel radioisotope thermoelectric generators (RTG) used in space missions. The demand for this fuel is increasing, yet there are currently no domestic sources of this material. Much of the fuel is material reprocessed from other sources. One rich source of Pu-238 residual material is that from contaminated combustible materials, such as cheesecloth, ion exchange resins and plastics. From both waste minimization and production efficiency standpoints, the best solution is to recover this material. One way to accomplish separation of the organic component from these residues is a flameless oxidation process using molten salt as the matrix for the breakdown of the organic to carbon dioxide and water. The plutonium is retained in the salt, and can be recovered by dissolution of the carbonate salt in an aqueous solution, leaving the insoluble oxide behind. Further aqueous scrap recovery processing is used to purify the plutonium oxide. Recovery of the plutonium from contaminated combustibles achieves two important goals. First, it increases the inventory of Pu-238 available for heat source fabrication. Second, it is a significant waste minimization process. Because of its thermal activity (0.567 W per gram), combustibles must be packaged for disposition with much lower amounts of Pu-238 per drum than other waste types. Specifically, cheesecloth residues in the form of pyrolyzed ash (for stabilization) are being stored for eventual recovery of the plutonium.

  11. Anthropogenic Pu distribution in Tropical East Pacific.

    PubMed

    Kinoshita, Norikazu; Sumi, Takahiro; Takimoto, Kiyotaka; Nagaoka, Mika; Yokoyama, Akihiko; Nakanishi, Takashi

    2011-04-15

    The geographical distribution of the anthropogenic radionuclides (238)Pu and (239+240)Pu in the Tropical East Pacific in 2003 was studied from the viewpoint of material migration. We measured the contents of Pu isotopes in seawater and in sediment from the sea bottom. The distributions of Pu isotopes, together with those of coexisting nitrate and phosphate species and dissolved oxygen, are discussed in relation to the potential temperature and potential density (sigma-θ). The Pu contents in sediment samples were compared with those in the seawater. Horizontal migration across the Equator from north to south was investigated at depths down to ~800m in the eastern Pacific. The Pu distribution at 0-400m correlated well with the distribution of potential temperature. Maximum Pu levels were observed in the subsurface layer at 600-800m, corresponding to the depth where sigma-θ≈27.0. It is suggested that the Pu distribution depends on the structure of the water mass and the particular temperature and salinity. The water column/sediment column inventory ratio and the vertical distribution of Pu may reflect the efficiency of scavenging in the relevant water areas. Copyright © 2011 Elsevier B.V. All rights reserved.

  12. Extinct 244Pu in ancient zircons.

    PubMed

    Turner, Grenville; Harrison, T Mark; Holland, Greg; Mojzsis, Stephen J; Gilmour, Jamie

    2004-10-01

    We have found evidence, in the form of fissiogenic xenon isotopes, for in situ decay of 244Pu in individual 4.1- to 4.2-billion-year-old zircons from the Jack Hills region of Western Australia. Because of its short half-life, 82 million years, 244Pu was extinct within 600 million years of Earth's formation. Detrital zircons are the only known relics to have survived from this period, and a study of their Pu geochemistry will allow us to date ancient metamorphic events and determine the terrestrial Pu/U ratio for comparison with the solar ratio.

  13. Design Studies of ``100% Pu'' Mox Lead Test Assembly

    SciTech Connect

    Pavlovichev, A.M.

    2001-01-11

    In this document the results of neutronics studies of <<100%Pu>> MOX LTA design are presented. The parametric studies of infinite MOX-UOX grids, MOX-UOX core fragments and of VVER-1000 core with 3 MOX LTAs are performed. The neutronics parameters of MOX fueled core have been performed for the chosen design MOX LTA using the Russian 3D code BIPR-7A and 2D code PERMAK-A with the constants prepared by the cell spectrum code TVS-M.

  14. New measurement of the 242Pu(n,γ) cross section at n_TOF

    NASA Astrophysics Data System (ADS)

    Lerendegui-Marco, J.; Guerrero, C.; Cortés-Giraldo, M. A.; Quesada, J. M.; Mendoza, E.; Cano-Ott, D.; Eberhardt, K.; Junghans, A.

    2016-03-01

    The use of MOX fuel (mixed-oxide fuel made of UO2 and PuO2) in nuclear reactors allows substituting a large fraction of the enriched Uranium by Plutonium reprocessed from spent fuel. With the use of such new fuel composition rich in Pu, a better knowledge of the capture and fission cross sections of the Pu isotopes becomes very important. In particular, a new series of cross section evaluations have been recently carried out jointly by the European (JEFF) and United States (ENDF) nuclear data agencies. For the case of 242Pu, the two only neutron capture time-of-flight measurements available, from 1973 and 1976, are not consistent with each other, which calls for a new time-of flight capture cross section measurement. In order to contribute to a new evaluation, we have perfomed a neutron capture cross section measurement at the n_TOF-EAR1 facility at CERN using four C6D6 detectors, using a high purity target of 95 mg. The preliminary results assessing the quality and limitations (background, statistics and γ-flash effects) of this new experimental data are presented and discussed, taking into account that the aimed accuracy of the measurement ranges between 7% and 12% depending on the neutron energy region.

  15. Analysis of plutonium isotope ratios including (238)Pu/(239)Pu in individual U-Pu mixed oxide particles by means of a combination of alpha spectrometry and ICP-MS.

    PubMed

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

    2017-04-01

    Isotope ratio analysis of individual uranium-plutonium (U-Pu) mixed oxide particles contained within environmental samples taken from nuclear facilities is proving to be increasingly important in the field of nuclear safeguards. However, isobaric interferences, such as (238)U with (238)Pu and (241)Am with (241)Pu, make it difficult to determine plutonium isotope ratios in mass spectrometric measurements. In the present study, the isotope ratios of (238)Pu/(239)Pu, (240)Pu/(239)Pu, (241)Pu/(239)Pu, and (242)Pu/(239)Pu were measured for individual Pu and U-Pu mixed oxide particles by a combination of alpha spectrometry and inductively coupled plasma mass spectrometry (ICP-MS). As a consequence, we were able to determine the (240)Pu/(239)Pu, (241)Pu/(239)Pu, and (242)Pu/(239)Pu isotope ratios with ICP-MS after particle dissolution and chemical separation of plutonium with UTEVA resins. Furthermore, (238)Pu/(239)Pu isotope ratios were able to be calculated by using both the (238)Pu/((239)Pu+(240)Pu) activity ratios that had been measured through alpha spectrometry and the (240)Pu/(239)Pu isotope ratios determined through ICP-MS. Therefore, the combined use of alpha spectrometry and ICP-MS is useful in determining plutonium isotope ratios, including (238)Pu/(239)Pu, in individual U-Pu mixed oxide particles. Copyright © 2016 Elsevier B.V. All rights reserved.

  16. Use of plutonium isotope activity ratios in dating recent sediments. [/sup 238/Pu//sup 239/Pu + /sup 240/Pu

    SciTech Connect

    Beasley, T. M.

    1982-01-01

    The majority of plutonium presently in the biosphere has come from the testing of nuclear devices. In the early 1950s, the Pu-238/239+240 activity ratio of fallout debris was > 0.04; in the more extensive test series of 1961 to 1962, the Pu-238/239+240 activity ratios were quite consistent at 0.02 to 0.03 and maximum fallout delivery occurred in mid-1963. A significant perturbation in Pu isotope activity ratios occurred in mid-1966 with the deposition of Pu-238 from the SNAP-9A reentry and burn-up. Recently deposited sediments have recorded these events and where accumulation rates are rapid (> 1 cm/y), changes in Pu isotope activity ratios can be used as a geochronological tool.

  17. Separate effects identification via casting process modeling for experimental measurement of U-Pu-Zr alloys

    NASA Astrophysics Data System (ADS)

    Crapps, J.; DeCroix, D. S.; Galloway, J. D.; Korzekwa, D. A.; Aikin, R.; Fielding, R.; Kennedy, R.; Unal, C.

    2013-11-01

    Computational simulations of gravity casting processes for metallic U-Pu-Zr nuclear fuel rods have been performed using a design-of-experiments technique to determine the fluid flow, liquid heat transfer, and solid heat transfer parameters which most strongly influence the process solidification speed and fuel rod porosity. The results are used to make recommendations for the best investment of experimental time and effort to measure process parameters.

  18. Biodegradation of PuEDTA and Impacts on Pu Mobility

    SciTech Connect

    Xun, Luying; Bolton, Jr., Harvey

    2001-06-01

    Ethylenediaminetetraacetate (EDTA) and nitrilotriacetate (NTA) are synthetic chelating agents, which can form strong water-soluble complexes with radionuclides and metals and has been used to decontaminate and process nuclear materials. Synthetic chelating agents were co-disposed with radionuclides (e.g., 60Co, Pu) and heavy metals enhancing their transport in the subsurface. An understanding of EDTA biodegradation is essential to help mitigate enhanced radionuclide transport by EDTA. The objective of this research is to develop fundamental data on factors that govern the biodegradation of radionuclide-EDTA. These factors include the dominant EDTA aqueous species, the biodegradation of various metal-EDTA complexes, the uptake of various metal-EDTA complexes into the cell, the distribution and mobility of the radionuclide during and after EDTA biodegradation, and the enzymology and genetics of EDTA biodegradation.

  19. Determination of the 240Pu/ 239Pu atomic ratio in soils from Palomares (Spain) by low-energy accelerator mass spectrometry

    NASA Astrophysics Data System (ADS)

    Chamizo, E.; García-León, M.; Synal, H.-A.; Suter, M.; Wacker, L.

    2006-08-01

    In 1966, the nuclear fuel of two thermonuclear bombs was released over the Spanish region of Palomares, due to a B52 bomber accident during a refuelling operation. Since then, much effort has been made to assess its impact to the different environmental compartments of this area in South-East Spain, mostly by measuring the 239+240Pu activity concentration and the 238Pu/239+240Pu activity ratio. Nevertheless, these measurements do not give enough information on the problem. In order to recognize unambiguously small traces of the weapon-grade plutonium released in the accident, the ratio of the two major isotopes of plutonium, 240Pu/239Pu, has to be determined. In this work, this ratio has been measured in low- and high-activity samples from Palomares by means of low-energy accelerator mass spectrometry (AMS). That way, we will show the potential of the new generation of compact AMS facilities in terms of plutonium characterization at ultra-trace levels.

  20. Software Aspects of PuMa-II

    NASA Astrophysics Data System (ADS)

    Karuppusamy, R.; Stappers, B.; Stappers, B.

    2006-08-01

    The Pulsar Machine II (PuMa-II) is a state of the art pulsar machine-installed at the Westerbork Synthesis Radio Telescope (WSRT), in December 2005. PuMa-II is a flexible instrument and is designed around an ensemble of 44 high-performance computers running the Linux operating system. Much of the flexibility of PuMa-II comes from the software that is being developed for this instrument. The radio signals reaching the telescope undergo several stages of electronic and software processing before a scientifically useful data product is generated. The electronic processing of signals includes the usual RF to IF conversion, analogue to digital conversion and telescope dependent electronic digital delay compensation that happen in the signal chain of WSRT. Within PuMa-II, this data is acquired, stored and suitably processed. In this poster we present various aspects of PuMa-II software and illustrate its pulsar signal processing capabilities.

  1. Fallout Pu in the Japanese diet

    SciTech Connect

    Hisamatsu, S.; Takizawa, Y.; Abe, T.

    1986-10-01

    The ingestion of fallout Pu from seven or six separate food groups collected in Japan during 1962 and 1983-84 are reported. The contribution of ingested Pu from algae was the highest among the food groups studied: approximately 60% of the total ingested Pu during 1962, and 74% in 1983-84. The contribution from whole marine products, i.e. algae and fish/shellfish, was approximately 70% of the total Pu ingested in 1962, and more than 90% in 1983-84. The higher Japanese ingestion rate than that of the United States is attributable to the higher consumption rate of marine food products in Japan. Also reported in this paper are the ingestion rate for /sup 137/Cs from fallout through the same food groups, and the concentration of Pu and other radioactive nuclides in specific individual foodstuffs and algae samples.

  2. Distribution of nuclear bomb Pu in Nishiyama area, Nagasaki, estimated by accurate and precise determination of 240Pu/239Pu ratio in soils.

    PubMed

    Yoshida, S; Muramatsu, Y; Yamazaki, S; Ban-Nai, T

    2007-01-01

    Plutonium isotopes in forest soils collected in Nishiyama area, Nagasaki, were successfully determined by high resolution inductively coupled plasma mass spectrometry after the treatment with a microwave decomposition system. The (240)Pu/(239)Pu atom ratios observed in the samples in the Nishiyama area were obviously lower than the range of the global fallout. The low ratios (minimum 0.032) observed in Nishiyama area indicated the influence of detonation of the Pu nuclear weapon in 1945. Since the area is contaminated also by global fallout, the (240)Pu/(239)Pu atom ratio can be more sensitive indicator of bomb-derived Pu than Pu activity concentration.

  3. Flowsheet Analysis of U-Pu Co-Crystallization Process as a New Reprocessing System

    SciTech Connect

    Shunji Homma; Jun-ichi Ishii; Jiro Koga; Shiro Matsumoto; Toshiaki Kikuchi; Takahiro Chikazawa; Atsuhiro Shibata

    2006-07-01

    A new fuel reprocessing system by U-Pu co-crystallization process is proposed and examined by flowsheet analysis. This reprocessing system is based on the fact that hexavalent plutonium in nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate does not exist in the solution. The system consists of five steps: dissolution of spent fuel, plutonium oxidation, U-Pu co-crystallization as a co-decontamination, re-dissolution of the crystals, and U re-crystallization as a U-Pu separation. The system requires a recycling of the mother liquor from the U-Pu co-crystallization step and the appropriate recycle ratio is determined by flowsheet analysis such that the satisfactory decontamination is achieved. Further flowsheet study using four different compositions of LWR spent fuels demonstrates that the constant ratio of plutonium to uranium in mother liquor from the re-crystallization step is achieved for every composition by controlling the temperature. It is also demonstrated by comparing to the Purex process that the size of the plant based on the proposed system is significantly reduced. (authors)

  4. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Shackleford, M.H.

    1958-12-16

    A fuel element possessing good stability and heat conducting properties is described. The fuel element comprises an outer tube formed of material selected from the group consisting of stainhess steel, V, Ti. Mo. or Zr, a fuel tube concentrically fitting within the outer tube and containing an oxide of an isotope selected from the group consisting of U/sup 235/, U/sup 233/, and Pu/sup 239/, and a hollow, porous core concentrically fitting within the fuel tube and formed of an oxide of an element selected from the group consisting of Mg, Be, and Zr.

  5. Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Novitrian, Waris, Abdul; Ismail, Suzuki, Mitsutoshi; Saito, Masaki

    2014-09-01

    Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by convertion rasio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissile material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loding scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.

  6. Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

    SciTech Connect

    Permana, Sidik; Novitrian,; Waris, Abdul; Ismail; Suzuki, Mitsutoshi; Saito, Masaki

    2014-09-30

    Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by conversion ratio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissile material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loading scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.

  7. Neutronics Simulations of 237Np Targets to Support Safety-Basis and 238Pu Production Assessment Efforts at the High Flux Isotope Reactor

    SciTech Connect

    Chandler, David; Ellis, Ronald James

    2015-01-01

    Fueled by two highly enriched uranium-bearing fuel elements surrounded by a large concentric ring of beryllium reflector, the High Flux Isotope Reactor (HFIR) provides one of the highest neutron fluxes in the world and is used to produce unique isotopes like plutonium-238. The National Aeronautics and Space Administration use radioisotope thermoelectric generators powered by 238Pu for deep-space missions. As part of the US Department of Energy s task to reestablish the domestic production of 238Pu, a technology demonstration sub-project has been initiated to establish a new 238Pu supply chain. HFIR safety-basis neutronics calculations are being performed to ensure the target irradiations have no adverse impacts on reactor performance and to calculate data required as input to follow-on thermal-structural, thermal-hydraulic and radionuclide/dose analyses. Plutonium-238 production assessments are being performed to estimate the amount of 238Pu that can be produced in HFIR s permanent beryllium reflector. It is estimated that a total of 0.96 1.12 kg 238Pu (~1.28 1.49 kg PuO2 at 85% 238Pu/Pu purity) could be produced per year in HFIR s permanent beryllium reflector irradiation facilities if they are all utilized.

  8. Thermodynamic assessment of the LiF-ThF4-PuF3-UF4 system

    NASA Astrophysics Data System (ADS)

    Capelli, E.; Beneš, O.; Konings, R. J. M.

    2015-07-01

    The LiF-ThF4-PuF3-UF4 system is the reference salt mixture considered for the Molten Salt Fast Reactor (MSFR) concept started with PuF3. In order to obtain the complete thermodynamic description of this quaternary system, two binary systems (ThF4-PuF3 and UF4-PuF3) and two ternary systems (LiF-ThF4-PuF3 and LiF-UF4-PuF3) have been assessed for the first time. The similarities between CeF3/PuF3 and ThF4/UF4 compounds have been taken into account for the presented optimization as well as in the experimental measurements performed, which have confirmed the temperatures predicted by the model. Moreover, the experimental results and the thermodynamic database developed have been used to identify potential compositions for the MSFR fuel and to evaluate the influence of partial substitution of ThF4 by UF4 in the salt.

  9. Resonant Photoemission in f Electron Systems: Pu& Gd

    SciTech Connect

    Tobin, J G; Chung, B W; Schulze, R K; Terry, J; Farr, J D; Shuh, D K; Heinzelman, K; Rotenberg, E; Waddill, G D; van der Laan, G

    2003-03-07

    Resonant photoemission in the Pu5f and Pu6p states is compared to that in the Gd4f and Gd5p states. Spectral simulations, based upon and atomic model with angular momentum coupling, are compared to the Gd and Pu results. Additional spectroscopic measurements of Pu, including core level photoemission and x-ray absorption are also presented.

  10. Resonant photoemission in f electron systems: Pu and Gd

    SciTech Connect

    Tobin, J.G.; Chung, B.W.; Waddill, G.D.; Schulze, R.K.; Terry,J.; Farr, J.D.; Zocco, T.; Shuh, D.K.; Heinzelman, K.; Rotenberg, E.; Vander Laan, G.

    2003-10-14

    Resonant photoemission in the Pu 5f and Pu 6p states is compared to that in the Gd 4f and Gd 5p states. Spectral simulations, based upon an atomic model with angular momentum coupling, are compared to the Gd and Pu results. Additional spectroscopic measurements of Pu, including core level photoemission and x-ray absorption, are also presented.

  11. Resonant photoemission in f-electron systems: Pu and Gd

    NASA Astrophysics Data System (ADS)

    Tobin, J. G.; Chung, B. W.; Schulze, R. K.; Terry, J.; Farr, J. D.; Shuh, D. K.; Heinzelman, K.; Rotenberg, E.; Waddill, G. D.; van der Laan, G.

    2003-10-01

    Resonant photoemission in the Pu 5f and Pu 6p states is compared to that in the Gd 4f and Gd 5p states. Spectral simulations, based upon an atomic model with angular momentum coupling, are compared to the Gd and Pu results. Additional spectroscopic measurements of Pu, including core level photoemission and x-ray absorption, are also presented.

  12. PuMa, a digital Pulsar Machine

    NASA Astrophysics Data System (ADS)

    Voûte, J. L. L.; Kouwenhoven, M. L. A.; van Haren, P. C.; Langerak, J. J.; Stappers, B. W.; Driesens, D.; Ramachandran, R.; Beijaard, Th. D.

    2002-04-01

    We have designed and constructed PuMa, a pulsar machine that has both a baseband recording and a digital filterbank mode. Its design is based on the use of digital signal processors (DSPs). Their operation is controlled by software, which makes PuMa reconfigurable, flexible and easy to operate. The maximum number of channels in the digital filterbank mode is 32 768 over a bandwidth of 80 MHz. This makes PuMa suitable for pulsar observations at low sky frequencies. The maximum bandwidth in baseband recording mode is two times 10 MHz. The machine was installed at the Westerbork Synthesis Radio Telescope in The Netherlands in 1998. We discuss in some detail PuMa's technical properties and capabilities. Recent observations, a sample of which are presented here, demonstrate its capabilities and that it is performing up to its specifications.

  13. Development of prototype induced-fission-based Pu accountancy instrument for safeguards applications.

    PubMed

    Seo, Hee; Lee, Seung Kyu; An, Su Jung; Park, Se-Hwan; Ku, Jeong-Hoe; Menlove, Howard O; Rael, Carlos D; LaFleur, Adrienne M; Browne, Michael C

    2016-09-01

    Prototype safeguards instrument for nuclear material accountancy (NMA) of uranium/transuranic (U/TRU) products that could be produced in a future advanced PWR fuel processing facility has been developed and characterized. This is a new, hybrid neutron measurement system based on fast neutron energy multiplication (FNEM) and passive neutron albedo reactivity (PNAR) methods. The FNEM method is sensitive to the induced fission rate by fast neutrons, while the PNAR method is sensitive to the induced fission rate by thermal neutrons in the sample to be measured. The induced fission rate is proportional to the total amount of fissile material, especially plutonium (Pu), in the U/TRU product; hence, the Pu amount can be calibrated as a function of the induced fission rate, which can be measured using either the FNEM or PNAR method. In the present study, the prototype system was built using six (3)He tubes, and its performance was evaluated for various detector parameters including high-voltage (HV) plateau, efficiency profiles, dead time, and stability. The system's capability to measure the difference in the average neutron energy for the FNEM signature also was evaluated, using AmLi, PuBe, (252)Cf, as well as four Pu-oxide sources each with a different impurity (Al, F, Mg, and B) and producing (α,n) neutrons with different average energies. Future work will measure the hybrid signature (i.e., FNEM×PNAR) for a Pu source with an external interrogating neutron source after enlarging the cavity size of the prototype system to accommodate a large-size Pu source (~600g Pu).

  14. A Neutronic Analysis of TRU Recycling in PWRs Loaded with MOX-UE Fuel (MOX with U-235 Enriched U Support)

    SciTech Connect

    G. Youinou; S. Bays

    2009-05-01

    This report presents the results of a study dealing with the homogeneous recycling of either Pu or Pu+Np or Pu+Np+Am or Pu+Np+Am+Cm in PWRs using MOX-UE fuel, i.e. standard MOX fuel with a U235 enriched uranium support instead of the standard tail uranium (0.25%) for standard MOX fuel. This approach allows to multirecycle Pu or TRU (Pu+MA) as long as U235 is available, by keeping the Pu or TRU content in the fuel constant and at a value ensuring a negative moderator void coefficient (i.e. the loss of the coolant brings imperatively the reactor to a subcritical state). Once this value is determined, the U235 enrichment of the MOX-UE fuel is adjusted in order to reach the target burnup (51 GWd/t in this study).

  15. Synthesis of Pu-Doped Ceramic

    SciTech Connect

    Anderson, E. B

    1998-09-02

    Plutonium-doped zircon containing about 10 wt% Pu was synthesized in this cooperative project between Russia and the United States conducted at the V. G. Khlopin Radium Institute. The sol-gel method was used for starting precursor preparation to provide complete mixing of initial components and to avoid dust formation inside the glove-box. The sol-gel process also gives interim Pu stabilization in the form of amorphous zirconium hydrosilicate (AZHS), which is a result of gel solidification. AZHS is a solid and relatively durable material that can be easy converted into crystalline zircon by pressureless sintering, thus avoiding significant radioactive contamination of laboratory equipment. A methanol-aqueous solution of tetraethoxysilane Si(OC2H5)4, Pu-nitrate, and zirconil oxynitrate was prepared in final stoichiometry of zircon (Zr,Pu)SiO4 80 wt% + zirconia (Zr,Pu)O2 20 wt%. Gelation occurred after 90 hours at room temperature. AZHS with excess of zirconia 20 wt% was obtained as an interim calcine product and then it was converted into zircon/zirconia ceramic by sintering at 1490 to 1500°C in air for different time periods. The samples obtained were studied by SRD and ESEM methods. It was found that both zircon yield and zircon cell parameters that are correlated with Pu incorporation depend on sintering time.

  16. The Spectroscopic Signature of Aging in (delta)-Pu(Ga)

    SciTech Connect

    Tobin, J G; Yu, S; Chung, B W

    2009-04-29

    The electronic structure of Pu is briefly discussed, with emphasis upon Aging effects. Photoelectron Spectroscopy and X-ray Absorption Spectroscopy have contributed greatly to our improved understanding of Pu electronic structure. (See Figure 1.) From these and related measurements, the following has been determined: (1) The Pu 5f spin-orbit splitting is large; (2) The number of Pu5f electrons is 5; and (3) The Pu 5f spin-orbit splitting effect dominates 5f itineracy.

  17. Investigation of Benefits from U/TRU Recycle - Quantification and Comparison to U/Pu Recycle

    SciTech Connect

    R. Wigeland; T. Taiwo; M. Todosow

    2015-09-01

    The recently completed comprehensive evaluation and screening of nuclear fuel cycle options identified a number of potentially promising fuel cycles for R&D that offer what could be considered by decision-makers as having the potential for significant improvement compared to the current U.S. fuel cycle. The fuel cycles that consistently performed the best were recycle fuel cycles that used self-sustaining fast reactors operating with either U/Pu or U/TRU recycle fuel and also included options where the fast reactors provided fissile materials to support operation of thermal reactors. However, based on the evaluation criteria and metrics used in the study, there was no difference in benefit between recycle of U/Pu and U/TRU (where TRU is plutonium and the minor actinides) while there were differences in the challenges for developing and deploying such fuel cycles, with U/TRU recycle being more challenging. This observation prompted the question as to the desirability of pursuing R&D on U/TRU recycle given that there may not be an increase in benefit. As a result, activities have been pursued to further investigate the performance differences between U/Pu and U/TRU recycle based on considering issues beyond those used in the evaluation and screening study to identify, if possible, areas where there are significant benefits of U/TRU recycle compared to U/Pu recycle. These new considerations focused on several areas, but especially on the impact on disposal of the HLW, which in the case of U/Pu recycle contains all of the minor actinides along with fission products, while in the case of U/TRU recycle only contains the losses of minor actinides from the reprocessing and recycle fuel fabrication operations. This difference in content has several implications. One impact is on the time dependent decay heat which can affect handling and the use of space in a geologic repository. Another impact concerns the HLW form and volume, since presence of minor actinides may

  18. Prevention of Pu(IV) polymerization in a PUREX-based process

    SciTech Connect

    Paviet-Hartmann, Patricia; Senentz, Gerald

    2007-07-01

    The US Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF) is being designed to produce MOX fuel assemblies for use in domestic, commercial nuclear power reactors, as part of the U.S. DOE efforts to dispose of surplus weapon-grade plutonium. The feed material is plutonium dioxide from surplus weapon grade plutonium. PuO{sub 2}, issued from a pit disassembly and conversion facility (PDCF), will be processed using a flowsheet derived from the La Hague reprocessing plant to remove impurities. The purified PuO{sub 2} will be blended with UO{sub 2} to form mixed oxide pellets, and loaded into fuel rods, to create MOX fuel assemblies based on the process and technology of the MELOX plant in France,. Safety studies are necessary to support the development of the design basis per regulation 10 CFR Part 70 to complete an integrated safety analysis for the MFFF facility. The formation of tetravalent plutonium polymers in certain process vessels of the aqueous polishing (AP) process has been identified as a potential hazard. Based on scientific literature, the following paper demonstrates that within the AP process units, the polymerization of Pu(IV) will not occur and/or will not create a criticality issue even where the acidity may drop below 0.5 N HNO{sub 3}. We will identify and control the conditions under which plutonium (IV) will not polymerize. (authors)

  19. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    SciTech Connect

    Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

    1999-03-23

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

  20. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    SciTech Connect

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

    1997-12-01

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr-Hf alloy or an alloy of Pu-Zr-Hf or a combination of both.

  1. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    DOEpatents

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

    1999-01-01

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both.

  2. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    DOEpatents

    Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

    1999-03-23

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

  3. An analysis of plutonium immobilization versus the "spent fuel" standard

    SciTech Connect

    Gray, W L; McKibben, J M

    1998-06-16

    Safe Pu management is an important and urgent task with profound environmental, national, and international security implications. Presidential Policy Directive 13 and analyses by scientific, technical, and international policy organizations brought about a focused effort within the Department of Energy (DOE) to identify and implement long-term disposition paths for surplus Pu. The principal goal is to render surplus Pu as inaccessible and unattractive for reuse in nuclear weapons as Pu in spent reactor fuel. In the Programmatic Environmental Impact Statement and Record of Decision for the Storage and Disposition of Weapons- Usable Fissile Materials (1997), DOE announced pursuit of two disposition technologies: (1) irradiation of Pu as MOX fuel in existing reactors and (2) immobilization of Pu into solid forms containing fission products as a radiation barrier. DOE chose an immobilization approach that includes "use of the can-in-canister option.. . for a portion of the surplus, non-pit Pu material." In the can-in-canister approach, cans of glass or ceramic forms containing Pu are encapsulated within canisters of HLW glass. In support of the selection process, a technical evaluation of retrievability and recoverability of Pu from glass and ceramic forms by a host nation and by rogue nations or subnational groups was completed. The evaluation involved determining processes and flowsheets for Pu recovery, comparing these processes against criteria and metrics established by the Fissile Materials Disposition Program and then comparing the recovery processes against each other and against SNF processes.

  4. Chemical potential of oxygen in (U, Pu) mixed oxide with Pu/(U+Pu) = 0.46

    NASA Astrophysics Data System (ADS)

    Dawar, Rimpi; Chandramouli, V.; Anthonysamy, S.

    2016-05-01

    Chemical potential of oxygen in (U,Pu) mixed oxide with Pu/(U + Pu) = 0.46 was measured for the first time using H2/H2O gas equilibration combined with solid electrolyte EMF technique at 1073, 1273 and 1473 K covering an oxygen potential range of -525 to -325 kJ mol-1. The effect of oxygen potential on the oxygen to metal ratio was determined. Increase in oxygen potential increases the O/M. In this study the minimum O/M obtained was 1.985 below which reduction was not possible. Partial molar enthalpy ΔHbar O2 and entropy ΔSbar O2 of oxygen were calculated from the oxygen potential data. The values of -752.36 kJ mol-1 and 0.25 kJ mol-1 were obtained for ΔHbar O2 and ΔSbar O2 respectively.

  5. MCSNA: Experimental Benchmarking of Pu Electronic Structure

    SciTech Connect

    Tobin, J G

    2007-01-29

    The objective of this work is to develop and/or apply advanced diagnostics to the understanding of aging of Pu. Advanced characterization techniques such as photoelectron and x-ray absorption spectroscopy will provide fundamental data on the electronic structure of Pu phases. These data are crucial for the validation of the electronic structure methods. The fundamental goal of this project is to narrow the parameter space for the theoretical modeling of Pu aging. The short-term goal is to perform experiments to validate electronic structure calculations of Pu. The long-term goal is to determine the effects of aging upon the electronic structure of Pu. Many of the input parameters for aging models are not directly measurable. These parameters will need to be calculated or estimated. Thus a First Principles-Approach Theory is needed, but it is unclear what terms are important in the Hamiltonian. (H{Psi} = E{Psi}) Therefore, experimental data concerning the 5f electronic structure are needed, to determine which terms in the Hamiltonian are important. The data obtained in this task are crucial for reducing the uncertainty of Task LL-01-developed models and predictions. The data impact the validation of electronic structure methods, the calculation of defect properties, the evaluation of helium diffusion, and the validation of void nucleation models. The importance of these activities increases if difficulties develop with the accelerating aging alloy approach.

  6. Microstructural Characterization of Cast Metallic Transmutation Fuels

    SciTech Connect

    J. I. Cole; D. D. Keiser; J. R. Kennedy

    2007-09-01

    As part of the Global Nuclear Energy Partnership (GNEP) and the Advanced Fuel Cycle Initiative (AFCI), the US Department of Energy (DOE) is participating in an international collaboration to irradiate prototypic actinide-bearing transmutation fuels in the French Phenix fast reactor (FUTURIX-FTA experiment). The INL has contributed to this experiment by fabricating and characterizing two compositions of metallic fuel; a non-fertile 48Pu-12Am-40Zr fuel and a low-fertile 35U-29Pu-4Am-2Np-30Zr fuel for insertion into the reactor. This paper highlights results of the microstructural analysis of these cast fuels, which were reasonably homogeneous in nature, but had several distinct phase constituents. Spatial variations in composition appeared to be more pronounced in the low-fertile fuel when compared to the non-fertile fuel.

  7. Closed DTU fuel cycle with Np recycle and waste transmutation

    SciTech Connect

    Beller, D.E.; Sailor, W.C.; Venneri, F.; Herring, J.S.

    1999-09-01

    A nuclear energy scenario for the 21st century that included a denatured thorium-uranium-oxide (DTU) fuel cycle and new light water reactors (LWRs) supported by accelerator-driven transmutation of waste (ATW) systems was previously described. This coupled system with the closed DTU fuel cycle provides several improvements beyond conventional LWR (CLWR) (once-through, UO{sub 2} fuel) nuclear technology: increased proliferation resistance, reduced waste, and efficient use of natural resources. However, like CLWR fuel cycles, the spent fuel in the first one-third core discharged after startup contains higher-quality Pu than the equilibrium fuel cycle. To eliminate this high-grade Pu, Np is separated and recycled with Th and U--rather than with higher actinides [(HA) including Pu]. The presence of Np in the LWR feed greatly increases the production of {sup 238}Pu so that a few kilograms of Pu generated enough alpha-decay heat that the separated Pu is highly resistant to proliferation. This alternate process also simplifies the pyrochemical separation of fuel elements (Th and U) from HAs. To examine the advantages of this concept, the authors modeled a US deployment scenario for nuclear energy that includes DTU-LWRs plus ATW`s to burn the actinides produced by these LWRs and to close the back-end of the DTU fuel cycle.

  8. The basic features of a closed fuel cycle without fast reactors

    NASA Astrophysics Data System (ADS)

    Bobrov, E. A.; Alekseev, P. N.; Teplov, P. S.

    2017-01-01

    In this paper the basic features of a closed fuel cycle with thermal reactors are considered. The three variants of multiple Pu and U recycling in VVER reactors was investigated. The comparison of MOX and REMIX fuel approaches for closed fuel cycle with thermal reactors is presented. All variants make possible to recycle several times the total amount of Pu and U obtained from spent fuel. The reported study was funded by RFBR according to the research project № 16-38-00021

  9. PU Disposition in Russian VVERs: Physics Studies of Lead Test Assembly Design

    SciTech Connect

    Ellis, R.J.

    2000-05-07

    As part of Fissile Materials Disposition Program (FMDP) physics support was given to the design of a MOX lead test assembly (LTA) for use in Russian VVER nuclear reactors. This paper discusses some of the pertinent findings and assessments for two distinct LTA designs for weapons-grade (WG) Pu dispositioning in Russian VVER-1000 nuclear reactors. The two assessed MOX LTA designs are the graded-zone full MOX LTA and the Island LTA (2 central zones of MOX pins surrounded by UO{sub 2} pins). The process of optimizing the graded Pu-content by zone in the fuel assembly is discussed. Eigenvalue and power peaking comparisons are made as a function of fuel burnup. Zero-power reactivity effects were calculated for the different LTA options. For the ORNL results, the n,{gamma}-transport lattice physics code HELIOS-1.4 was used with nuclear data libraries (based on ENDF/B-VI) in 89 and 190 neutron energy groups. Some comparisons are made between the ORNL HELIOS results and corresponding Russian LTA calculations by the RRC-KI (Kurchatov Institute) using the code TVS-M. Also in this paper, pertinent results are discussed from a study of void reactivity effects for LEU, RG MOX and WG MOX fuels in PWR and VVER-1000 nuclear reactors. These void reactivity calculations were performed for a large range of LEU enrichments (2-20 wt% {sup 235}U), and large ranges of Pu-content (2-20 wt% Pu) in RG and WG MOX fuel.

  10. Localized 5f electrons in superconducting PuCoIn₅: consequences for superconductivity in PuCoGa₅.

    PubMed

    Bauer, E D; Altarawneh, M M; Tobash, P H; Gofryk, K; Ayala-Valenzuela, O E; Mitchell, J N; McDonald, R D; Mielke, C H; Ronning, F; Griveau, J-C; Colineau, E; Eloirdi, R; Caciuffo, R; Scott, B L; Janka, O; Kauzlarich, S M; Thompson, J D

    2012-02-08

    The physical properties of the first In analog of the PuMGa(5) (M = Co, Rh) family of superconductors, PuCoIn(5), are reported. With its unit cell volume being 28% larger than that of PuCoGa(5), the characteristic spin-fluctuation energy scale of PuCoIn(5) is three to four times smaller than that of PuCoGa(5), which suggests that the Pu 5f electrons are in a more localized state relative to PuCoGa(5). This raises the possibility that the high superconducting transition temperature T(c) = 18.5 K of PuCoGa(5) stems from the proximity to a valence instability, while the superconductivity at T(c) = 2.5 K of PuCoIn(5) is mediated by antiferromagnetic spin fluctuations associated with a quantum critical point.

  11. The geochemistry of fallout plutonium in the North Atlantic: II. 240Pu /239Pu ratios and their significance

    NASA Astrophysics Data System (ADS)

    Buesseler, Ken O.; Sholkovitz, Edward R.

    1987-10-01

    A systematic decrease in the 240Pu /239Pu ratio in marine sediments is found with increasing water depth along a transect of cores between Woods Hole and Bermuda. The 240Pu /239Pu atom ratios range from ≅O.18 on the shelf to ≅O.10 at 5000 m but do not change with depth in individual cores. A model is presented which can account for the range of 240Pu /239Pu ratios found in this and other similar studies ( NOSHKIN and GATROUSIS, 1974; SCOTTet al., 1983). We propose that there have been at least two distinct sources of fallout Pu to this region. The major source of Pu is global stratospheric fallout, characterized by a 240Pu /239Pu ratio of 0.18 and a relatively long residence time in seawater. The second source is characterized by a much lower 240Pu /239Pu ratio, and relative to global fallout it must have been much more efficiently removed from the water column to deep-sea sediments. We suggest that surface-based low yield testing at the Nevada Test Site is the only source of low ratio fallout Pu which could account for the timing, inventories, and refractory characteristics of this second component of fallout Pu inputs to the North Atlantic.

  12. Progress of nitride fuel cycle research for transmutation of minor actinides

    SciTech Connect

    Arai, Yasuo; Akabori, Mitsuo; Minato, Kazuo

    2007-07-01

    Recent progress of nitride fuel cycle research for transmutation of MA is summarized. Preparation of MA-bearing nitride pellets, such as (Np,Am)N, (Am,Pu)N and (Np,Pu,Am,Cm)N, was carried out. Irradiation behavior of U-free nitride fuel was investigated by the irradiation test of (Pu,Zr)N and PuN+TiN fuels, in which ZrN and TiN were added as a possible diluent material. Further, pyrochemical process of spent nitride fuel was developed by electrorefining in a molten chloride salt and subsequent re-nitridation of actinides in liquid Cd cathode electro-deposits. Nitride fuel cycle for transmutation of MA has been demonstrated in a laboratory scale by the experimental study with MA and Pu. (authors)

  13. Isochronal Annealing Studies in Pu and Pu Alloys Using Magnetic Susceptibility

    SciTech Connect

    McCall, S. K.; Fluss, M. J.; Chung, B. W.; McElfresh, M. W.; Chapline, G.F.; Jackson, D. D.; Haire, Richard {Dick} G

    2007-01-01

    The isochronal annealing of the low temperature accumulated damage from the radioactive decay of plutonium in {alpha}-Pu, {delta}-Pu{sub 1-x}Ga{sub x} (x = 0.043) and {delta}-Pu{sub 1-x}Am{sub x} (x = 0.224) was characterized using magnetic susceptibility. In each specimen, thermal annealing, as tracked by magnetic susceptibility, only commenced when T > 33 K and the magnetic susceptibility changes due to defects were fully annealed at T not, vert, similar 300 K. The {alpha}-Pu magnetic susceptibility isochronal annealing data is similar to earlier measurements of resistivity characterized isochronal annealing. However, the {delta}-Pu{sub 1-x}Ga{sub x} (x = 0.043) magnetic susceptibility isochronal annealing data, when compared with similar resistivity data, indicates that for this alloy magnetic susceptibility studies are more sensitive to vacancies than to the interstitials accumulated at low temperatures. The Pu{sub 1-x}Am{sub x} (x = 0.224) alloy shows a remarkable change in properties, over a limited temperature range beginning where interstitial defects are first mobile, and characterized by an induced effective moment of order 1.1 {mu}{sub B}/Pu. This transient behavior may be evidence for a disorder driven low temperature phase transition, perhaps indicative of a compositional and structural proximity to a state possessing significant magnetic moments.

  14. Effect of polyurethane (PU) - bioactive glass (BG) ratio on the development of BG reinforced PU scaffold

    NASA Astrophysics Data System (ADS)

    Lip, Lim Weng; Abdullah, Tuti Katrina; Zubir, Syazana Ahmad

    2016-12-01

    Nowadays, variety of biomaterials may be used to produce implanted scaffolds such as metal-based, ceramic-based and polymer-based materials. In this study, porous bioactive glass (BG) reinforced polyurethane (PU) composite scaffolds with different PU:BG mass ratio (10 to 40 wt%) were fabricated as a potential candidate for synthetic bone graft. The PU-BG scaffolds were prepared using solvent casting combined with salt leaching (SCPL) method and were subjected to several characterizations including fourier transform infra-red (FTIR) spectroscopy, scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDX). FTIR spectrum showed the trace of BG particles in the PU-BG scaffolds with high concentration of BG (30 and 40 wt%). EDX confirmed that the white particles in the PU-BG scaffold as observed via SEM micrograph were BG particles. A slightly round and irregular pore structures were observed for the PU-BG scaffolds prepared in this study. More homogeneous pore structures were observed as the amount of BG in the PU-BG scaffold is increased. The overall pore size for all scaffolds was in the range of 130 to 400 µm which is suitable for the growth of bone tissue.

  15. Oxidation behaviour of plutonium rich (U, Pu)C and (U, Pu)O2

    NASA Astrophysics Data System (ADS)

    Sali, S. K.; Kulkarni, N. K.; Phatak, Rohan; Agarwal, Renu

    2016-10-01

    Oxidation behaviour of (U0.3Pu0.7)C1.06 was investigated in air by heating samples up to 1073 K and 1273 K. Thermogravimetry (TG) of the samples and X-ray powder diffraction (XRD) of the intermediate products were used to understand the phenomenon taking place during this process. Theoretical calculations were carried out to understand the multiple phase changes taking place during oxidation of carbide. Theoretical results were validated by XRD analysis of the products obtained at different stages of oxidation. The final oxidized products were found to be a single FCC phase with O/M = 2.15 (M = U + Pu). Oxidation kinetic studies of (U0.3Pu0.7)O2 and (U0.47Pu0.53)O2 were carried out in dry air, using thermogravimetry, under non-isothermal conditions. The activation energy of oxidation was found to be 49 and 70 kJ/mol, respectively. Lattice parameter dependence on Pu/M and O/M of plutonium rich mixed oxide (MOX) was established using combined results of XRD and TG analysis of (U0.3Pu0.7)O2+x and (U0.47Pu0.53)O2+x.

  16. Polymeric Species of Pu in Low Ionic Strength Media

    SciTech Connect

    Romanovski, V V; Palmer, C E; Shaw, H F; Bourcier, W L; Jardine, L J

    2000-01-27

    The US Government has declared that approximately 50 tons of plutonium is surplus to US needs and should be set aside for eventual disposition. The US is currently following a dual path for the disposition of this plutonium: immobilization and irradiation of mixed-oxide fuel. Some fraction of this plutonium material that is undesirable for use in mixed-oxide fuel will be immobilized in a titanate ceramic and disposed of in a geologic repository for high level waste. The reminder of Pu will be fabricated into mixed-oxide fuel and irradiated in domestic light-water reactors. The resulting spent fuel would also be disposed of in a geologic repository for high level waste. The proposed US repository would be at the Yucca Mountain site in Nevada. Plutonium present in the disposal forms, either ceramics or spent fuel, must remain isolated from the biosphere over the geologic repository regulatory performance period, which is currently 10,000 years. Contamination of the biosphere could result from slow dissolution of the disposal forms followed by transport of the dissolution products into the biosphere by flowing ground water. Measurable amounts of apparently soluble plutonium can be released if plutonium dioxide is exposed to water under some conditions. Furthermore, recent studies in Nevada near the Yucca Mountain Site revealed that plutonium, associated with the colloidal fraction of the groundwater, was detected over a kilometer from its source within 30 years after it was placed underground for a nuclear weapons testing. It was not clear whether plutonium was transported as an intrinsic plutonium colloid or as plutonium sorbed onto colloidal clay or zeolite particles.

  17. [Acupuncture messenger--Pu Xiang-cheng].

    PubMed

    Du, Huai-bin; Liang, Fan-rong

    2011-06-01

    PU Xiang-cheng is the eminent acupuncture master in modern history of China. He studied diligently in early years and devoted his life to the cause of acupuncture practice and education in Chinese medicine. Combination of acupuncture and herbal medicine, coordination of acupuncture and moxibustion, unique application of acupoints, flexible combination of acupoints and focusing on needling techniques are the essence of his academic thoughts. The life of PU Xiang-cheng, the acupuncture master, and his major academic thoughts are described in this paper, so as to commemorate his contributions to acupuncture theory, practice and promotions.

  18. Neutron Capture Cross Section of 239Pu

    NASA Astrophysics Data System (ADS)

    Mosby, S.; Arnold, C.; Bredeweg, T. A.; Couture, A.; Jandel, M.; O'Donnell, J. M.; Rusev, G.; Ullmann, J. L.; Chyzh, A.; Henderson, R.; Kwan, E.; Wu, C. Y.

    2014-09-01

    The 239Pu(n,γ) cross section has been measured over the energy range 10 eV - 10 keV using the Detector for Advanced Neutron Capture Experiments (DANCE) as part of a campaign to produce precision (n,γ) measurements on 239Pu in the keV region. Fission coincidences were measured with a PPAC and used to characterize the prompt fission γ-ray spectrum in this region. The resulting spectra will be used to better characterize the fission component of another experiment with a thicker target to extend the (n,γ) cross section measurement well into the keV region.

  19. A practical strategy for reducing the future security risk of United States spent nuclear fuel

    SciTech Connect

    Chodak, P. III; Buksa, J.J.

    1997-06-01

    Depletion calculations show that advanced oxide (AOX) fuels can be used in existing light water reactors (LWRs) to achieve and maintain virtually any desired level of US (US) reactor-grade plutonium (R-Pu) inventory. AOX fuels are composed of a neutronically inert matrix loaded with R-Pu and erbium. A 1/2 core load of 100% nonfertile, 7w% R-Pu AOX and 3.9 w% UO{sub 2} has a net total plutonium ({sup TOT}Pu) destruction rate of 310 kg/yr. The 20% residual {sup TOT}Pu in discharged AOX contains > 55% {sup 242}Pu making it unattractive for nuclear explosive use. A three-phase fuel-cycle development program sequentially loading 60 LWRs with 100% mixed oxide, 50% AOX with a nonfertile component displacing only some of the {sup 238}U, and 50% AOX, which is 100% nonfertile, could reduce the US plutonium inventory to near zero by 2050.

  20. Metallic fuel development

    SciTech Connect

    Walters, L.C.

    1987-01-01

    Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising.

  1. Transport of 137Cs and 239,240Pu with ice-rafted debris in the Arctic Ocean

    USGS Publications Warehouse

    Landa, E.R.; Reimnitz, E.; Beals, D.M.; Pochkowski, J.M.; Winn, W.G.; Rigor, I.

    1998-01-01

    Ice rafting is the dominant mechanism responsible for the transport of fine-grained sediments from coastal zones to the deep Arctic Basin. Therefore, the drift of ice-rafted debris (IRD) could be a significant transport mechanism from the shelf to the deep basin for radionuclides originating from nuclear fuel cycle activities and released to coastal Arctic regions of the former Soviet Union. In this study, 28 samples of IRD collected from the Arctic ice pack during expeditions in 1989-95 were analyzed for 137Cs by gamma spectrometry and for 239Pu and 240Pu by thermal ionization mass spectrometry. 137Cs concentrations in the IRD ranged from less than 0.2 to 78 Bq??kg-1 (dry weight basis). The two samples with the highest 137Cs concentrations were collected in the vicinity of Franz Josef Land, and their backward trajectories suggest origins in the Kara Sea. Among the lowest 137Cs values are seven measured on sediments entrained on the North American shelf in 1989 and 1995, and sampled on the shelf less than six months later. Concentrations of 239Pu + 240Pu ranged from about 0.02 to 1.8 Bq??kg-1. The two highest values came from samples collected in the central Canada Basin and near Spitsbergen; calculated backward trajectories suggest at least 14 years of circulation in the Canada Basin in the former case, and an origin near Severnaya Zemlya (at the Kara Sea/Laptev Sea boundary) in the latter case. While most of the IRD samples showed 240Pu/239Pu ratios near the mean global fallout value of 0.185, five of the samples had lower ratios, in the 0.119 to 0.166 range, indicative of mixtures of Pu from fallout and from the reprocessing of weapons-grade Pu. The backward trajectories of these five samples suggest origins in the Kara Sea or near Severnaya Zemlya.

  2. First-principles calculations of PuO(2+/-x).

    PubMed

    Petit, L; Svane, A; Szotek, Z; Temmerman, W M

    2003-07-25

    The electronic structure of PuO(2+/-x) was studied using first-principles quantum mechanics, realized with the self-interaction corrected local spin density method. In the stoichiometric PuO2 compound, Pu occurs in the Pu(IV) oxidation state, corresponding to a localized f4 shell. If oxygen is introduced onto the octahedral interstitial site, the nearby Pu atoms turn into Pu(V) (f3) by transferring electrons to the oxygen. Oxygen vacancies cause Pu(III) (f5) to form by taking up electrons released by oxygen. At T = 0, the PuO2 compound is stable with respect to free oxygen, but the delicate energy balance suggests the possible deterioration of the material during long-term storage.

  3. THERMOSTATICS AND KINETICS OF TRANSFORMATIONS IN PU-BASED ALLOYS

    SciTech Connect

    Turchi, P; Kaufman, L; Liu, Z

    2006-06-30

    CALPHAD assessment of the thermodynamic properties of a series of Pu-based alloys is briefly presented together with some results on the kinetics of phase formation and transformations in Pu-Ga alloys.

  4. Pb-Pu superlattices: an example of nanostructured actinide materials.

    PubMed

    Rudin, Sven P

    2007-03-16

    Density functional theory applied to Pb-Pu superlattices reveals two competing phases separated by a Mott transition between itinerant and localized 5f electrons. One phase, corresponding to Pu's bulk alpha phase, exhibits paired up Pu planes, thereby broadening the 5f bandwidth. Allowing spin polarization to emulate the energetics of electron correlation leads to another phase with larger volume, narrow 5f bandwidth, and more uniform local crystal structure, similar to bulk fcc Pu.

  5. Experimental Benchmarking of Pu Electronic Structure

    SciTech Connect

    Tobin, J.G.; Moore, K.T.; Chung, B.W.; Wall, M.A.; Schwartz, A.J.; Ebbinghaus, B.B.; Butterfield, M.T.; Teslich, Jr., N.E.; Bliss, R.A.; Morton, S.A.; Yu, S.W.; Komesu, T.; Waddill, G.D.; van der Laan, G.; Kutepov, A.L.

    2008-10-30

    The standard method to determine the band structure of a condensed phase material is to (1) obtain a single crystal with a well defined surface and (2) map the bands with angle resolved photoelectron spectroscopy (occupied or valence bands) and inverse photoelectron spectroscopy (unoccupied or conduction bands). Unfortunately, in the case of Pu, the single crystals of Pu are either nonexistent, very small and/or having poorly defined surfaces. Furthermore, effects such as electron correlation and a large spin-orbit splitting in the 5f states have further complicated the situation. Thus, we have embarked upon the utilization of unorthodox electron spectroscopies, to circumvent the problems caused by the absence of large single crystals of Pu with well-defined surfaces. Our approach includes the techniques of resonant photoelectron spectroscopy, x-ray absorption spectroscopy, electron energy loss spectroscopy, Fano Effect measurements, and Bremstrahlung Isochromat Spectroscopy, including the utilization of micro-focused beams to probe single-crystallite regions of polycrystalline Pu samples.

  6. Experimental Benchmarking of Pu Electronic Structure

    SciTech Connect

    Tobin, J G; Moore, K T; Chung, B W; Wall, M A; Schwartz, A J; Ebbinghaus, B B; Butterfield, M T; Teslich, Jr., N E; Bliss, R A; Morton, S A; Yu, S W; Komesu, T; Waddill, G D; der Laan, G v; Kutepov, A L

    2005-10-13

    The standard method to determine the band structure of a condensed phase material is to (1) obtain a single crystal with a well defined surface and (2) map the bands with angle resolved photoelectron spectroscopy (occupied or valence bands) and inverse photoelectron spectroscopy (unoccupied or conduction bands). Unfortunately, in the case of Pu, the single crystals of Pu are either nonexistent, very small and/or having poorly defined surfaces. Furthermore, effects such as electron correlation and a large spin-orbit splitting in the 5f states have further complicated the situation. Thus, we have embarked upon the utilization of unorthodox electron spectroscopies, to circumvent the problems caused by the absence of large single crystals of Pu with well-defined surfaces. Our approach includes the techniques of resonant photoelectron spectroscopy [1], x-ray absorption spectroscopy [1,2,3,4], electron energy loss spectroscopy [2,3,4], Fano Effect measurements [5], and Bremstrahlung Isochromat Spectroscopy [6], including the utilization of micro-focused beams to probe single-crystallite regions of polycrystalline Pu samples. [2,3,6

  7. Statistical properties of 243Pu, and 242Pu(n ,γ ) cross section calculation

    NASA Astrophysics Data System (ADS)

    Laplace, T. A.; Zeiser, F.; Guttormsen, M.; Larsen, A. C.; Bleuel, D. L.; Bernstein, L. A.; Goldblum, B. L.; Siem, S.; Garotte, F. L. Bello; Brown, J. A.; Campo, L. Crespo; Eriksen, T. K.; Giacoppo, F.; Görgen, A.; Hadyńska-KlÈ©k, K.; Henderson, R. A.; Klintefjord, M.; Lebois, M.; Renstrøm, T.; Rose, S. J.; Sahin, E.; Tornyi, T. G.; Tveten, G. M.; Voinov, A.; Wiedeking, M.; Wilson, J. N.; Younes, W.

    2016-01-01

    The level density and γ -ray strength function (γ SF ) of 243Pu have been measured in the quasicontinuum using the Oslo method. Excited states in 243Pu were populated using the 242Pu(d ,p ) reaction. The level density closely follows the constant-temperature level density formula for excitation energies above the pairing gap. The γ SF displays a double-humped resonance at low energy as also seen in previous investigations of actinide isotopes. The structure is interpreted as the scissors resonance and has a centroid of ωSR=2.42 (5 ) MeV and a total strength of BSR=10.1 (15 ) μN2 , which is in excellent agreement with sum-rule estimates. The measured level density and γ SF were used to calculate the 242Pu(n ,γ ) cross section in a neutron energy range for which there were previously no measured data.

  8. (239)Pu fallout across continental Australia: Implications on (239)Pu use as a soil tracer.

    PubMed

    Lal, R; Fifield, L K; Tims, S G; Wasson, R J

    2017-09-19

    At present there is a need for the development of new radioisotopes for soil erosion and sediment tracing especially as fallout (137)Cs levels become depleted. Recent studies have shown that (239)Pu can be a useful new soil erosion and sediment radioisotope tracer. (239)Pu was released in the major atmospheric nuclear weapons tests of 1950's and 1960's. However (239)Pu has a half-life of 24110 years and more than 99% of this isotope is still present in the environment today. In contrast (137)Cs with a half-life of 30.07 year has decayed to <35% of initially deposited activities and this isotope will become increasingly difficult to measure in the coming decades especially in the southern hemisphere, which received only about a third of the total global fallout from the atmospheric tests (UNSCEAR, 2000). In this study an assessment of the (239)Pu fallout in Australia was carried out from comparison of measured (239)Pu inventories with expected (239)Pu inventories from fallout models. (239)Pu inventories were also compared with rainfall and measured (240)Pu/(239)Pu ratios across Australia. (239)Pu fallout inventories ranged from 430 to 1461 μB/cm(2). Central Australia, with fallout 107% in excess of expected values, seems to be strongly impacted by local fallout deposition. In comparison other sites typically show 5-40% variation between expected and measured fallout values. The fallout inventories were found to weakly correlate (using power functions, y = ax(b)) with rainfall with r(2) = 0.50 across the southern catchments (25-40°S latitude band). Across the northern catchments (10-25°S latitude band) fallout showed greater variability with rainfall with r(2) = 0.24. Central Australia and Alice Springs which seem to be strongly impacted by local fallout are excluded from the rainfall correlation data (with these sites included r(2) = 0.08 and r(2) < 0.01 respectively). (240)Pu/(239)Pu atom ratios range from 0.045 to 0.197, with averages of 0

  9. [Dialectic research on "shu pu" (referring a symptom) and "shu pu" (referring a place)].

    PubMed

    Wang, Z

    2001-01-01

    The name of "shu pu" come from Shen shi fang (Prescriptions of Master Shen), which was quoted by Wai tai mi yao (Medical Secrets of an Official), compiled by Wang Tao of the Tang dynasty. In the successive processes of Zheng lei ben cao (Classified Materia Medica) and Ben cao gang mu (Compendium of Materia Medica), it was mixed up with "shu pu", a symptom, in Su wen ci jin lun (On Banning of Acupuncture in Plain Questions), and "shu pu", a place, in Zhen jiu jia yi jing (A -- B Classin of Acupuncture and Moxibustion). In fact, "shu pu" was the meat of rats with no processing, which can be used externally or administered orally to treat injury due to metallic tools and abscess.

  10. Melting behavior of mixed U-Pu oxides under oxidizing conditions

    NASA Astrophysics Data System (ADS)

    Strach, Michal; Manara, Dario; Belin, Renaud C.; Rogez, Jacques

    2016-05-01

    In order to use mixed U-Pu oxide ceramics in present and future nuclear reactors, their physical and chemical properties need to be well determined. The behavior of stoichiometric (U,Pu)O2 compounds is relatively well understood, but the effects of oxygen stoichiometry on the fuel performance and stability are often still obscure. In the present work, a series of laser melting experiments were carried out to determine the impact of an oxidizing atmosphere, and in consequence the departure from a stoichiometric composition on the melting behavior of six mixed uranium plutonium oxides with Pu content ranging from 14 to 62 wt%. The starting materials were disks cut from sintered stoichiometric pellets. For each composition we have performed two laser melting experiments in pressurized air, each consisting of four shots of different duration and intensity. During the experiments we recorded the temperature at the surface of the sample with a pyrometer. Phase transitions were qualitatively identified with the help of a reflected blue laser. The observed phase transitions occur at a systematically lower temperature, the lower the Pu content of the studied sample. It is consistent with the fact that uranium dioxide is easily oxidized at elevated temperatures, forming chemical species rich in oxygen, which melt at a lower temperature and are more volatile. To our knowledge this campaign is a first attempt to quantitatively determine the effect of O/M on the melting temperature of MOX.

  11. Electronic structure of delta-Pu and PuCoGa[sub 3] from photoemission and the mixed level model

    SciTech Connect

    Joyce, J. J.; Wills, J. M.; Durakiewicz, T.; Butterfield, M. T.; Guziewicz, E.; Sarrao, John L.,; Arko, A. J.; Moore, D. P.; Morales, L. A.; Eriksson, O.

    2004-01-01

    The electronic structure of {delta}-phase Pu metal and the Pu-based superconductor PuCoGa{sub 5} is explored using photoelectron spectroscopy and a novel theoretical scheme. Excellent agreement between calculation and experiment defines a path forward for understanding electronic structure aspects of Pu-based materials. The photoemission results show two separate regions of 5f electron spectral intensity, one at the Fermi energy and another centered 1.2 eV below the Fermi level. A comparison is made between the photoemission data and five computational schemes for {delta}-Pu. The results for {delta}-Pu and PuCoGa{sub 5} indicate 5f electron behavior on the threshold between localized and itinerant and a broader framework for understanding the fundamental electronic properties of the Pu 5f levels in general within two configurations, one localized and one itinerant.

  12. PROPERTIES AND BEHAVIOR OF 238PU RELEVANT TO DECONTAMINATION OF BUILDING 235-F

    SciTech Connect

    Duncan, A.; Kane, M.

    2009-11-24

    This report was prepared to document the physical, chemical and radiological properties of plutonium oxide materials that were processed in the Plutonium Fuel Form Facility (PuFF) in building 235-F at the Savannah River Plant (now known as the Savannah River Site) in the late 1970s and early 1980s. An understanding of these properties is needed to support current project planning for the safe and effective decontamination and deactivation (D&D) of PuFF. The PuFF mission was production of heat sources to power Radioisotope Thermoelectric Generators (RTGs) used in space craft. The specification for the PuO{sub 2} used to fabricate the heat sources required that the isotopic content of the plutonium be 83 {+-} 1% Pu-238 due to its high decay heat of 0.57 W/g. The high specific activity of Pu-238 (17.1 Ci/g) due to alpha decay makes this material very difficult to manage. The production process produced micron-sized particles which proved difficult to contain during operations, creating personnel contamination concerns and resulting in the expenditure of significant resources to decontaminate spaces after loss of material containment. This report examines high {sup 238}Pu-content material properties relevant to the D&D of PuFF. These relevant properties are those that contribute to the mobility of the material. Physical properties which produce or maintain small particle size work to increase particle mobility. Early workers with {sup 238}PuO{sub 2} felt that, unlike most small particles, Pu-238 oxide particles would not naturally agglomerate to form larger, less mobile particles. It was thought that the heat generated by the particles would prevent water molecules from binding to the particle surface. Particles covered with bound water tend to agglomerate more easily. However, it is now understood that the self-heating effect is not sufficient to prevent adsorption of water on particle surfaces and thus would not prevent agglomeration of particles. Operational

  13. Temporal variation of 240Pu/239Pu atom ratio and 239+240Pu inventory in water columns of the Japan Sea.

    PubMed

    Yamada, Masatoshi; Zheng, Jian

    2010-11-01

    The (239+240)Pu concentrations and (240)Pu/(239)Pu atom ratios were determined by alpha spectrometry and double-focusing SF-ICP-MS for seawater samples obtained in 1984 and 1993 from the Yamato and Tsushima Basins of the Japan Sea in the western North Pacific margin. The total (239+240)Pu inventories in the whole water columns were approximately doubled during the period from 1984 to 1993 in the two basins. The increasing rates were estimated to be 5.1 Bq m(-2)yr(-1) in the Yamato Basin and 4.2 Bq m(-2)yr(-1) in the Tsushima Basin and they corresponded to ~0.02% of the annual (239+240)Pu inflow rate into the Japan Sea through the Tsushima Strait. The mean (240)Pu/(239)Pu atom ratios were ~0.240 and significantly higher than the mean global fallout ratio of 0.18. Furthermore, there were no temporal or spatial variations of (240)Pu/(239)Pu atom ratios during this period in the Japan Sea. The total (239+240)Pu inventories originating from the close-in fallout increased from 17.6 Bq m(-2) to 34.6 Bq m(-2) in the Yamato Basin and from 20.1 Bq m(-2) to 34.6 Bq m(-2) in the Tsushima Basin; however, the relative percentage of ~40% from the close-in fallout was unchanged during this period. A likely mechanism for the increasing Pu inventory would be the continuous inflow of the Tsushima Current from the western North Pacific, and the removal of Pu from surface waters by scavenging onto the settling particles, followed by regeneration of Pu from the settling particles during the downward transport. Copyright © 2010 Elsevier B.V. All rights reserved.

  14. Vertical distributions of radionuclides ((239+240)Pu, (240)Pu/(239)Pu, and (137)Cs) in sediment cores of Lake Bosten in Northwestern China.

    PubMed

    Liao, Haiqing; Bu, Wenting; Zheng, Jian; Wu, Fengchang; Yamada, Masatoshi

    2014-04-01

    Artificial radionuclides ((137)Cs, (239+240)Pu, (241)Pu, (241)Am) deposited in lacustrine sediments have been used for dating as well as radionuclide source identification. In the present work, we investigated the vertical distributions of (239+240)Pu and (137)Cs activities, (240)Pu/(239)Pu atom ratios, and (239+240)Pu/(137)Cs activity ratios in sediment cores collected from Lake Bosten, which is the lake closest to the Lop Nor Chinese Nuclear Weapon Test site in northwestern China. Uniformly high concentrations of (239+240)Pu and (137)Cs were found in the upper layers deposited since 1964 in the sediment cores, and these were controlled by the resuspension of soil containing radionuclides from the nearby land surface. As the Chinese nuclear tests varied remarkably in yield, the mixing of the tropospheric deposition from these tests and the stratospheric deposition of global fallout has led to a (240)Pu/(239)Pu atom ratio that is similar to that of global fallout and to a (239+240)Pu/(137)Cs activity ratio that is slightly higher than that of global fallout. However, a low (240)Pu/(239)Pu atom ratio of 0.080 and high (239+240)Pu/(137)Cs activity ratio of 0.087, significantly different from the global fallout values, were observed in one sediment core (07BS10-2), indicating the inhomogenous tropospheric deposition from the Chinese nuclear tests in Lake Bosten during 1967-1973. These results are important to understand the influence of the CNTs on the radionuclide contamination in Lake Bosten.

  15. Investigating Pu and U isotopic compositions in sediments: a case study in Lake Obuchi, Rokkasho Village, Japan using sector-field ICP-MS and ICP-QMS.

    PubMed

    Zheng, Jian; Yamada, Masatoshi

    2005-08-01

    The objectives of the present work were to study isotope ratios and the inventory of plutonium and uranium isotope compositions in sediments from Lake Obuchi, which is in the vicinity of several nuclear fuel facilities in Rokkasho, Japan. Pu and its isotopes were determined using sector-field ICP-MS and U and its isotopes were determined with ICP-QMS after separation and purification with a combination of ion-exchange and extraction chromatography. The observed (240)Pu/(239)Pu atom ratio (0.186 +/- 0.016) was similar to that of global fallout, indicating that the possible early tropospheric fallout Pu did not deliver Pu from the Pacific Proving Ground to areas above 40 degrees N. The previously reported higher Pu inventory in the deep water area of Lake Obuchi could be attributed to the lateral transportation of Pu deposited in the shallow area which resulted from the migration of deposited global fallout Pu from the land into the lake by river runoff and from the Pacific Ocean by tide movement and sea water scavenging, as well as from direct soil input by winds. The (235)U/(238)U atom ratios ranged from 0.00723 to 0.00732, indicating the natural origin of U in the sediments. The average (234)U/(238)U activity ratio of 1.11 in a sediment core indicated a significant sea water U contribution. No evidence was found for the release of U containing wastes from the nearby nuclear facilities. These results will serve as a reference baseline on the levels of Pu and U in the studied site so that any further contamination from the spent nuclear fuel reprocessing plants, the radioactive waste disposal and storage facilities, and the uranium enrichment plant can be identified, and the impact of future release can be rapidly assessed.

  16. Determination of (239)Pu, (240)Pu, (241)Pu and (242)Pu at femtogram and attogram levels - evidence for the migration of fallout plutonium in an ombrotrophic peat bog profile.

    PubMed

    Quinto, Francesca; Hrnecek, Erich; Krachler, Michael; Shotyk, William; Steier, Peter; Winkler, Stephan R

    2013-04-01

    The isotopic composition of plutonium ((239)Pu, (240)Pu, (241)Pu and (242)Pu) was investigated in a ∼0.5 m long peat core from an ombrotrophic bog (Black Forest, Germany) using clean room procedures and accelerator mass spectrometry (AMS). This sophisticated analytical approach was ultimately needed to detect reliably the Pu concentrations present in the peat samples at femtogram (fg) and attogram (ag) levels. The mean (240)Pu/(239)Pu isotopic ratio of 0.19 ± 0.02 (N = 32) in the peat layers, representing approximately the last 80 years, was in good agreement with the accepted value of 0.18 for the global fallout in the Northern Hemisphere. This finding is largely supported by the corresponding and rather constant (241)Pu/(239)Pu (0.0012 ± 0.0005) and (242)Pu/(239)Pu (0.004 ± 0.001) ratios. Since the Pu isotopic composition characteristic of the global fallout was also identified in peat samples pre-dating the period of atmospheric atom bomb testing (AD 1956-AD 1980), migration of Pu within the peat profile is clearly indicated. These results highlight, for the first time, the mobility of Pu in a peat bog with implications for the migration of Pu in other acidic, organic rich environments such as forest soils and other wetland types. These findings constitute a direct observation of the behaviour of Pu at fg and ag levels in the environment. The AMS measurements of Pu concentrations (referring to a corresponding activity of (240+239)Pu from 0.07 mBq g(-1) to 5 mBq g(-1)) essentially confirm our a priori estimates based on existing (241)Am and (137)Cs data in the investigated peat core and agree well with the global fallout levels from the literature. Exclusively employing the Pu isotope ratios established for the peat samples, the date of the Pu irradiation (AD 1956, correctable to AD 1964) was calculated and subsequently compared to the (210)Pb age of the peat layers; this comparison provided an additional hint that global fallout derived Pu is not fixed in

  17. Modeling of anodic dissolution of U Pu Zr ternary alloy in the molten LiCl KCl electrolyte

    NASA Astrophysics Data System (ADS)

    Iizuka, Masatoshi; Kinoshita, Kensuke; Koyama, Tadafumi

    2005-02-01

    The metallic fuel anode in the molten salt electrorefining step for the pyrometallurgical reprocessing was modeled based on the findings from the anodic dissolution tests using a U Pu Zr ternary alloy. This anode model simulates selective dissolution of uranium and plutonium at lower anode potential, growth of a diffusion controlling layer consisting of a mixture of the molten salt electrolyte and the remaining zirconium metal, and simultaneous dissolution of all the constituents at higher anode potential. The calculation with this model reproduced well the actual anodic behavior of the U Pu Zr ternary alloy such as two-step rapid rise in the anode potential.

  18. Isotopic compositions of (236)U and Pu isotopes in "black substances" collected from roadsides in Fukushima prefecture: fallout from the Fukushima Dai-ichi nuclear power plant accident.

    PubMed

    Sakaguchi, Aya; Steier, Peter; Takahashi, Yoshio; Yamamoto, Masayoshi

    2014-04-01

    Black-colored road dusts were collected in high-radiation areas in Fukushima Prefecture. Measurement of (236)U and Pu isotopes and (134,137)Cs in samples was performed to confirm whether refractory elements, such as U and Pu, from the fuel core were discharged and to ascertain the extent of fractionation between volatile and refractory elements. The concentrations of (134,137)Cs in all samples were exceptionally high, ranging from 0.43 to 17.7 MBq/kg, respectively. (239+240)Pu was detected at low levels, ranging from 0.15 to 1.14 Bq/kg, and with high (238)Pu/(239+240)Pu activity ratios of 1.64-2.64. (236)U was successfully determined in the range of (0.28 to 6.74) × 10(-4) Bq/kg. The observed activity ratios for (236)U/(239+240)Pu were in reasonable agreement with those calculated for the fuel core inventories, indicating that trace amounts of U from the fuel cores were released together with Pu isotopes but without large fractionation. The quantities of U and (239+240)Pu emitted to the atmosphere were estimated as 3.9 × 10(6) Bq (150 g) and 2.3 × 10(9) Bq (580 mg), respectively. With regard to U, this is the first report to give a quantitative estimation of the amount discharged. Appreciable fractionation between volatile and refractory radionuclides associated with the dispersal/deposition processes with distance from the Fukushima Dai-ichi Nuclear Power Plant was found.

  19. Biokinetics of sup 237 Pu citrate and nitrate in the rat: Implications for Pu studies in man

    SciTech Connect

    Talbot, R.J.; Knight, D.A.; Morgan, A. )

    1990-08-01

    Plutonium-237 decays mainly by electron capture with a half-life of 45 d. Alpha particles are emitted in only 5 x 10(-3)% of its disintegrations. This nuclide can now be produced with relatively small amounts of alpha-emitting contaminants so that, in principle, {sup 237}Pu can be used for studies of Pu biokinetics in man. However, because of its high specific activity, there was some doubt that its metabolism would be the same as that of the alpha- and beta-emitting isotopes of Pu normally encountered in the nuclear industry. In this study, the biokinetics of nearly pure, high specific activity {sup 237}Pu are compared with those of lower specific activity, impure {sup 237}Pu containing significant amounts of alpha-emitting Pu, following administration to rats by intravenous injection as the citrate. Both the distribution and excretion of the pure and impure {sup 237}Pu used in the two studies were similar and also in good agreement with the results of previously reported studies using {sup 239}Pu and {sup 241}Pu citrate, thus validating the use of {sup 237}Pu for studies of Pu metabolism in man. Data on the biokinetics of {sup 237}Pu nitrate are also included.

  20. Redox reactions of Pu(IV) and Pu(III) in the presence of acetohydroxamic acid in HNO(3) solutions.

    PubMed

    Tkac, Peter; Precek, Martin; Paulenova, Alena

    2009-12-21

    The reduction of Pu(IV) in the presence of acetohydroxamic acid (HAHA) was monitored by vis-NIR spectroscopy. All experiments were performed under low HAHA/Pu(IV) ratios, where only the Pu(IV)-monoacetohydroxamate complex and Pu uncomplexed with HAHA were present in relevant concentrations. Time dependent concentrations of all absorbing species were resolved using molar extinction coefficients for Pu(IV), Pu(III), and the Pu(AHA)(3+) complex by deconvolution of spectra. From fitting of the experimental data by rate equations integrated by a numeric method three reactions were proposed to describe a mechanism responsible for the reduction and oxidation of plutonium in the presence of HAHA and HNO(3). Decomposition of Pu(AHA)(3+) follows a second order reaction mechanism with respect to its own concentration and leads to the formation of Pu(III). At low HAHA concentrations, a two-electron reduction of uncomplexed Pu(IV) with HAHA also occurs. Formed Pu(III) is unstable and slowly reoxidizes back to Pu(IV), which, at the point when all HAHA is decomposed, can be catalyzed by the presence of nitrous acid.

  1. Pu speciation in actual and simulated aged wastes

    SciTech Connect

    Lezama-pacheco, Juan S; Conradson, Steven D

    2008-01-01

    X-ray Absorption Fine Structure Spectroscopy (XAFS) at the Pu L{sub II/III} edge was used to determine the speciation of this element in (1) Hanford Z-9 Pu crib samples, (2) deteriorated waste resins from a chloride process ion-exchange purification line, and (3) the sediments from two Waste Isolation Pilot Plant Liter Scale simulant brine systems. The Pu speciation in all of these samples except one is within the range previously displayed by PuO{sub 2+x-2y}(OH){sub y}{center_dot}zH{sub 2}O compounds, which is expected based on the putative thermodynamic stability of this system for Pu equilibrated with excess H{sub 2}O and O{sub 2} under environmental conditions. The primary exception was a near neutral brine experiment that displayed evidence for partial substitution of the normal O-based ligands with Cl{sup -} and a concomitant expansion of the Pu-Pu distance relative to the much more highly ordered Pu near neighbor shell in PuO{sub 2}. However, although the Pu speciation was not necessarily unusual, the Pu chemistry identified via the history of these samples did exhibit unexpected patterns, the most significant of which may be that the presence of the Pu(V)-oxo species may decrease rather than increase the overall solubility of these compounds. Several additional aspects of the Pu speciation have also not been previously observed in laboratory-based samples. The molecular environmental chemistry of Pu is therefore likely to be more complicated than would be predicted based solely on the behavior of PuO{sub 2} under laboratory conditions.

  2. Complementary Pu Resuspension Study at Palomares, Spain

    SciTech Connect

    Shinn, J

    2002-10-01

    Soil in an area near Palomares, Spain, was contaminated with plutonium as a result of a mid-air collision of U.S. military aircraft in January 1966. The assessment for potential inhalation dose can be found in Iranzo et al., (1987). Long-term monitoring has been used to evaluate remedial actions (Iranzo et al., 1988) and there are many supporting studies of the Pu contamination at Palomares that have been carried out by the Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT) in Madrid. The purpose of this study is to evaluate the resuspension of Pu from the soil in terms of Pu-concentrations in air and resuspension rates in a complementary investigation to those of CIEMAT but in an intensive short-term field effort. This study complements the resuspension studies of CIEMAT at Palomares with additional information, and with confirmation of their previous studies. Observed mass loadings (M) were an average of 70 mg/m{sup 3} with peaks in the daytime of 130 mg/m{sup 3} and low values at night below 30 {micro}g/m{sup 3}. The Pu-activity of aerosols (A) downwind of plot 2-1 was 0.12 Bq/g and the enhancement factor (E{sub f}) had a value of 0.3, which is low but similar to a typical value of 0.7 for other undisturbed sites. This E{sub f} value may increase further away from ground zero. The particle size distribution of the Pu in air measured by cascade impactors was approximately lognormal with a median aerodynamic diameter of 3.7 {micro}m and a geometric standard deviation of 3.5 in the respirable range. This peak midway between 1 ? m and 10 {micro}m in the respirable range is commonly observed. Daily fluctuations in the Pu concentration in air (C) detected by the UHV were lognormally distributed with a geometric standard deviation of 4.9 indicating that the 98th percentile would be 24 times as high as the median. Downwind of plot 2-1 the mean Pu concentration in air, C, was 8.5 {micro}Bq/m{sup 3}. The resuspension factor (Sf) was 2.4 x 10

  3. Plutonium partitioning in uranium and plutonium co-recovery system for fast reactor fuel recycling with enhanced nuclear proliferation resistance

    SciTech Connect

    Nakahara, Masaumi; Koma, Yoshikazu; Nakajima, Yasuo

    2013-07-01

    For enhancement of nuclear proliferation resistance, a 'co-processing' method for U and Pu co-recovery was studied. Two concepts, no U scrubbing and no Pu reduction partitioning, were employed to formulate two types of flow sheets by using a calculation code. Their process performance was demonstrated using radioactive solutions derived from an irradiated fast reactor fuel. These experimental results indicated that U and Pu were co-recovered in the U/Pu product, and the Pu content in the U/Pu product increased approximately 2.3 times regardless of using reductant. The proposed no U scrubbing and no Pu reductant flow sheet is applicable to fast reactor fuel reprocessing and enhances its resistance to nuclear proliferation. (authors)

  4. Irradiation performance of full-length metallic IFR fuels

    SciTech Connect

    Tsai, H.; Neimark, L.A.

    1992-07-01

    An assembly irradiation of 169 full-length U-Pu-Zr metallic fuel pins was successfully completed in FFTF to a goal burnup of 10 at.%. All test fuel pins maintained their cladding integrity during the irradiation. Postirradiation examination showed minimal fuel/cladding mechanical interaction and excellent stability of the fuel column. Fission-gas release was normal and consistent with the existing data base from irradiation testing of shorter metallic fuel pins in EBR-II.

  5. Formation and stability of metastable structures and amorphous phases in PU-V, PU-TA, and PU-YB systems with positive heats of mixing

    NASA Astrophysics Data System (ADS)

    Rizzo, H. F.; Zocco, T.; Massalski, T. B.; Nastasi, M.; Echeverria, A.

    1994-08-01

    The triode sputtering technique with a “split-target” arrangement was used to obtain metastable crystalline and amorphous phases in the Pu-V, Pu-Ta, and Pu-Yb systems. The proposed phase diagrams for these systems all exhibit liquid immiscibility. The heats of mixing are estimated to be highly positive, and the atomic radii of the component atoms differ by at least 10 pct. Extended amorphous and body-centered cubic (bcc) solid-solution regions were observed in the Pu-V and Pu-Ta systems. The corresponding lattice parameters appear to follow in each case an assumed Vegard’s Law extension. In the Pu-Yb system, no amorphous phase was obtained, but an extended face-centered cubic (fcc) solid-solution region (24 to 78 at. pct Yb) was observed with a large positive deviation of the lattice parameter (˜9 pct at 40 at. pct Yb) from a linear Vegard’s Law between the pure fcc components. The observed ranges of amorphous and metastable solid-solution phases have been interpreted in terms of predicated heats of formation for these phases using Miedema’s thermodynamic approximations that include chemical, elastic, and structural contributions. The effect of the high deposition rates on the formation of amorphous and metastable phases has also been considered. Thermal annealing of Pu-Ta amorphous alloys brings about a rapid diffusion of Pu to the free surface of the amorphous phase without crystallization of the remaining Ta-rich amorphous phase. Microhardness measurements indicate that amorphous Pu-V and Pu-Ta alloys are softer than the crystalline bcc solid-solution alloys in the same composition range. Several similarities in the formation of mixed phase regions (amorphous and solid solutions), microhardness, and resistance to decomposition on heating were noted between the Pu-Ta and Pu-V systems and the Cu-W system studied previously.

  6. FCRD Transmutation Fuels Handbook 2015

    SciTech Connect

    Janney, Dawn Elizabeth; Papesch, Cynthia Ann

    2015-09-01

    Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. It is, therefore, important to understand the properties of U-Pu-Zr alloys, both with and without minor actinide additions. In addition to requiring extensive safety precautions, alloys containing U and Pu are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phase-transformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, and that general acceptance of results sometimes indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, and alloys of two or three of these elements. It contains information about phase diagrams and related information (including phases and phase transformations); heat capacity, entropy, and enthalpy; thermal expansion; and thermal conductivity and diffusivity. In addition to presenting information about materials properties, it attempts to provide information about how well the property is known and how much variation exists between measurements. Although the handbook includes some references to publications about modeling

  7. PuPO4(cr, hyd.) Solubility Product and Pu3+ Complexes With Phosphate and Ethylenediaminetetraacetic Acid

    SciTech Connect

    Rai, Dhanpat; Moore, Dean A.; Felmy, Andrew R.; Rosso, Kevin M.; Bolton, Harvey

    2010-06-15

    To determine the solubility product of PuPO4(cr, hyd.) and the complexation constants of Pu(III) with phosphate and EDTA, the solubility of PuPO4(cr, hyd.) was investigated as a function of: 1) time and pH varying from 1.0 to 12.0 and at a fixed 0.00032 M phosphate concentration; 2) NaH2PO4 concentrations varying from 0.0001 M to 1.0 M and at a fixed pH value of 2.5; 3) time and pH varying from 1.3 to 13.0 at fixed concentrations of 0.00032 M phosphate and 0.0004 M or 0.002 M Na2H2EDTA; and 4) Na2H2EDTA concentrations varying from 0.00005 M to 0.0256 M at a fixed 0.00032 M phosphate concentration and at pH values of approximately 3.5, 10.6, and 12.6. A combination of solvent extraction and spectrophotometric techniques confirmed that the use of hydroquinone and Na2S2O4 helped maintain Pu as Pu(III). The solubility data were interpreted using Pitzer and SIT models, and both provided similar values for the solubility product of PuPO4(cr, hyd.) and for the formation constant of PuEDTA-. The log10 of the solubility product of PuPO4(cr, hyd.) (PuPO4(cr, hyd.) = Pu3+ + PO4 ) was determined to be –(24.42 ± 0.38). Pitzer modeling showed that phosphate interactions with Pu3+ were extremely weak and did not require any phosphate complexes (e.g., PuPO4(aq), PuH2PO42+, Pu(H2PO4)2+, Pu(H2PO4)3(aq), and Pu(H2PO4)4-), as proposed in existing literature, to explain the experimental data. SIT modeling, however, required the inclusion of PuH2PO42+ to explain the data in high NaH2PO4 concentrations; this illustrates the differences one can expect when using these two chemical models to interpret the data. As the Pu(III)-EDTA species, only PuEDTA- was needed to interpret the experimental data in a large range in pH values (1.3–12.9) and EDTA concentrations (0.00005–0.256 M). Calculations based on density functional theory support the existence of PuEDTA- (with prospective stoichiometry as Pu(OH2)3EDTA-) as the chemically and structurally stable species. The log10 of the

  8. ²³⁹Pu and ²⁴⁰Pu inventories and ²⁴⁰Pu/²³⁹Pu atom ratios in the equatorial Pacific Ocean water column.

    PubMed

    Yamada, Masatoshi; Zheng, Jian

    2012-07-15

    The (239+240)Pu concentrations and (240)Pu/(239)Pu atom ratios were determined by alpha spectrometry and inductively coupled plasma mass spectrometry for seawater samples from two stations, one at the equator and the other in the equatorial South Pacific. To better understand the fate of Pu isotopes, this study dealt with the contribution of the close-in fallout Pu from the Pacific Proving Grounds (PPG) in water columns of the Pacific Ocean. The (239)Pu, (240)Pu and (239+240)Pu inventories over the depth interval 0-3000 m at the equator station were 10.4, 8.9 and 19.3 Bq m(-2), respectively. Further, no noticeable difference was observed in (239)Pu, (240)Pu and (239+240)Pu inventories over the depth interval 0-3000 m between the two stations. The total (239+240)Pu inventories were significantly higher than the expected cumulative deposition density of global fallout. Water column (239+240)Pu inventories measured in this study were lower than those reported for comparable stations in the Geochemical Ocean Sections Study, indicating that these inventories have been decreasing at average rates of 0.89 ± 0.07 and 0.16 ± 0.07 Bq m(-2)yr(-1) at the equator and equatorial South Pacific stations, respectively, from 1973 to 1990. The obtained (240)Pu/(239)Pu atom ratios were higher than the mean global fallout ratio of 0.18. These high atom ratios proved the existence of close-in tropospheric fallout Pu from the PPG in the Marshall Islands. The (239+240)Pu inventories originating from the close-in fallout in the entire water column were estimated to be 11.1 Bq m(-2) at the equator station and 7.1 Bq m(-2) at the equatorial South Pacific Ocean station, and the relative percentages of close-in fallout Pu were 40% at the former and 34% at the latter. A significant amount of close-in fallout Pu originating from the PPG has been transported to deep layers below the 1000 m depth in the equatorial Pacific Ocean.

  9. Assessment of Startup Fuel Options for a Test or Demonstration Fast Reactor

    SciTech Connect

    Carmack, Jon; Hayes, Steven; Walters, L. C.

    2015-09-01

    This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO2 and UO2-PuO2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availability are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.

  10. Pu-content verification by calorimetry

    SciTech Connect

    Beets, C.; Carchon, R.; Fettweis, P.; Corbellini, M.; D'Adama, D.; Guardini, S.; Rodenburg, W.W.; Strohm, W.W.; Fiarman, S.; Keddar, A.

    1984-05-01

    The aim of this paper is to present calorimetric assay measurements that were performed on a set of high burnup plutonium samples using the Mound No. 150 twin resistance bridge calorimeter. The samples and the calorimeter are described in the first part of the paper and the experimental results are discussed in a second part. The isotopic composition, obtained by destructive analysis, was combined with the measured power values for the Pu-assay.

  11. Nevada test site fallout atom ratios: /sup 240/Pu//sup 239/Pu and /sup 241/Pu//sup 239/Pu

    SciTech Connect

    Hicks, H.G.; Barr, D.W.

    1984-02-01

    The exposure of the population in Utah to external gamma radiation from the fallout from nuclear weapons tests carried out between 1951 and 1958 at the Nevada Test Site (NTS) has been reconstructed from recent measurements of /sup 137/Cs and plutonium in soil. The fraction of /sup 137/Cs in the fallout from NTS events was calculated from the total plutonium and the /sup 240/Pu//sup 239/Pu ratios measured in the soil, using the values of 0.180 +- 0.006 and 0.032 +- 0.003 for that ratio in global fallout and NTS fallout, respectively. The total population exposure from NTS events was then calculated on the basis of exposure rates resulting from short-lived radionuclides associated with the /sup 137/Cs at the time of deposition. While the /sup 240/Pu//sup 239/Pu ratio is constant in global fallout, this ratio varies greatly in the fallout from individual events. While the composition of fallout on Utah from NTS events is rather uniform, the Off-Site Radiation Exposure Review Project is currently reconstructing radiation exposures for locations close to NTS where the fallout may be predominantly from one event. Therefore, the authors compiled the pertinent ratios in order to provide information concerning the exposure resulting from any individual event. The plutonium ratios measured at 30 days postshot were compiled from unpublished values in the archives of the Nuclear Chemistry Division of LLNL and INC-11 of LANL. These ratios are pertinent to fallout data. Dates for each event were taken from a publication by the Nevada Operations Office of the Department of Energy. 3 references.

  12. Gallium Content in PuO{sub 2} Using Laser Induced Breakdown Spectroscopy (LIBS)

    SciTech Connect

    Smith, C.A.; Martinez, M.A.; Veirs, D.K.

    1999-08-29

    Laser Induced Breakdown Spectroscopy (LIBS) has been applied to the semi-quantitative analysis of gallium in plutonium oxide at the Los Alamos Plutonium Facility. The oxide samples were generated by the Thermally Induced Gallium Removal (TIGR) process, a pretreatment step prior to MOX fuel processing. The TIGR process uses PuO{sub 2} containing 1 wt% gallium (nominal) as feed material. Following the TIGR process, gallium content was analyzed by LIBS and also by conventional wet chemical analysis (ICP-MS). Although the data range was insufficient to obtain an adequate calibration, general agreement between the two techniques was good. LIBS was found to have a useful analytical range of 34-400 ppm for Ga in PuO{sub 2}.

  13. Nuclear Resonance Fluorescence Excitations Near 2 MeV in 235U and 239Pu

    SciTech Connect

    Bertozzi, W; Caggiano, J A; Hensley, W K; Johnson, M S; Korbly, S E; Ledoux, R J; McNabb, D P; Norman, E B; Park, W H; Warren, G A

    2006-12-27

    A search for nuclear resonance fluorescence excitations in {sup 235}U and {sup 239}Pu within the energy range of 1.0- to 2.5-MeV was performed using a 4-MeV continuous bremsstrahlung source at the High Voltage Research Laboratory at the Massachusetts Institute of Technology. Measurements utilizing high purity Ge detectors at backward angles identified 9 photopeaks in {sup 235}U and 12 photopeaks in {sup 239}Pu in this energy range. These resonances provide unique signatures that allow the materials to be non-intrusively detected in a variety of environments including fuel cells, waste drums, vehicles and containers. The presence and properties of these states may prove useful in understanding the mechanisms for mixing low-lying collective dipole excitations with other states at low excitations in heavy nuclei.

  14. Recent advances in the study of the UO2-PuO2 phase diagram at high temperatures

    NASA Astrophysics Data System (ADS)

    Böhler, R.; Welland, M. J.; Prieur, D.; Cakir, P.; Vitova, T.; Pruessmann, T.; Pidchenko, I.; Hennig, C.; Guéneau, C.; Konings, R. J. M.; Manara, D.

    2014-05-01

    Recently, novel container-less laser heating experimental data have been published on the melting behaviour of pure PuO2 and PuO2-rich compositions in the uranium dioxide-plutonium dioxide system. Such data showed that previous data obtained by more traditional furnace heating techniques were affected by extensive interaction between the sample and its containment. It is therefore paramount to check whether data so far used by nuclear engineers for the uranium-rich side of the pseudo-binary dioxide system can be confirmed or not. In the present work, new data are presented both in the UO2-rich part of the phase diagram, most interesting for the uranium-plutonium dioxide based nuclear fuel safety, and in the PuO2 side. The new results confirm earlier furnace heating data in the uranium-dioxide rich part of the phase diagram, and more recent laser-heating data in the plutonium-dioxide side of the system. As a consequence, it is also confirmed that a minimum melting point must exist in the UO2-PuO2 system, at a composition between x(PuO2) = 0.4 and x(PuO2) = 0.7 and 2900 K ⩽ T ⩽ 3000 K. Taking into account that, especially at high temperature, oxygen chemistry has an effect on the reported phase boundary uncertainties, the current results should be projected in the ternary U-Pu-O system. This aspect has been extensively studied here by X-ray diffraction and X-ray absorption spectroscopy. The current results suggest that uncertainty bands related to oxygen behaviour in the equilibria between condensed phases and gas should not significantly affect the qualitative trend of the current solid-liquid phase boundaries.

  15. Magnetic Properties of Radiation Damage in Pu

    SciTech Connect

    McCall, S; Fluss, M J; Chung, B W; McElfresh, M; Chapline, G; Jackson, D

    2004-10-27

    First, we review earlier studies reporting possible magnetic characteristics for radiation defects in Pu. We then report, for {alpha}-Pu, two studies of the excess magnetic susceptibility (EMS) due to radiation damage, as a function of time and temperature. We have observed several annealing stages associated with the EMS of the accumulated self-damage and we report that annealing begins at {approx}31K, while below that temperature the displacement damage from self-irradiation of the Pu alpha particle emission and the U recoil are immobile. A detailed investigation was made of this EMS well below the first annealing stage as a function of temperature (2K < T < 15K) and time in a magnetic field of 2T. A linear increase in magnetic susceptibility is seen as a function of time for all isotherms. The excess susceptibility per alpha decay, determined from a linear fit of the slope of the time dependent EMS, is reasonably described with a Curie-Weiss law exhibiting a small negative Weiss temperature. We conclude by describing some future experiments in light of the present results.

  16. 240Pu/239Pu isotopic ratios and 239 + 240Pu total measurements in surface and deep waters around Mururoa and Fangataufa atolls compared with Rangiroa atoll (French Polynesia).

    PubMed

    Chiappini, R; Pointurier, F; Millies-Lacroix, J C; Lepetit, G; Hemet, P

    1999-09-30

    The average values of 240Pu/239Pu mass isotopic ratios of plutonium deposited in Mururoa and Fangataufa atoll sediments by French atmospheric nuclear tests range from 3.5 to 5%. In order to assess the near field and far field influence of those deposits in the open ocean, two water profiles were measured for 239 + 240Pu and 240Pu/239Pu using, for the first time, an Inductively Coupled Plasma Mass Spectrometer which was developed to achieve femtogram detection limits. One site was located at the limit of the French territorial waters, which is 22 km distant from Mururoa. The second site was located close to Rangiroa atoll, at a distance of approximately 1200-km from French nuclear test sites. The sample volumes were approximately 500 litres and plutonium was purified prior to mass spectrometry and alpha spectrometry measurements. In Rangiroa, the 239 + 240Pu profile is comparable with those already determined in world open oceans but the maximum detected activity, 9 mBq/m3 at 500-600 m is a lot lower than those measured in the northern hemisphere. 240Pu/239Pu ratios were measured between 500 and 1000 m and were not statistically different from the typical 0.18 +/- 0.01 ratio which characterises the global fallout. Consequently, any influence of plutonium from the tests in Mururoa and Fangataufa is not apparent at Rangiroa. The vertical distribution of 239 + 240Pu near Mururoa shows similar changes with depth but with a slight increase in concentration. 240Pu/239Pu mass ratios vary with depth, from 7 to 10% in the upper 500 m and in the deep waters (below 1000 m) to 15-16% between 600 and 1000 m. A contribution from plutonium deposited in the sediments at Mururoa and Fangataufa is observed at the limit of territorial waters, especially in surface and deep waters.

  17. Optimization of burnable poison design for Pu incineration in fully fertile free PWR core

    SciTech Connect

    Fridman, E.; Shwageraus, E.; Galperin, A.

    2006-07-01

    The design challenges of the fertile-free based fuel (FFF) can be addressed by careful and elaborate use of burnable poisons (BP). Practical fully FFF core design for PWR reactor has been reported in the past [1]. However, the burnable poison option used in the design resulted in significant end of cycle reactivity penalty due to incomplete BP depletion. Consequently, excessive Pu loading were required to maintain the target fuel cycle length, which in turn decreased the Pu burning efficiency. A systematic evaluation of commercially available BP materials in all configurations currently used in PWRs is the main objective of this work. The BP materials considered are Boron, Gd, Er, and Hf. The BP geometries were based on Wet Annular Burnable Absorber (WABA), Integral Fuel Burnable Absorber (IFBA), and Homogeneous poison/fuel mixtures. Several most promising combinations of BP designs were selected for the full core 3D simulation. All major core performance parameters for the analyzed cases are very close to those of a standard PWR with conventional UO{sub 2} fuel including possibility of reactivity control, power peaking factors, and cycle length. The MTC of all FFF cores was found at the full power conditions at all times and very close to that of the UO{sub 2} core. The Doppler coefficient of the FFF cores is also negative but somewhat lower in magnitude compared to UO{sub 2} core. The soluble boron worth of the FFF cores was calculated to be lower than that of the UO{sub 2} core by about a factor of two, which still allows the core reactivity control with acceptable soluble boron concentrations. The main conclusion of this work is that judicial application of burnable poisons for fertile free fuel has a potential to produce a core design with performance characteristics close to those of the reference PWR core with conventional UO{sub 2} fuel. (authors)

  18. Comparative Analysis on Nuclear Fuel Sustainability Aspect of FBR

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Irwanto, Dwi; Suzuki, Mitsutoshi; Saito, Masaki

    2017-01-01

    Recycle program of spent nuclear fuel (SNF) will have some challanges in term of fuel cycle capability and its facilities as well as nuclear non-proliferation concern of special nuclear materials. A different analysis approach as a comparative study have been analyzed based on breeding ratio and heavy metal inventory ratio concepts in fast breeder reactor (FBR) type. Breeding ratio and heavy metal inventory obtain higher than unity which shows breeding gain or surplus inventory of heavy metals are obtained. Breeding ratio indicates the fuel conversion capability from conversion process of fertile materials into fissile material such as fertile materials of U-238, Pu-238, Pu-240 and fissile materials of Pu-239 and Pu-241. Inventory ratio approaches are appropriate to estimate some selected actinide as a mass inventory production such as plutonium inventory ratio which estimate the surplus mass inventory from the ratio of produced plutonium at the net of operation to the initial inventory ratio.

  19. X-ray excited Auger transitions of Pu compounds

    SciTech Connect

    Nelson, Art J. Grant, William K.; Stanford, Jeff A.; Siekhaus, Wigbert J.; Allen, Patrick G.; McLean, William

    2015-05-15

    X-ray excited Pu core–valence–valence and core–core–valence Auger line-shapes were used in combination with the Pu 4f photoelectron peaks to characterize differences in the oxidation state and local electronic structure for Pu compounds. The evolution of the Pu 4f core-level chemical shift as a function of sputtering depth profiling and hydrogen exposure at ambient temperature was quantified. The combination of the core–valence–valence Auger peak energies with the associated chemical shift of the Pu 4f photoelectron line defines the Auger parameter and results in a reliable method for definitively determining oxidation states independent of binding energy calibration. Results show that PuO{sub 2}, Pu{sub 2}O{sub 3}, PuH{sub 2.7}, and Pu have definitive Auger line-shapes. These data were used to produce a chemical state (Wagner) plot for select plutonium oxides. This Wagner plot allowed us to distinguish between the trivalent hydride and the trivalent oxide, which cannot be differentiated by the Pu 4f binding energy alone.

  20. Pu-241 in samples of forest soil from Poland.

    PubMed

    Mietelski, J W; Dorda, J; Was, B

    1999-10-01

    241Pu activity measurements in coniferous forest soil samples from Poland are presented. The results were obtained in two ways: by alpha spectrometric remeasurements of the plutonium sources 3-4 years after their preparation (i.e. by the 241Am ingrowth) and by direct measurements of 241Pu using liquid scintillation (LS) spectrometry. Both methods gave consistent results. The maximum observed activity concentration obtained by the 241Am ingrowth was (254 +/- 43) Bq/kg, and by direct measurements it was (284 +/- 31) Bq/kg (the same sample, activities calculated for May 1, 1986). Enhanced levels of 241Pu were observed in all samples from the farthest north-eastern Poland. The estimated 241Pu maximum deposition from Chernobyl fallout in this area (sum of deposition for two examined layers of one site) was (1.025 +/- 0.110) kBq/m2. This will result after 70 years in an additional 241Am activity of about (30.1 +/- 3.2) Bq/m2. The average ratio for 241Pu to total 238,239,240Pu was of the order of 25. The obtained average Chernobyl ratios for 241Pu to 239,240Pu were about 86, and those for 241Pu to 238,239,240Pu were 56.

  1. Sintering of compacts of UN, (U,Pu)N, and PuN

    DOEpatents

    Tennery, V.J.; Godfrey, T.G.; Bomar, E.S.

    1973-10-16

    >A method is provided for preparing a densified compact of a metal nitride selected from the group consisting of UN, (U,Pu)N, and PuN which comprises heating a green compact of at least one selected nitride in the mononitride single-phase region, as displayed by a phase diagram of the mononitride of said compact, in a nitrogen atmosphere at a pressure of nitrogen less than 760 torr. At a given temperature, this process produces a singlephase structure and a maximal sintered density as measured by mercury displacement. (Official Gazette)

  2. 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.

  3. The optimization of an AP1000 fuel assembly for the transmutation of plutonium and minor actinides

    NASA Astrophysics Data System (ADS)

    Washington, Jeremy A.

    The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. The goal of this thesis is to examine the potential of light water reactors for plutonium and minor actinides transmutation as a near-term solution. This thesis screens the available nuclear isotope database to identify potential absorbers as coatings on a transmutation fuel in a light water reactor. A spectral shift absorber coating tunes the neutron energy spectrum experienced by the underlying target fuel. Eleven different spectral shift absorbers (B4C, CdO, Dy2O3, Er 2O3, Eu2O3, Gd2O3, HfO2, In2O3, Lu2O3, Sm2O3, and TaC) have been selected for further evaluation. A model developed using the NEWT module of SCALE 6.1 code provided performance data for the burnup of the target fuel rods. Irradiation of the target fuels occurs in a Westinghouse 17x17 XL Robust Fuel Assembly over a 1400 Effective Full Power Days (EFPD) interval. The fuels evaluated in this thesis include PuO2, Pu3Si2, PuN, MOX, PuZrH, PuZrHTh, PuZrO 2, and PuUZrH. MOX (5 wt% PuO2), Pu0.31ZrH 1.6Th1.08, and PuZrO2MgO (8 wt%) are selected for detailed analysis in a multi-pin transmutation assembly. A coupled model optimized the resulting transmutation fuel elements. The optimization considered three stages of fuel assemblies containing target fuel pins. The first stage optimized four target fuel pins adjacent to the central instrumentation channel. The second stage evaluated a variety of assemblies with multiple target fuel pins and the third stage re-optimized target fuel pins in the second-stage assembly. A PuZrO2MgO (8 wt%) target fuel with a coating of Lu 2O3 resulted in the greatest reduction in curium-244

  4. Iron Corrosion Observations: Pu(VI)-Fe Reduction Studies

    SciTech Connect

    Reed, Donald T.; Swanson, Juliet S.; Richmann, Michael K.; Lucchini, Jean-Francois; Borkowski, Marian

    2012-09-11

    Iron and Pu Reduction: (1) Very different appearances in iron reaction products were noted depending on pH, brine and initial iron phase; (2) Plutonium was associated with the Fe phases; (3) Green rust was often noted at the higher pH; (4) XANES established the green rust to be an Fe2/3 phase with a bromide center; and (5) This green rust phase was linked to Pu as Pu(IV).

  5. Accident tolerant fuels for LWRs: A perspective

    NASA Astrophysics Data System (ADS)

    Zinkle, S. J.; Terrani, K. A.; Gehin, J. C.; Ott, L. J.; Snead, L. L.

    2014-05-01

    The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms.

  6. Size and density of a /sup 242/Pu colloid

    SciTech Connect

    Rundberg, R.S.; Mitchell, A.J.; Torstenfelt, N.B.

    1987-01-01

    The size and density of a /sup 242/Pu colloid has been measured by autocorrelation photon spectrometry. The density of the colloid was determined by ultraspeed centrifugation. From the concentration profiles of /sup 242/Pu in the centrifuged test tubes, a standard sedimentation formula was used to calculate the density; the size of the colloid was known from the light scattering experiments. The determined density of the /sup 242/Pu colloid was unexpectedly low compared to the density of crystalline PuO/sub 2/. 5 refs., 1 tab.

  7. Effect of equilibration time on Pu desorption from goethite

    SciTech Connect

    Wong, Jennifer C.; Zavarin, Mavrik; Begg, James D.; Kersting, Annie B.; Powell, Brian A.

    2015-01-28

    Strongly sorbing ions such as plutonium may become irreversibly bound to mineral surfaces over time implicates near- and far-field transport of Pu. Batch adsorption–desorption data were collected as a function of time and pH to study the surface stability of Pu on goethite. Pu(IV) was adsorbed to goethite over the pH range 4.2 to 6.6 for different periods of time (1, 6, 15, 34 and 116 d). Moreover, following adsorption, Pu was leached from the mineral surface with desferrioxamine B (DFOB), a complexant capable of effectively competing with the goethite surface for Pu. The amount of Pu desorbed from the goethite was found to vary as a function of the adsorption equilibration time, with less Pu removed from the goethite following longer adsorption periods. This effect was most pronounced at low pH. Logarithmic desorption distribution ratios for each adsorption equilibration time were fit to a pH-dependent model. Model slopes decreased between 1 and 116 d adsorption time, indicating that overall Pu(IV) surface stability on goethite surfaces becomes less dependent on pH with greater adsorption equilibration time. The combination of adsorption and desorption kinetic data suggest that non-redox aging processes affect Pu sorption behavior on goethite.

  8. Effect of equilibration time on Pu desorption from goethite

    DOE PAGES

    Wong, Jennifer C.; Zavarin, Mavrik; Begg, James D.; ...

    2015-01-28

    Strongly sorbing ions such as plutonium may become irreversibly bound to mineral surfaces over time implicates near- and far-field transport of Pu. Batch adsorption–desorption data were collected as a function of time and pH to study the surface stability of Pu on goethite. Pu(IV) was adsorbed to goethite over the pH range 4.2 to 6.6 for different periods of time (1, 6, 15, 34 and 116 d). Moreover, following adsorption, Pu was leached from the mineral surface with desferrioxamine B (DFOB), a complexant capable of effectively competing with the goethite surface for Pu. The amount of Pu desorbed from the goethitemore » was found to vary as a function of the adsorption equilibration time, with less Pu removed from the goethite following longer adsorption periods. This effect was most pronounced at low pH. Logarithmic desorption distribution ratios for each adsorption equilibration time were fit to a pH-dependent model. Model slopes decreased between 1 and 116 d adsorption time, indicating that overall Pu(IV) surface stability on goethite surfaces becomes less dependent on pH with greater adsorption equilibration time. The combination of adsorption and desorption kinetic data suggest that non-redox aging processes affect Pu sorption behavior on goethite.« less

  9. Neutron Capture Cross Section of 239Pu

    NASA Astrophysics Data System (ADS)

    Mosby, S.; Arnold, C.; Bredeweg, T. A.; Chyzh, A.; Couture, A.; Henderson, R.; Jandel, M.; Kwan, E.; O'Donnell, J. M.; Rusev, G.; Ullmann, J. L.; Wu, C. Y.

    2014-05-01

    The Detector for Advanced Neutron Capture Experiments (DANCE) has been used to measure the 239Pu(n,γ) cross section from 10 eV to the keV region. Three experimental run conditions were used to characterize the prompt fission γ-ray spectrum across the entire energy regime, measure the cross section in the resolved resonance region, and obtain necessary count rate well into the keV region. The preliminary cross sections are in good agreement with current evaluations from 10 eV to 80 keV.

  10. Simulation of radiation driven fission gas diffusion in UO2, ThO2 and PuO2

    NASA Astrophysics Data System (ADS)

    Cooper, M. W. D.; Stanek, C. R.; Turnbull, J. A.; Uberuaga, B. P.; Andersson, D. A.

    2016-12-01

    Below 1000 K it is thought that fission gas diffusion in nuclear fuel during irradiation occurs through atomic mixing due to radiation damage. Here we present a molecular dynamics (MD) study of Xe, Kr, Th, U, Pu and O diffusion due to irradiation. It is concluded that the ballistic phase does not sufficiently account for the experimentally observed diffusion. Thermal spike simulations are used to confirm that electronic stopping remedies the discrepancy with experiment and the predicted diffusivities lie within the scatter of the experimental data. Our results predict that the diffusion coefficients are ordered such that DO* > DKr* > DXe* > DU*. For all species >98.5% of diffusivity is accounted for by electronic stopping. Fission gas diffusivity was not predicted to vary significantly between ThO2, UO2 and PuO2, indicating that this process would not change greatly for mixed oxide fuels.

  11. Neutron induced fission cross section measurements of 240Pu and 242Pu

    NASA Astrophysics Data System (ADS)

    Belloni, F.; Eykens, R.; Heyse, J.; Matei, C.; Moens, A.; Nolte, R.; Plompen, A. J. M.; Richter, S.; Sibbens, G.; Vanleeuw, D.; Wynants, R.

    2017-09-01

    Accurate neutron induced fission cross section of 240Pu and 242Pu are required in view of making nuclear technology safer and more efficient to meet the upcoming needs for the future generation of nuclear power plants (GEN-IV). The probability for a neutron to induce such reactions figures in the NEA Nuclear Data High Priority Request List [1]. A measurement campaign to determine neutron induced fission cross sections of 240Pu and 242Pu at 2.51 MeV and 14.83 MeV has been carried out at the 3.7 MV Van De Graaff linear accelerator at Physikalisch-Technische Bundesanstalt (PTB) in Braunschweig. Two identical Frisch Grid fission chambers, housing back to back a 238U and a APu target (A = 240 or A = 242), were employed to detect the total fission yield. The targets were molecular plated on 0.25 mm aluminium foils kept at ground potential and the employed gas was P10. The neutron fluence was measured with the proton recoil telescope (T1), which is the German primary standard for neutron fluence measurements. The two measurements were related using a De Pangher long counter and the charge as monitors. The experimental results have an average uncertainty of 3-4% at 2.51 MeV and for 6-8% at 14.81 MeV and have been compared to the data available in literature.

  12. Statistical properties of Pu243 , and Pu242(n,γ) cross section calculation

    DOE PAGES

    Laplace, T. A.; Zeiser, F.; Guttormsen, M.; ...

    2016-01-29

    The level density and γ-ray strength function (γSF) of 243Pu have been measured in the quasicontinuum using the Oslo method. Excited states in 243Pu were populated using the 242Pu(d,p) reaction. The level density closely follows the constant-temperature level density formula for excitation energies above the pairing gap. The γSF displays a double-humped resonance at low energy as also seen in previous investigations of actinide isotopes. The structure is interpreted as the scissors resonance and has a centroid of ωSR = 2.42(5) MeV and a total strength of BSR = 10.1(15) μ2N, which is in excellent agreement with sum-rule estimates. Lastly,more » the measured level density and γSF were used to calculate the 242Pu(n,γ) cross section in a neutron energy range for which there were previously no measured data.« less

  13. Fuel flexible fuel injector

    DOEpatents

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  14. 239Pu(n,2n) 238Pu cross section inferred from IDA calculations and GEANIE measurements

    SciTech Connect

    Chen, H; Ormand, W E; Dietrich, F S

    2000-09-01

    This report presents the latest {sup 239}Pu(n,2n){sup 238}Pu cross sections inferred from calculations performed with the nuclear reaction-modeling code system, IDA, coupled with experimental measurements of partial {gamma}-ray cross sections for incident neutron energies ranging from 5.68 to 17.18 MeV. It is found that the inferred {sup 239}Pu(n,2n){sup 238}Pu cross section peaks at E{sub inc} {approx} 11.4 MeV with a peak value of approximately 326 mb. At E{sub inc} {approx} 14 MeV, the inferred {sup 239}Pu(n,2n){sup 238}Pu cross section is found to be in good agreement with previous radio-chemical measurements by Lockheed. However, the shape of the inferred {sup 239}Pu(n,2n){sup 238}Pu cross section differs significantly from previous evaluations of ENDL, ENDF/B-V and ENDF/B-VI. In our calculations, direct, preequilibrium, and compound reactions are included. Also considered in the modeling are fission and {gamma}-cascade processes in addition to particle emission. The main components of physics adopted and the parameters used in our calculations are discussed. Good agreement of the inferred {sup 239}Pu(n,2n){sup 238}Pu cross sections derived separately from IDA and GNASH calculations is shown. The two inferences provide an estimate of variations in the deduced {sup 239}Pu(n,2n){sup 238}Pu cross section originating from modeling.

  15. 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.

  16. Theoretical investigation on electronic and mechanical properties of ternary actinide (U, Np, Pu) nitrides

    NASA Astrophysics Data System (ADS)

    Zhang, Yu-Juan; Zhou, Zhang-Jian; Lan, Jian-Hui; Bo, Tao; Ge, Chang-Chun; Chai, Zhi-Fang; Shi, Wei-Qun

    2017-09-01

    Actinide mononitrides as a promising advanced nuclear fuel have recently earned much attention. We herein studied the electronic and mechanical properties of the ternary actinide mixed mononitrides A0.5B0.5 N (A, B = U, Np, and Pu) using the density functional theory +U method. It is found that in the studied ternary mixed mononitrides, the 5f electronic states of all actinide atoms maintain the local electronic character and do not overlap with each other. Compared with their corresponding binary mononitrides, the U-N bond becomes more ionic, where the Np-N and Pu-N bonds become more covalent in ternary actinide mixed mononitrides. The mechanical properties (such as bulk and shear moduli, Young's modulus, and Poisson's ratio) of three ternary actinide (U-Pu) mononitrides are found to be similar to that of their corresponding binary actinide mononitrides and thus are expected not to misbehave with actinide mononitrides in respect of mechanics. In addition, all the three ternary actinide mononitrides have no imaginary frequencies in their vibration curves and correspondingly satisfy the stability criteria for elastic constants of tetragonal structures.

  17. Incorporation of excess weapons material into the IFR fuel cycle

    SciTech Connect

    Hannum, W.H.; Wade, D.C.

    1993-09-01

    The Integral Fast Reactor (IFR) provides both a diversion resistant closed fuel cycle for commercial power generation and a means of addressing safeguards concerns related to excess nuclear weapons material. Little head-end processing and handling of dismantled warhead materials is required to convert excess weapons plutonium (Pu) to IFR fuel and a modest degree of proliferation protection is available immediately by alloying weapons Pu to an IFR fuel composition. Denaturing similar to that of spent fuel is obtained by short cycle (e.g. 45 day) use in an IFR reactor, by mixing which IFR recycle fuel, or by alloying with other spent fuel constituents. Any of these permanent denaturings could be implemented as soon as an operating IFR and/or an IFR recycle capability of reasonable scale is available. The initial Pu charge generated from weapons excess Pu can then be used as a permanent denatured catalyst, enabling the IFR to efficiently and economically generate power with only a natural or depleted uranium feed. The Pu is thereafter permanently safeguarded until consumed, with essentially none going to a waste repository.

  18. Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Edwards, Geoffrey W R; Priest, Nicholas D

    2014-11-01

    The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low

  19. On the multi-reference nature of plutonium oxides: PuO22+, PuO2, PuO3 and PuO2(OH)2

    NASA Astrophysics Data System (ADS)

    Boguslawski, Katharina; Réal, Florent; Tecmer, Paweł; Duperrouzel, Corinne; Gomes, André Severo Pereira; Legeza, Örs; Ayers, Paul W.; Vallet, Valérie

    Actinide-containing complexes present formidable challenges for electronic structure methods due to the large number of degenerate or quasi-degenerate electronic states arising from partially occupied 5f and 6d shells. Conventional multi-reference methods can treat active spaces that are often at the upper limit of what is required for a proper treatment of species with complex electronic structures, leaving no room for verifying their suitability. In this work we address the issue of properly defining the active spaces in such calculations, and introduce a protocol to determine optimal active spaces based on the use of the Density Matrix Renormalization Group algorithm and concepts of quantum information theory. We apply the protocol to elucidate the electronic structure and bonding mechanism of volatile plutonium oxides (PuO$_3$ and PuO$_2$(OH)$_2$), species associated with nuclear safety issues for which little is known about the electronic structure and energetics. We show how, within a scalar relativistic framework, orbital-pair correlations can be used to guide the definition of optimal active spaces which provide an accurate description of static/non-dynamic electron correlation, as well as to analyse the chemical bonding beyond a simple orbital model. From this bonding analysis we are able to show that the addition of oxo- or hydroxo-groups to the plutonium dioxide species considerably changes the pi-bonding mechanism with respect to the bare triatomics, resulting in bent structures with considerable multi-reference character.

  20. 239Pu Prompt Fission Neutron Spectra Impact on a Set of Criticality and Experimental Reactor Benchmarks

    NASA Astrophysics Data System (ADS)

    Peneliau, Y.; Litaize, O.; Archier, P.; De Saint Jean, C.

    2014-04-01

    A large set of nuclear data are investigated to improve the calculation predictions of the new neutron transport simulation codes. With the next generation of nuclear power plants (GEN IV projects), one expects to reduce the calculated uncertainties which are mainly coming from nuclear data and are still very important, before taking into account integral information in the adjustment process. In France, future nuclear power plant concepts will probably use MOX fuel, either in Sodium Fast Reactors or in Gas Cooled Fast Reactors. Consequently, the knowledge of 239Pu cross sections and other nuclear data is crucial issue in order to reduce these sources of uncertainty. The Prompt Fission Neutron Spectra (PFNS) for 239Pu are part of these relevant data (an IAEA working group is even dedicated to PFNS) and the work presented here deals with this particular topic. The main international data files (i.e. JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0, BRC-2009) have been considered and compared with two different spectra, coming from the works of Maslov and Kornilov respectively. The spectra are first compared by calculating their mathematical moments in order to characterize them. Then, a reference calculation using the whole JEFF-3.1.1 evaluation file is performed and compared with another calculation performed with a new evaluation file, in which the data block containing the fission spectra (MF=5, MT=18) is replaced by the investigated spectra (one for each evaluation). A set of benchmarks is used to analyze the effects of PFNS, covering criticality cases and mock-up cases in various neutron flux spectra (thermal, intermediate, and fast flux spectra). Data coming from many ICSBEP experiments are used (PU-SOL-THERM, PU-MET-FAST, PU-MET-INTER and PU-MET-MIXED) and French mock-up experiments are also investigated (EOLE for thermal neutron flux spectrum and MASURCA for fast neutron flux spectrum). This study shows that many experiments and neutron parameters are very sensitive to

  1. Ab Initio Enhanced calphad Modeling of Actinide-Rich Nuclear Fuels

    SciTech Connect

    Morgan, Dane; Yang, Yong Austin

    2013-10-28

    The process of fuel recycling is central to the Advanced Fuel Cycle Initiative (AFCI), where plutonium and the minor actinides (MA) Am, Np, and Cm are extracted from spent fuel and fabricated into new fuel for a fast reactor. Metallic alloys of U-Pu-Zr-MA are leading candidates for fast reactor fuels and are the current basis for fast spectrum metal fuels in a fully recycled closed fuel cycle. Safe and optimal use of these fuels will require knowledge of their multicomponent phase stability and thermodynamics (Gibbs free energies). In additional to their use as nuclear fuels, U-Pu-Zr-MA contain elements and alloy phases that pose fundamental questions about electronic structure and energetics at the forefront of modern many-body electron theory. This project will validate state-of-the-art electronic structure approaches for these alloys and use the resulting energetics to model U-Pu-Zr-MA phase stability. In order to keep the work scope practical, researchers will focus on only U-Pu-Zr-{Np,Am}, leaving Cm for later study. The overall objectives of this project are to: Provide a thermodynamic model for U-Pu-Zr-MA for improving and controlling reactor fuels; and, Develop and validate an ab initio approach for predicting actinide alloy energetics for thermodynamic modeling.

  2. Determination of plutonium content in TRR spent fuel by nondestructive neutron counting

    NASA Astrophysics Data System (ADS)

    Chen, Yen-Fu; Sheu, Rong-Jiun; Chiao, Ling-Huan; Yuan, Ming-Chen; Jiang, Shiang-Huei

    2010-07-01

    For the nuclear safeguard purpose, this work aims to nondestructively determine the plutonium content in the Taiwan Research Reactor (TRR) spent fuel rods in the storage pool before the stabilization process, which transforms the metal spent fuel rods into oxide powder. A SPent-fuel-Neutron-Counter (SPNC) system was designed and constructed to carry out underwater scan measurements of neutrons emitting from the spent fuel rod, from which the 240Pu mass in the fuel rod will be determined. The SAS2 H control module of the SCALE 5.1 code package was applied to calculate the 240Pu-to-Pu mass ratio in the TRR spent fuel rod according to the given power history. This paper presents the methodology and design of our detector system as well as the measurements of four TRR spent fuel rods in the storage pool and the comparison of the measured results with the facility declared values.

  3. Is Octavalent Pu(VIII) Possible? Mapping the Plutonium Oxyfluoride Series PuO(n)F(8-2n) (n = 0-4).

    PubMed

    Huang, Wei; Pyykkö, Pekka; Li, Jun

    2015-09-08

    While the oxidation state Pu(VIII) is shown to be less stable than Pu(V) in the PuO4 molecule, it is not clear if the more electronegative fluorine can help to stabilize Pu(VIII). Our calculations on PuO(n)F(8-2n) (n = 0-4) molecules notably confirm that PuO2F4 has both (1)D(4h) and (5)C(2v) minima with the oxidation states Pu(VIII) and Pu(V), respectively, with the latter having lower energy. The hybrid-DFT, CCSD(T), and CASSCF methods all give the same result. The results conform to a superoxide ligand when n ≥ 2. PuF8 in a (1)O(h) state can decompose to PuF6 and F2, and PuOF6 in a (1)C(2v) state also can break down to PuF6 and 1/2 O2. The Pu(VIII) anion PuO2F5(-) does have a D(5h) minimum, which also lies above a (5)C(2v) Pu(V) peroxide structure. However, the energy differences between the different minima are not large, indicating that metastable species with oxidation states higher than Pu(V) cannot be completely excluded.

  4. A comparative study of nitride purity and Am fabrication losses in PuN materials by the powder and internal gelation production routes

    NASA Astrophysics Data System (ADS)

    Hedberg, Marcus; Ekberg, Christian

    2016-12-01

    Fabrication of plutonium containing fuels through the internal gelation method has mostly been studied in mixed metal systems such as U, Pu or Zr,Pu. In this work production of undiluted PuN has been performed by carbothermal reduction on both oxide powder and Pu microspheres produced by the internal gelation method. Nitride purities reached using the different methods have been studied together with final densities achieved during pellet fabrication as well as losses of ingrown Am during the different production steps. Formation of Pu microspheres was successfully performed using the internal gelation method, although extensive microsphere fracturing occurred during thermal treatment. Final densities of PuN pellets produced by cold pressing and sintering reached 70-80% of theoretical density. Am losses during the carbothermal reduction step was on average about 3.7%. After sintering about 11% of Am was lost in total through the entire production process if sintering in N2 + 5% H2 atmosphere while about 50% of the Am in total was lost when using Ar as sintering atmosphere.

  5. Isotopic Pu, Am and Cm signatures in environmental samples contaminated by the Fukushima Dai-ichi Nuclear Power Plant accident.

    PubMed

    Yamamoto, M; Sakaguchi, A; Ochiai, S; Takada, T; Hamataka, K; Murakami, T; Nagao, S

    2014-06-01

    Dust samples from the sides of roads (black substances) have been collected together with litter and soil samples at more than 100 sites contaminated heavily in the 20-km exclusion zones around Fukushima Dai-ichi Nuclear Power Plant (FDNPP) (Minamisoma City, and Namie, Futaba and Okuma Towns), in Iitate Village located from 25 to 45 km northwest of the plant and in southern areas from the plant. Isotopes of Pu, Am and Cm have been measured in the samples to evaluate their total releases into the environment from the FDNPP and to get the isotopic compositions among these nuclides. For black substances and litter samples, in addition to Pu isotopes, (241)Am, (242)Cm and (243,244)Cm were determined for most of samples examined, while for soil samples, only Pu isotopes were determined. The results provided a coherent data set on (239,240)Pu inventories and isotopic composition among these transuranic nuclides. When these activity ratios were compared with those for fuel core inventories in the FDNPP accident estimated by a group at JAEA, except (239,240)Pu/(137)Cs activity ratios, fairly good agreements were found, indicating that transuranic nuclides, probably in the forms of fine particles, were released into the environment without their large fractionations. The obtained data may lead to more accurate information about the on-site situation (e.g., burn-up, conditions of fuel during the release phase, etc.), which would be difficult to get otherwise, and more detailed information on the dispersion and deposition processes of transuranic nuclides and the behavior of these nuclides in the environment.

  6. Ostwald Ripening and Its Effect on PuO2 Particle Size in Hanford Tank Waste

    SciTech Connect

    Delegard, Calvin H.

    2011-09-29

    Between 1944 and 1989, the Hanford Site produced 60 percent (54.5 metric tons) of the United States weapons plutonium and produced an additional 12.9 metric tons of fuels-grade plutonium. High activity wastes, including plutonium lost from the separations processes used to isolate the plutonium, were discharged to underground storage tanks during these operations. Plutonium in the Hanford tank farms is estimated to be {approx}700 kg but may be up to {approx}1000 kg. Despite these apparent large quantities, the average plutonium concentration in the {approx}200 million liter tank waste volume is only about 0.003 grams per liter ({approx}0.0002 wt%). The plutonium is largely associated with low solubility metal hydroxide/oxide sludges where its low concentration and intimate mixture with neutron-absorbing elements (e.g., iron) are credited in nuclear criticality safety. However, concerns have been expressed that plutonium, in the form of plutonium hydrous oxide, PuO{sub 2} {center_dot} xH{sub 2}O, could undergo sufficient crystal growth through Ostwald ripening in the alkaline tank waste to potentially be separable from neutron absorbing constituents by settling or sedimentation. It was found that plutonium that entered the alkaline tank waste by precipitation through neutralization from acid solution is initially present as 2- to 3-nm (0.002- to 0.003-{mu}m) scale PuO{sub 2} {center_dot} xH{sub 2}O crystallite particles and grows from that point at exceedingly slow rates, posing no risk to physical segregation. These conclusions are reached by both general considerations of Ostwald ripening and specific observations of the behaviors of PuO{sub 2} and PuO{sub 2} {center_dot} xH{sub 2}O upon aging in alkaline solution.

  7. PU.1 silencing leads to terminal differentiation of erythroleukemia cells

    SciTech Connect

    Atar, Orna; Levi, Ben-Zion . E-mail: blevi@technion.ac.il

    2005-04-22

    The transcription factor PU.1 plays a central role in development and differentiation of hematopoietic cells. Evidence from PU.1 knockout mice indicates a pivotal role for PU.1 in myeloid lineage and B-lymphocyte development. In addition, PU.1 is a key player in the development of Friend erythroleukemia disease, which is characterized by proliferation and differentiation arrest of proerythrocytes. To study the role of PU.1 in erythroleukemia, we have used murine erythroleukemia cells, isolated from Friend virus-infected mice. Expression of PU.1 small interfering RNA in these cells led to significant inhibition of PU.1 levels. This was accompanied by inhibition of proliferation and restoration in the ability of the proerythroblastic cells to produce hemoglobin, i.e., reversion of the leukemic phenotype. The data suggest that overexpression of PU.1 gene is the immediate cause for maintaining the leukemic phenotype of the disease by retaining the self-renewal capacity of transformed erythroblastic cells and by blocking the terminal differentiation program towards erythrocytes.

  8. Synchrotron-Radiation-based Investigationsof the Electronic Structure of Pu

    SciTech Connect

    Tobin, J; Chung, B; Terry, J; Schulze, R; Farr, J; Heinzelman, K; Rotenberg, E; Shuh, D

    2004-09-27

    Synchrotron radiation from the Advanced Light Source has been used to investigate the electronic structure of {alpha}-Pu and {delta}-Pu. Measurements include core level and valence band photoelectron spectroscopy, Resonant Photoelectron Spectroscopy (REPES), and X-ray Absorption Spectroscopy (XAS).

  9. 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.

  10. Global distribution of Pu isotopes and 237Np.

    PubMed

    Kelley, J M; Bond, L A; Beasley, T M

    1999-09-30

    Inventories and compositions of Pu isotopes and 237Np in archived soil samples collected in the 1970s from 54 locations around the world were determined to provide regional baselines for recognizing possible future environmental inputs of non-fallout Pu and Np. As sample sizes used in this work were small (typically 1 g), inhomogeneities in Pu and Np concentrations were easily recognizable and, as a result, we were able to determine that atypical debris in South America, from French testing in the South Pacific, is more widely and uniformly distributed than previously supposed. From our results we conclude that fallout 237Np/239Pu atom ratios are generally lower in the Southern Hemisphere (approximately 0.35) than in the Northern Hemisphere (approximately 0.47.) Moreover, 237Np/239Pu atom ratios are more device-dependent, hence more variable, than counterpart 240Pu/239Pu atom ratios. Given predictable trends caused by sample inhomogeneities, with only two exceptions, the Pu results of this work are entirely consistent with (and in several instances improve on) results previously reported for these same samples. However, unlike earlier interpretations used to explain these results, we recommend that fallout isotopic signatures be represented by mixing lines, rather than averages, to better reflect regional variations of stratospheric fallout inventories relative to tropospheric fallout inventories, and provide the theoretical basis for doing so. Finally, the Np results of this work constitute one of the largest single compilations of such data reported to date.

  11. Performance and design considerations in metal fueled cores. [LMFBR

    SciTech Connect

    Orechwa, Y.; Khalil, H.; Turski, R.B.

    1984-01-01

    To focus future metal fuel development requirements a study was performed to quantify the relationship between some critical core design parameters. The fuel studied was U-Pu-Zr alloy. Of interest are performance parameters, such as peak Pu enrichment, burnup swing, fast fluence, breeding ratio, and their relation to core parameters such as reactor size, degree of core heterogeneity, pin diameter, and linear heat rating. These performance parameters, while numericaly different from those of ceramic fuels, were found to exhibit the same qualitative dependence on the key design variables.

  12. Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel [1]. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as PuFP. The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the PuFP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  13. Technical overview: CANDU MOX fuel dual irradiation experiment

    SciTech Connect

    Dimayuga, F.C.; M.R. Floyd, M.R.; Schankula, M.H.; Sullivan, J.D.

    1996-02-01

    This Technical Overview describes: the technical objectives and rational for the choice of MOX fuel fabrication parameters that are to be investigated; the pre-irradiation fuel characterization plan; the NRU irradiation plan; the post-irradiation examination plan; and a summary of the evaluations that can be extracted from the Parallex data. This Technical Overview is based on the 37-element reference CANDU MOX fuel design established in the 1994 Pu Dispositioning Study. An extension to this study is currently underway, aimed at increasing the Pu disposition rates of the mission. The results of this new study will likely specify a higher Pu loading for the CANDU MOX fuel. If confirmed, this Technical Overview document will be revised and the Parallex test matrix could be modified accordingly.

  14. ANALYSIS AND EXAMINATION OF MOX FUEL FROM NONPROLIFERATION PROGRAMS

    SciTech Connect

    McCoy, Kevin; Machut, Dr McLean; Morris, Robert Noel; Blanpain, Patrick; Hemrick, James Gordon

    2013-01-01

    The U.S. Department of Energy has decided to dispose of a portion of the nation s surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. Four lead assemblies were manufactured and irradiated to a maximum fuel rod burnup of 47.3 MWd/kg heavy metal. This was the first commercial irradiation of MOX fuel with a 240Pu/239Pu ratio of less than 0.10. Five fuel rods with varying burnups and plutonium contents were selected from one of the assemblies and shipped to Oak Ridge National Laboratory for hot cell examination. The performance of the rods was analyzed with AREVA s next-generation GALILEO code. The results of the analysis confirmed that the fuel rods had performed safely and predictably, and that GALILEO is applicable to MOX fuel with a low 240Pu/239Pu ratio as well as to standard MOX. The results are presented and compared to the GALILEO database. In addition, the fuel cladding was tested to confirm that traces of gallium in the fuel pellets had not affected the mechanical properties of the cladding. The irradiated cladding was found to remain ductile at both room temperature and 350 C for both the axial and circumferential directions.

  15. Criticality safety evaluation report for FFTF 42% fuel assemblies

    SciTech Connect

    Richard, R.F.

    1997-10-28

    An FFTF tritium/isotope production mission will require a new fuel supply. The reference design core will use a mixed oxide fuel nominally enriched to 40 wt% Pu. This enrichment is significantly higher than that of the standard Driver Fuel Assemblies used in past operations. Consequently, criticality safety for handling and storage of this fuel must be addressed. The purpose of this document is to begin the process by determining the minimum critical number for these new fuel assemblies in water, sodium and air. This analysis is preliminary and further work can be done to refine the results reported here. Analysis was initially done using 45 wt 5 PuO. Additionally, a preliminary assessment is done concerning storage of these fuel assemblies in Interim Decay Storage (IDS), Fuel Storage Facility (FSF), and Core Component Containers/Interim Storage Casks (CCC/ISC).

  16. Resolving global versus local/regional Pu sources in the environment using sector ICP-MS

    USGS Publications Warehouse

    Ketterer, M.E.; Hafer, K.M.; Link, C.L.; Kolwaite, D.; Wilson, Jim; Mietelski, J.W.

    2004-01-01

    Sector inductively coupled plasma mass spectrometry is a versatile method for the determination of plutonium activities and isotopic compositions in samples containing this element at fallout levels. Typical detection limits for 239+240Pu are 0.1, 0.02 and 0.002 Bq kg -1Pu for samples sizes of 0.5 g, 3 g, and 50 g of soil, respectively. The application of sector ICP-MS-based Pu determinations is demonstrated in studies in sediment chronology, soil Pu inventory and depth distribution, and the provenance of global fallout versus local or regional Pu sources. A sediment core collected from Sloans Lake (Denver, Colorado, USA) exhibits very similar 137Cs and 239+240Pu activity profiles; 240Pu/239Pu atom ratios indicate possible small influences from the Nevada Test Site and/or the Rocky Flats Environmental Technology Site. An undisturbed soil profile from Lockett Meadow (Flagstaff, Arizona, USA) exhibits an exponential decrease in 239+240Pu activity versus depth; 240Pu/239Pu in the top 3 cm is slightly lower than the global fallout range of 0.180 ?? 0.014 due to possible regional influence of Nevada Test Site fallout. The 239??240Pu inventory at Lockett Meadow is 56 ?? 4 Bq m-2, consistent with Northern Hemisphere mid-latitude fallout. Archived NdF3 sources, prepared from Polish soils, demonstrate that substantial 239+240Pu from the 1986 Chernobyl disaster has been deposited in north eastern regions of Poland; compared to global fallout, Chernobyl Pu exhibits higher abundances of 240Pu and 241Pu. The ratios 240Pu/239pu and 241Pu/239Pu co-vary and range from 0.186-0.348 and 0.0029-0.0412, respectively, in forest soils (241Pu/239Pu = 0.2407??[240Pu/239Pu] - 0.0413; r2 = 0.9924). ?? The Royal Society of Chemistry 2004.

  17. Fission xenon from extinct Pu-244 in 14,301.

    NASA Technical Reports Server (NTRS)

    Drozd, R.; Hohenberg, C. M.; Ragan, D.

    1972-01-01

    Xenon extracted in step-wise heating of lunar breccia 14,301 contains a fission-like component in excess of that attributable to uranium decay during the age of the solar system. There seems to be no adequate source for this component other than Pu-244. Verification that this component is in fact due to the spontaneous fission of extinct Pu-244 comes from the derived spectrum which is similar to that observed from artificially produced Pu-244. It thus appears that Pu-244 was extant at the time lunar crustal material cooled sufficiently to arrest the thermal diffusion of xenon. Subsequent history has apparently maintained the isotopic integrity of plutonium fission xenon. Of major importance are details of the storage itself. Either the fission component is the result of in situ fission of Pu-244 and subsequent storage in 14,301 material, or the fission xenon was stored in an intermediate reservoir before incorporation into 14,301.

  18. Laboratory studies of actinide partitioning relevant to 244Pu chronometry

    NASA Technical Reports Server (NTRS)

    Benjamin, T.; Heuser, W. R.; Burnett, D. S.

    1978-01-01

    Actinide partitioning and light lanthanide fractionation have been studied to gain an understanding of Pu chemistry under meteoritic and lunar conditions. The goal of the study was to identify conditions and samples from which chronological information can be retrieved. The laboratory investigations involved particle track radiography of the crystal/liquid partitioning of Th, U and Pu among diopsidic clinopyroxene, whitlockite and liquid. It is found that trivalent Pu plays an important role in partitioning for lunar and most meteoritic conditions. The use of Pu/Nd for relative age assessments is supported to some extent by the investigations; samples with unfractionated U, Th and Nd abundances (relative to average solar system values) may be suitable for Pu chronometry.

  19. Laboratory studies of actinide partitioning relevant to 244Pu chronometry

    NASA Technical Reports Server (NTRS)

    Benjamin, T.; Heuser, W. R.; Burnett, D. S.

    1978-01-01

    Actinide partitioning and light lanthanide fractionation have been studied to gain an understanding of Pu chemistry under meteoritic and lunar conditions. The goal of the study was to identify conditions and samples from which chronological information can be retrieved. The laboratory investigations involved particle track radiography of the crystal/liquid partitioning of Th, U and Pu among diopsidic clinopyroxene, whitlockite and liquid. It is found that trivalent Pu plays an important role in partitioning for lunar and most meteoritic conditions. The use of Pu/Nd for relative age assessments is supported to some extent by the investigations; samples with unfractionated U, Th and Nd abundances (relative to average solar system values) may be suitable for Pu chronometry.

  20. Method for estimating the systemic burden of Pu from urinalyses

    SciTech Connect

    Leggett, R.W.; Eckerman, K.F.

    1987-03-01

    It is generally agreed that Langham's model for urinary excretion of Pu substantially overestimates the systemic burden several years after exposure. Improved estimates can be derived from information obtained since the development of that model, including comparative urine and autopsy data for occupationally exposed persons; reanalyzed and updated data for human subjects injected with Pu; and a large body of general physiological and Pu-specific information on the processes governing the behavior of Pu in the body. We examine modeling approaches based on each of these sets of information and show that the three approaches yield fairly consistent estimates of the urinary excretion rate over three decades after contamination of blood. Estimates from the various approaches are unified to obtain a single set of predicted urinary excretion rates that, in effect, is based on all three bodies of information. A simple method is described for using these excretion rates to estimate intakes and systemic burdens of Pu.

  1. Thermodynamic evaluation of the NaCl-MgCl 2-UCl 3-PuCl 3 system

    NASA Astrophysics Data System (ADS)

    Beneš, O.; Konings, R. J. M.

    2008-04-01

    A full thermodynamic description of the quaternary NaCl-MgCl 2-UCl 3-PuCl 3 system, a Molten Salt Fast Breeder Fuel, is presented. The binary phase diagrams have been assessed in this study and the data were used to extrapolate the higher order systems. To optimize the excess parameters of the liquid phase the modified quasi chemical model has been used, while for the solid solution the classical polynomial model has been applied. From the obtained results a possible fuel composition for the Molten Salt Reactor has been evaluated.

  2. Oxidation states, geometries, and electronic structures of plutonium tetroxide PuO4 isomers: is octavalent Pu viable?

    PubMed

    Huang, Wei; Xu, Wen-Hua; Su, Jing; Schwarz, W H E; Li, Jun

    2013-12-16

    In neutral chemical compounds, the highest known oxidation state of all elements in the Periodic Table is +VIII. While PuO4 is viewed as an exotic Pu(+VIII) complex, we have shown here that no stable electronic homologue of octavalent RuO4 and OsO4 exists for PuO4, even though Pu has the same number of eight valence electrons as Ru and Os. Using quantum chemical approaches at the levels of quasi-relativistic DFT, MP2, CCSD(T), and CASPT2, we find the ground state of PuO4 as a quintet (5)C2v-(PuO2)(+)(O2)(-) complex with the leading valence configuration of an (f(3))plutonyl(V) unit, loosely coupled to a superoxido (π*(3))O2(-) ligand. This stable isomer is likely detectable as a transient species, while the previously suggested planar (1)D4h-Pu(VIII)O4 isomer is only metastable. Through electronic structure analyses, the bonding and the oxidation states are explained and rationalized. We have predicted the characteristics of the electronic and vibrational spectra to assist future experimental identification of (PuO2)(+)(O2)(-) by IR, UV-vis, and ionization spectroscopy.

  3. Distinguishing Pu Metal from Pu Oxide and Determining alpha-ratio using Fast Neutron Counting

    SciTech Connect

    Verbeke, J. M.; Chapline, G. F.; Nakae, L. F.; Prasad, M. K.; Sheets, S. A.; Snyderman, N. J.

    2015-01-07

    We describe a new method for determining the ratio of the rate of (α, n) source neutrons to the rate of spontaneous fission neutrons, the so called α-ratio. This method is made possible by fast neutron counting with liquid scintillator detectors, which can determine the shape of the fast neutron spectrum. The method utilizes the spectral difference between fission spectrum neutrons from Pu metal and the spectrum of (α, n) neutrons from PuO2. Our method is a generalization of the Cifarelli-Hage method for determining keff for fissile assemblies, and also simultaneously determines keff along with the α-ratio.

  4. Microscopic Calculations of 240Pu Fission

    SciTech Connect

    Younes, W; Gogny, D

    2007-09-11

    Hartree-Fock-Bogoliubov calculations have been performed with the Gogny finite-range effective interaction for {sup 240}Pu out to scission, using a new code developed at LLNL. A first set of calculations was performed with constrained quadrupole moment along the path of most probable fission, assuming axial symmetry but allowing for the spontaneous breaking of reflection symmetry of the nucleus. At a quadrupole moment of 345 b, the nucleus was found to spontaneously scission into two fragments. A second set of calculations, with all nuclear moments up to hexadecapole constrained, was performed to approach the scission configuration in a controlled manner. Calculated energies, moments, and representative plots of the total nuclear density are shown. The present calculations serve as a proof-of-principle, a blueprint, and starting-point solutions for a planned series of more comprehensive calculations to map out a large set of scission configurations, and the associated fission-fragment properties.

  5. LLNL Workshop on TEM of Pu

    SciTech Connect

    King, W.E.

    1996-09-10

    On Sept. 10, 1996, LLNL hosted a workshop aimed at answering the question: Is it possible to carry out transmission electron microscopy (TEM) on plutonium metal in an electron microscope located outside the LLNL plutonium facility. The workshop focused on evaluation of a proposed plan for Pu microscopy both from a technical and environment, health, and safety point of view. After review and modification of the plan, workshop participants unanimously concluded that: (1) the technical plan is sound, (2) this technical plan, including a proposal for a new TEM, provides significant improvements and unique capabilities compared with the effort at LANL and is therefore complementary, (3) there is no significant environment, health, and safety obstacle to this plan.

  6. Preparation of Pu{sup 239} sources

    SciTech Connect

    Holcomb, H.P.

    1988-08-05

    The Separations Technology Laboratory has prepared four sources to be used for calibrating a waste assay system (Passive/Active Neutron Assay) in Building 724-8G (Burial Ground). The four sources contain 0.5, 0.1, 0.05, and 0.01 grams Pu{sup 239}, respectively. The sources were prepared using aliquots from a single solution provided by the Quality Control (QC) group of Laboratories Department. The solution contained weapons-grade plutonium dissolved in nitric acid. Final solution acidity was 3M. Coulometry had been used to obtain a total plutonium content per unit volume. The weight percent of the plutonium isotopes present was obtained via mass spectrometry.

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

    NASA Astrophysics Data System (ADS)

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

    2013-06-01

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

  8. Some neutron and gamma radiation characteristics of plutonium cermet fuel for isotopic power sources

    NASA Technical Reports Server (NTRS)

    Neff, R. A.; Anderson, M. E.; Campbell, A. R.; Haas, F. X.

    1972-01-01

    Gamma and neutron measurements on various types of plutonium sources are presented in order to show the effects of O-17, O-18 F-19, Pu-236, age of the fuel, and size of the source on the gamma and neutron spectra. Analysis of the radiation measurements shows that fluorine is the main contributor to the neutron yields from present plutonium-molybdenum cermet fuel, while both fluorine and Pu-236 daughters contribute significantly to the gamma ray intensities.

  9. Multirecycling of Plutonium from LMFBR Blanket in Standard PWRs Loaded with MOX Fuel

    SciTech Connect

    Sonat Sen; Gilles Youinou

    2013-02-01

    It is now well-known that, from a physics standpoint, Pu, or even TRU (i.e. Pu+M.A.), originating from LEU fuel irradiated in PWRs can be multirecycled also in PWRs using MOX fuel. However, the degradation of the isotopic composition during irradiation necessitates using enriched U in conjunction with the MOX fuel either homogeneously or heterogeneously to maintain the Pu (or TRU) content at a level allowing safe operation of the reactor, i.e. below about 10%. The study is related to another possible utilization of the excess Pu produced in the blanket of a LMFBR, namely in a PWR(MOX). In this case the more Pu is bred in the LMFBR, the more PWR(MOX) it can sustain. The important difference between the Pu coming from the blanket of a LMFBR and that coming from a PWR(LEU) is its isotopic composition. The first one contains about 95% of fissile isotopes whereas the second one contains only about 65% of fissile isotopes. As it will be shown later, this difference allows the PWR fed by Pu from the LMFBR blanket to operate with natural U instead of enriched U when it is fed by Pu from PWR(LEU)

  10. Postirradiation examination of the HT9 clad fuel test X425 at 2.9% burnup

    SciTech Connect

    Pahl, R G; Beck, W N; Sanecki, J E

    1987-11-01

    The X425 experiment was the first EBR-II subassembly to be irradiated with U-Pu-Zr metallic fuel clad in the HT9 alloy. This report summarizes our initial postirradiation examination of selected elements from X425 at 2.9% peak burnup. Fuel microstructure, swelling behavior, fission gas release, and fuel/clad chemical interaction are discussed.

  11. Next-generation purex flowsheets with acetohydroxamic acid as complexant for FBR and thermal-fuel reprocessing

    SciTech Connect

    Kumar, Shekhar; Koganti, S.B.

    2008-07-01

    Acetohydroxamic acid (AHA) is a novel complexant for recycle of nuclear-fuel materials. It can be used in ordinary centrifugal extractors, eliminating the need for electro-redox equipment or complex maintenance requirements in a remotely maintained hot cell. In this work, the effect of AHA on Pu(IV) distribution ratios in 30% TBP system was quantified, modeled, and integrated in SIMPSEX code. Two sets of batch experiments involving macro Pu concentrations (conducted at IGCAR) and one high-Pu flowsheet (literature) were simulated for AHA based U-Pu separation. Based on the simulation and validation results, AHA based next-generation reprocessing flowsheets are proposed for co-processing based FBR and thermal-fuel reprocessing as well as evaporator-less macro-level Pu concentration process required for MOX fuel fabrication. Utilization of AHA results in significant simplification in plant design and simpler technology implementations with significant cost savings. (authors)

  12. Properties measurements of (U{sub 0.7}Pu{sub 0.3})O{sub 2-x} in PO{sub 2}-controlled atmosphere

    SciTech Connect

    Kato, M.; Murakami, T.; Sunaoshi, T.; Nelson, A.T.; McClellan, K.J.

    2013-07-01

    The investigation of physical properties of uranium and plutonium mixed oxide (MOX) fuels is important for the development of fast reactor fuels. It is well known that MOX is a nonstoichiometric oxide, and the physical properties change drastically with the Oxygen-to-Metal (O/M) ratio. A control technique for O/M ratio was established for measurements of high temperature properties of uranium and plutonium mixed oxide fuels. Sintering behavior, thermal expansion and O/M change of (U{sub 0.7}Pu{sub 0.3})O{sub 2.00} and (U{sub 0.7}Pu{sub 0.3})O{sub 1.99} were investigated in PO{sub 2}-controlled atmosphere which was controlled by H{sub 2}/H{sub 2}O gas system. Sintering behavior changed drastically with O/M ratio, and shrinkage of (U{sub 0.7}Pu{sub 0.3})O{sub 2.00} was faster and more advanced at lower temperatures as compared with (U{sub 0.7}Pu{sub 0.3})O{sub 1.99}. Thermal expansion was observed to be slightly increased with decreasing O/M ratio. (authors)

  13. An Investigation of the Use of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in Pressurized Water Reactors

    SciTech Connect

    Gentry, Cole A; Godfrey, Andrew T; Terrani, Kurt A; Gehin, Jess C; Powers, Jeffrey J; Maldonado, G Ivan

    2014-01-01

    An investigation of the utilization of TRistructural- ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized water reactors is presented. Though numerous studies have evaluated the burning of transuranic isotopes in light water reactors (LWRs), this work differentiates itself by employing Pu/Np-loaded TRISO particles embedded within a silicon carbide (SiC) matrix and formed into pellets, constituting the fully ceramic microencapsulated (FCM) fuel concept that can be loaded into standard LWR fuel element cladding. This approach provides the capability of Pu/Np burning and, by virtue of the multibarrier TRISO particle design and SiC matrix properties, will allow for greater burnup of Pu/Np material, plus improved fuel reliability and thermal performance. In this study, a variety of heterogeneous assembly layouts, which utilize a mix of FCM rods and typical UO2 rods, and core loading patterns were analyzed to demonstrate the neutronic feasibility of Pu/Np-loaded TRISO fuel. The assembly and core designs herein reported are not fully optimized and require fine-tuning to flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, Pu/Np-loaded TRISO fuel and core designs that are capable of balancing Pu/Np production and destruction can be designed within the standard constraints for thermal and reactivity performance in pressurized water reactors.

  14. 242Pu absolute neutron-capture cross section measurement

    NASA Astrophysics Data System (ADS)

    Buckner, M. Q.; Wu, C. Y.; Henderson, R. A.; Bucher, B.; Chyzh, A.; Bredeweg, T. A.; Baramsai, B.; Couture, A.; Jandel, M.; Mosby, S.; O'Donnell, J. M.; Ullmann, J. L.

    2017-09-01

    The absolute neutron-capture cross section of 242Pu was measured at the Los Alamos Neutron Science Center using the Detector for Advanced Neutron-Capture Experiments array along with a compact parallel-plate avalanche counter for fission-fragment detection. During target fabrication, a small amount of 239Pu was added to the active target so that the absolute scale of the 242Pu(n,γ) cross section could be set according to the known 239Pu(n,f) resonance at En,R = 7.83 eV. The relative scale of the 242Pu(n,γ) cross section covers four orders of magnitude for incident neutron energies from thermal to ≈ 40 keV. The cross section reported in ENDF/B-VII.1 for the 242Pu(n,γ) En,R = 2.68 eV resonance was found to be 2.4% lower than the new absolute 242Pu(n,γ) cross section.

  15. Determination of alpha-emitting Pu isotopes in environmental samples.

    PubMed

    Vioque, I; Manjón, G; García-Tenorio, R; El-Daoushy, F

    2002-04-01

    This paper presents an improved radiochemical procedure for the determination of alpha-emitting Pu isotopes in environmental samples (soils, sediments, vegetation) by alpha-particle spectrometry. Quantitative Pu recovery yields were obtained (average 60%), 0.1 mBq being the average minimum detectable activity by the complete technique. Special efforts were made to ensure the removal of traces of different natural alpha-emitting radionuclides, which can interfere with the correct determination of 239+240Pu and 238Pu concentrations. The radiochemical procedure was validated by application to reference material and by participation in intercomparison exercises. This radiochemical procedure was applied to the different layers of a high-resolution sediment core taken from a lake in Sweden. The 239+240Pu and 238Pu/239+240Pu profiles obtained in the high-resolution sediment core correctly reproduced the expected evolution of these quantities as observed historically in the atmosphere, validating the procedure for this purpose and showing the power of these radionuclides for dating purposes.

  16. Electronic Structure, Localization and 5f Occupancy in Pu Materials

    SciTech Connect

    Joyce, John J.; Beaux, Miles F.; Durakiewicz, Tomasz; Graham, Kevin S.; Bauer, Eric D.; Mitchell, Jeremy N.; Tobash, Paul H.; Richmond, Scott

    2012-05-03

    The electronic structure of delta plutonium ({delta}-Pu) and plutonium compounds is investigated using photoelectron spectroscopy (PES). Results for {delta}-Pu show a small component of the valence electronic structure which might reasonably be associated with a 5f{sup 6} configuration. PES results for PuTe are used as an indication for the 5f{sup 6} configuration due to the presence of atomic multiplet structure. Temperature dependent PES data on {delta}-Pu indicate a narrow peak centered 20 meV below the Fermi energy and 100 meV wide. The first PES data for PuCoIn5 indicate a 5f electronic structure more localized than the 5fs in the closely related PuCoGa{sub 5}. There is support from the PES data for a description of Pu materials with an electronic configuration of 5f{sup 5} with some admixture of 5f{sup 6} as well as a localized/delocalized 5f{sup 5} description.

  17. Density functional study of Pu2C3

    NASA Astrophysics Data System (ADS)

    Yang, Rong; Tang, Bin; Gao, Tao; Ao, Bing Yun

    2017-08-01

    The structural, magnetic, electronic, vibrational, thermodynamic and elastic properties of plutonium sesquicarbide (Pu2C3) are investigated based on density functional theory. The use of the Hubbard term to describe the 5 f electrons of plutonium is discussed according the lattice parameters and magnetism. The calculated lattice constants, magnetism and density of states agree well with the experimental data or other theoretical calculations. The Pu-C bonds of Pu2C3 have a mixture of covalent character and ionic character, while covalent character is stronger than ionic character. The phonon frequencies and the assignment of infrared-active, Raman-active and silent modes at Γ point are obtained. Furthermore, the enthalpy difference H-H298, entropy S, heat capacity and linear thermal expansion coefficient α of Pu2C3 have been calculated and compared with the available data. Lastly, the calculated elastic properties predict that Pu2C3 is ductile metal. In addition, the effect of spin-orbit coupling on the structural, magnetic, and electronic properties of Pu2C3 has been discussed. We hope that our results can provide a useful reference for further theoretical and experimental research on Pu2C3.

  18. Heavy metal inventory and fuel sustainability of recycling TRU in FBR design

    SciTech Connect

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-06

    Nuclear fuel materials from spent fuel of light water reactors have a potential to be used for destructive devices with very huge energy release or in the same time, it can be utilized as a peaceful energy or civil applications, for generating electricity, desalination of water, medical application and others applications. Several research activities showed some recycled spent fuel can be used as additional fuel loading for increasing fuel breeding capability as well as improving intrinsic aspect of nuclear non-proliferation. The present investigation intends to evaluate the composition of heavy metals inventories and fuel breeding capability in the FBR design based on the loaded fuel of light water reactor (LWR) spent fuel (SF) of 33 GWd/t with 5 years cooling time by adopting depletion code of ORIGEN. Whole core analysis of FBR design is performed by adopting and coupling codes such as SLAROM code, JOINT and CITATION codes. Nuclear data library, JFS-3-J-3.2R which is based on the JENDL 3.2 has been used for nuclear data analysis. JSFR design is the basis design reference which basically adopted 800 days cycle length for 4 batches system. Higher inventories of plutonium of MOX fuel and TRU fuel types at equilibrium composition than initial composition have been shown. Minor actinide (MA) inventory compositions obtain a different inventory trends at equilibrium composition for both fuel types. Higher Inventory of MA is obtained by MOX fuel and less MA inventory for TRU fuel at equilibrium composition than initial composition. Some different MA inventories can be estimated from the different inventory trend of americium (Am). Higher americium inventory for MOX fuel and less americium inventory for TRU fuel at equilibrium condition. Breeding ratio of TRU fuel is relatively higher compared with MOX fuel type. It can be estimated from relatively higher production of Pu-238 (through converted MA) in TRU fuel, and Pu-238 converts through neutron capture to produce Pu-239

  19. Heavy metal inventory and fuel sustainability of recycling TRU in FBR design

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-01

    Nuclear fuel materials from spent fuel of light water reactors have a potential to be used for destructive devices with very huge energy release or in the same time, it can be utilized as a peaceful energy or civil applications, for generating electricity, desalination of water, medical application and others applications. Several research activities showed some recycled spent fuel can be used as additional fuel loading for increasing fuel breeding capability as well as improving intrinsic aspect of nuclear non-proliferation. The present investigation intends to evaluate the composition of heavy metals inventories and fuel breeding capability in the FBR design based on the loaded fuel of light water reactor (LWR) spent fuel (SF) of 33 GWd/t with 5 years cooling time by adopting depletion code of ORIGEN. Whole core analysis of FBR design is performed by adopting and coupling codes such as SLAROM code, JOINT and CITATION codes. Nuclear data library, JFS-3-J-3.2R which is based on the JENDL 3.2 has been used for nuclear data analysis. JSFR design is the basis design reference which basically adopted 800 days cycle length for 4 batches system. Higher inventories of plutonium of MOX fuel and TRU fuel types at equilibrium composition than initial composition have been shown. Minor actinide (MA) inventory compositions obtain a different inventory trends at equilibrium composition for both fuel types. Higher Inventory of MA is obtained by MOX fuel and less MA inventory for TRU fuel at equilibrium composition than initial composition. Some different MA inventories can be estimated from the different inventory trend of americium (Am). Higher americium inventory for MOX fuel and less americium inventory for TRU fuel at equilibrium condition. Breeding ratio of TRU fuel is relatively higher compared with MOX fuel type. It can be estimated from relatively higher production of Pu-238 (through converted MA) in TRU fuel, and Pu-238 converts through neutron capture to produce Pu-239

  20. 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)

  1. Measurement of the 242Pu neutron capture cross section

    NASA Astrophysics Data System (ADS)

    Buckner, M. Q.; Wu, C. Y.; Henderson, R. A.; Bucher, B.; Bredeweg, T. A.; Baramsai, B.; Couture, A.; Jandel, M.; Mosby, S.; O'Donnell, J. M.; Ullmann, J. L.; Chyzh, A.; Dance Collaboration

    2015-10-01

    Precision (n,f) and (n, γ) cross sections are important for the network calculations of the radiochemical diagnostic chain for the U.S. DOE's Stockpile Stewardship Program. 242Pu(n, γ) cross section is relevant to the network calculations of Pu and Am. Additionally, new reactor concepts have catalyzed considerable interest in the measurement of improved cross sections for neutron-induced reactions on key actinides. To date, little or no experimental data has been reported on 242Pu(n, γ) for incident neutron energy below 50 keV. A new measurement of the 242Pu(n, γ) reaction was performed with the DANCE together with an improved PPAC for fission-fragment detection at LANSCE during FY14. The relative scale of the 242Pu(n, γ) cross section spans four orders of magnitude for incident neutron energies from thermal to ~ 30 keV. The absolute scale of the 242Pu(n, γ) cross section is set according to the measured 239Pu(n,f) resonance at 7.8 eV; the target was spiked with 239Pu for this measurement. The absolute 242Pu(n, γ) neutron capture cross section is ~ 30% higher than the cross section reported in ENDF for the 2.7 eV resonance. Latest results to be reported. Funded by U.S. DOE Contract No. DE-AC52-07NA27344 (LLNL) and DE-AC52-06NA25396 (LANL). U.S. DOE/NNSA Office of Defense Nuclear Nonproliferation Research and Development. Isotopes (ORNL).

  2. Exposure assessment of lovastatin in Pu-erh tea.

    PubMed

    Zhao, Zhen-Jun; Pan, You-Zhao; Liu, Qin-Jin; Li, Xing-Hui

    2013-06-03

    This paper reports the results of an extensive survey on the levels of lovastatin in Pu-erh tea samples. The microbial source of lovastatin was assessed by testing the ability of fungi with higher isolation frequency in the Pu-erh tea samples to produce lovastatin on Czapek yeast extract agar (CYA). Lovastatin was not detected in any of the raw Pu-erh tea samples without storage but was found in almost all the ripe Pu-erh tea samples, with lovastatin contents ranging from 20.61 ng/gdw to 226.38 ng/gdw. After five years' storage, the lovastatin levels increased obviously in ripe Pu-erh tea samples and 55% of raw Pu-erh tea samples from 2007 were found to contain lovastatin with concentrations ranging between 28.41 ng/gdw and 228.61 ng/gdw. With increasing storage time, lovastatin concentration in ripe Pu-erh tea, and the occurrence and concentration of lovastatin for raw Pu-erh tea increased significantly. Three genera of fungi: Aspergillus, Penicillium and Trichoderma were often isolated from Pu-erh tea samples. A total of 40 strains from 3 fungal genera were selected to test their ability to produce lovastatin. Only 6 strains, Aspergillus tubingensis, Aspergillus wentii, Aspergillus fumigatus, Penicillium chrysogenum, Trichoderma asperellum and Trichoderma citrinoviride, were able to produce lovastatin reaching concentrations of 9.59 ± 0.42 ng/g CYA, 2.33 ± 0.21 ng/g CYA, 2.77 ± 0.13 ng/g CYA, 3.36 ± 0.69 ng/g CYA, 4.8 ± 0.17 ng/g CYA, and 1.47 ± 0.36 ng/g CYA respectively in Czapek yeast extract agar. Copyright © 2013 Elsevier B.V. All rights reserved.

  3. An alternative structure of Pu{sub 4}O{sub 9} (''PuO{sub 2.25}'') incorporating interstitial hydroxyl rather than oxide

    SciTech Connect

    Penneman, R.A.; Paffett, M.T. . E-mail: mtp@lanl.gov

    2005-02-15

    Bond length/bond strength relationships are applied to the Pu{sub 4}O{sub 9} ('PuO{sub 2.25}') structure proposed by others and support one Pu(V) with a central hydroxyl ion but not a central oxide ion nor formation of Pu(VI). Bond distances and bond strengths are normal for a central ion of unit charge, and reconcile the finding that cell dimensions are so minimally changed from those of PuO{sub 2}. Substitution of hydroxyl for oxide accounts for the 'excess' oxygen content of PuO{sub 2.265,} yielding PuO{sub 2}(OH){sub 0.249.} The short range, local order (structure) of the 'Pu{sub 4}O{sub 9}' entity is alternatively formulated as Pu{sub 4}O{sub 8}OH.

  4. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    SciTech Connect

    Not Available

    1994-01-15

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

  5. Process parameters optimization in ion exchange 238Pu aqueous processing

    NASA Astrophysics Data System (ADS)

    Pansoy-Hjelvik, M. E.; Nixon, J.; Laurinat, J.; Brock, J.; Silver, G.; Reimus, M.; Ramsey, K. B.

    2000-07-01

    This paper describes bench-scale efforts (5-7 grams of 238Pu) to optimize the ion exchange process for 234U separation with minimal 238Pu losses to the effluent and wash liquids. The bench-scale experiments also determine the methodology to be used for the full-scale process: 5 kg238Pu annual throughput. Heat transfer calculations used to determine the thermal gradients expected during ion exchange processing are also described. The calculations were performed in collaboration with Westinghouse Savannah River Technology Center (WSRTC) and provide information for the design of the full-scale ion exchange equipment.

  6. Elastic properties of gamma-Pu by resonant ultrasound spectroscopy

    SciTech Connect

    Migliori, Albert; Betts, J; Trugman, A; Mielke, C H; Mitchell, J N; Ramos, M; Stroe, I

    2009-01-01

    Despite intense experimental and theoretical work on Pu, there is still little understanding of the strange properties of this metal. We used resonant ultrasound spectroscopy method to investigate the elastic properties of pure polycrystalline Pu at high temperatures. Shear and longitudinal elastic moduli of the {gamma}-phase of Pu were determined simultaneously and the bulk modulus was computed from them. A smooth linear and large decrease of all elastic moduli with increasing temperature was observed. We calculated the Poisson ratio and found that it increases from 0.242 at 519K to 0.252 at 571K.

  7. An Alternative Model for Electron Correlation in Pu

    SciTech Connect

    Yu, S; Tobin, J; Soderlind, P

    2007-10-23

    Using a density functional theory based approach that treats the 5f electrons relativistically, a Pu electronic structure with zero net magnetic moment is obtained, where the 5f orbital and 5f spin moments cancel each other. By combining the spin and orbital specific densities of states with state, spin and polarization specific transition moments, it is possible to reconstruct the experimentally observed photoemission spectra from Pu. Extrapolating to a spin-resolving Fano configuration, it is shown how this would resolve the extant controversy over Pu electronic structure.

  8. 239Pu Resonance Evaluation for Thermal Benchmark System Calculations

    SciTech Connect

    Leal, Luiz C; Noguere, G; De Saint Jean, C; Kahler, A.

    2013-01-01

    Analyses of thermal plutonium solution critical benchmark systems have indicated a deciency in the 239Pu resonance evaluation. To investigate possible solutions to this issue, the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) Working Party for Evaluation Cooperation (WPEC) established Subgroup 34 to focus on the reevaluation of the 239Pu resolved resonance parameters. In addition, the impacts of the prompt neutron multiplication (nubar) and the prompt neutron ssion spectrum (PFNS) have been investigated. The objective of this paper is to present the results of the 239Pu resolved resonance evaluation eort.

  9. 239Pu Resonance Evaluation for Thermal Benchmark System Calculations

    NASA Astrophysics Data System (ADS)

    Leal, L. C.; Noguere, G.; de Saint Jean, C.; Kahler, A. C.

    2014-04-01

    Analyses of thermal plutonium solution critical benchmark systems have indicated a deficiency in the 239Pu resonance evaluation. To investigate possible solutions to this issue, the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) Working Party for Evaluation Cooperation (WPEC) established Subgroup 34 to focus on the reevaluation of the 239Pu resolved resonance parameters. In addition, the impacts of the prompt neutron multiplicity (νbar) and the prompt neutron fission spectrum (PFNS) have been investigated. The objective of this paper is to present the results of the 239Pu resolved resonance evaluation effort.

  10. Studies on preparation of (U0.47,Pu0.53)O2 microspheres by internal gelation process

    NASA Astrophysics Data System (ADS)

    Kumar, Ashok; Radhakrishna, J.; Kumar, N.; Pai, Rajesh V.; Dehadrai, J. V.; Deb, A. C.; Mukerjee, S. K.

    2013-03-01

    It is proposed to irradiate experimental fuel pins containing 750 μm (U0.47,Pu0.53)O2 microspheres (coarse fraction) and 110 μm UO2 microspheres (fine fraction) in the Fast Breeder Test Reactor at Kalpakkam, with a view to develop sphere-pac fuel for Fast Breeder Reactors in India. This communication describes optimization of the internal gelation process to produce dense sintered (U0.47,Pu0.53)O2 microspheres. Major modifications were incorporated in the feed preparation and the calcination and sintering steps. Use of pre-heated HMTA and urea and plutonium nitrate solution with optimum nitrate to plutonium ratio led to desired gelation kinetics and growth of crystallites which is necessary to obtain good gel microspheres. The calcination process was optimised to remove residual organics without cracking of microspheres. The studies showed that dense crack free sintered (U0.47,Pu0.53)O2 microspheres of about 98% theoretical density (TD) can be prepared using the internal gelation method.

  11. Nuclear weapons produced (236)U, (239)Pu and (240)Pu archived in a Porites Lutea coral from Enewetak Atoll.

    PubMed

    Froehlich, M B; Tims, S G; Fallon, S J; Wallner, A; Fifield, L K

    2017-05-16

    A slice from a Porites Lutea coral core collected inside the Enewetak Atoll lagoon, within 15 km of all major nuclear tests conducted at the atoll, was analysed for (236)U, (239)Pu and (240)Pu over the time interval 1952-1964 using a higher time resolution than previously reported for a parallel slice from the same core. In addition two sediment samples from the Koa and Oak craters were analysed. The strong peaks in the concentrations of (236)U and (239)Pu in the testing years are confirmed to be considerably wider than the flushing time of the lagoon. This is likely due to the growth mechanism of the coral. Following the last test in 1958 atom concentrations of both (236)U and (239)Pu decreased from their peak values by more than 95% and showed a seasonal signal thereafter. Between 1959 and 1964 the weighted average of the (240)Pu/(239)Pu atom ratio is 0.124 ± 0.008 which is similar to that in the lagoon sediments (0.129 ± 0.006) but quite distinct from the global fallout value of ∼0.18. This, and the high (239,240)Pu and (236)U concentrations in the sediments, provides clear evidence that the post-testing signal in the coral is dominated by remobilisation of the isotopes from the lagoon sediments rather than from global fallout. Copyright © 2017 Elsevier Ltd. All rights reserved.

  12. Proliferation resistant fuel for pebble bed modular reactors

    SciTech Connect

    Ronen, Y.; Aboudy, M.; Regev, D.; Gilad, E.

    2012-07-01

    We show that it is possible to denature the Plutonium produced in Pebble Bed Modular Reactors (PBMR) by doping the nuclear fuel with either 3050 ppm of {sup 237}Np or 2100 ppm of Am vector. A correct choice of these isotopes concentration yields denatured Plutonium with isotopic ratio {sup 238}Pu/Pu {>=} 6%, for the entire fuel burnup cycle. The penalty for introducing these isotopes into the nuclear fuel is a subsequent shortening of the fuel burnup cycle, with respect to a non-doped reference fuel, by 41.2 Full Power Days (FPDs) and 19.9 FPDs, respectively, which correspond to 4070 MWd/ton and 1965 MWd/ton reduction in fuel discharge burnup. (authors)

  13. FCRD Advanced Reactor (Transmutation) Fuels Handbook

    SciTech Connect

    Janney, Dawn Elizabeth; Papesch, Cynthia Ann

    2016-09-01

    Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. U-Pu-Zr alloys are well suited for electrolytic refining, which leads to incorporation rare-earth fission products such as La, Ce, Pr, and Nd. It is, therefore, important to understand not only the properties of U-Pu-Zr alloys but also those of U-Pu-Zr alloys with concentrations of minor actinides (Np, Am) and rare-earth elements (La, Ce, Pr, and Nd) similar to those in reprocessed fuel. In addition to requiring extensive safety precautions, alloys containing U, Pu, and minor actinides (Np and Am) are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phasetransformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, rapid oxidation, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Although less toxic, rare-earth elements such as La, Ce, Pr, and Nd are also difficult to study for similar reasons. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, particularly those that also contain minor actinides and rare-earth elements. General acceptance of results commonly indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, Np, Am, La, Ce, Pr, and Nd and

  14. PU Vulpeculae: an eclipsing symbiotic nova.

    NASA Astrophysics Data System (ADS)

    Nussbaumer, H.; Vogel, M.

    1996-03-01

    A series of IUE observation from 1992 to 1995 has definitely established PU Vul as an eclipsing binary. The outburst of this symbiotic nova began in 1977. An extended fading in 1980 gave rise to various interpretations, the eclipse scenario being one of them, dust formation being another. From AFOEV and AAVSO observations we find a period of 4900+/-100days, or 13.42+/-0.27years. An eclipsing object of such a long period signifies that we see the binary system at an orbital inclination close to 90deg. ESO observations in the near infrared give an orbital velocity of 4.7km/s and a mass function of m_f_=~0.05. Assuming a white dwarf mass between 0.4Msun_ and 0.5Msun_ gives for the red giant 0.7<=M/Msun_<=1.1. From the length of the eclipse the radius of the red giant is determined as R_giant_>=82Rsun_. We discuss IUE, HST and ground based observations of PU Vulpeculae before and during its second observed eclipse of the hot component by the cool giant which lasted from 1993 to 1995, mid-eclipse was in April 1994. Line profiles, particularly those taken by HST, allow a neat distinction between narrow nebular lines and broader wind lines which prove the existence of a fast wind from the hot star in the binary system of v=~1000km/s. That wind has relatively high densities (N_e_>10^12^cm^-3^) and is optically thick to radiation at λ<228A. Nebular lines have half widths corresponding to v=~70km/s. During the 1994 eclipse the more highly ionized lines were strongly eclipsed, whereas the lowly ionized nebular lines were hardly affected. This proves that the lowly ionized nebular lines are emitted in a very extended region, and not only close to the cool giant. From 1990 to 1994 relative C/N/O abundances of the nebular and wind emission regions have not changed beyond observational uncertainties.

  15. PU IMMOBILIZATION - INDUCTION MELTING ND OFFGAS TESTING

    SciTech Connect

    Marra, J

    2006-11-28

    The Cylindrical Induction Melter (CIM) at the Aiken County Technology Laboratory (ACTL) has been operated by the Savannah River National Laboratory (SRNL) to support the Pu Disposition Conceptual Design (CD-0) development effort. The primary purpose of this report is to summarize the offgas sampling tests conducted in the CIM to capture and analyze the particulate and vapors emitted from lanthanide borosilicate (LaBS) Frit X with HfO{sub 2} as a surrogate for PuO{sub 2} and added impurities. In addition, this report describes several initial tests of the CIM for the vitrification of LaBS Frit X with HfO{sub 2}. The activities required to produce Frit X from batch chemical oxides for subsequent milling to yield glass frit of nominally 20 micron particle size are also discussed. The tests with impurities added showed that alkali salts such as NaCl and KCl were substantially emitted into the offgas system as the salt particulate, HCl, or Cl{sub 2}. Retention of Na and K in the glass were about 80 and 55%, respectively. Chloride retention was about 35%; chloride remaining in the glass was 0.29-0.37 wt%. Based on a material balance, approximately 83% of F fed was retained in the glass at about 0.09 wt % (F could not be measured directly at this concentration). Transition metals (Ni, Cu, Fe, Mo, Cr) were also volatilized to varying extents. A very small amount (<0.1 g) of nickel compounds and KCl were found in crystals deposited on the melter offgas line. Overall, about 58-72% of the impurities added were volatilized. Virtually all of the particulate species were collected on the nominal 0.3 {micro}m filter. The particulate evolution rate ranged from 2-8 g/kg glass/h. The particulate was found to be as small as 0.2 {micro}m and have an approximate median size of 0.5 {micro}m. The particulate salt was also found to stick together by forming bridges between particles. Further runs without washable salts are recommended. Measurements of particle size distribution for use in

  16. FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION

    SciTech Connect

    Carter, J.

    2011-01-03

    The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S. (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated. (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass. (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.

  17. FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION

    SciTech Connect

    Jones, R.; Carter, J.

    2010-10-13

    The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S; (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated; (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass; and (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.

  18. (239)Pu, (240)Pu, and (241)Am determination in hot particles by low level gamma-spectrometry.

    PubMed

    Jiménez-Ramos, M C; Hurtado, S; Chamizo, E; García-Tenorio, R; León-Vintró, L; Mitchell, P I

    2010-06-01

    A nondestructive method based on low-energy, high-resolution photon spectrometry is presented which allows accurate determination of (239)Pu, (240)Pu, and (241)Am (as a daughter of (241)Pu) activities in radioactive particles containing relatively high levels of plutonium isotopes. The proposed method requires only one measurement for the establishment of an absolute efficiency curve. Since the density and composition of the radioactive particles of interest may vary, a self-absorption correction is required for the accurate determination of isotopic activities and ratios. This correction is carried out for each individual particle using the convenient gamma-ray emissions of (241)Am.

  19. Optimization of hybrid-type instrumentation for Pu accountancy of U/TRU ingot in pyroprocessing.

    PubMed

    Seo, Hee; Won, Byung-Hee; Ahn, Seong-Kyu; Lee, Seung Kyu; Park, Se-Hwan; Park, Geun-Il; Menlove, Spencer H

    2016-02-01

    One of the final products of pyroprocessing for spent nuclear fuel recycling is a U/TRU ingot consisting of rare earth (RE), uranium (U), and transuranic (TRU) elements. The amounts of nuclear materials in a U/TRU ingot must be measured as precisely as possible in order to secure the safeguardability of a pyroprocessing facility, as it contains the most amount of Pu among spent nuclear fuels. In this paper, we propose a new nuclear material accountancy method for measurement of Pu mass in a U/TRU ingot. This is a hybrid system combining two techniques, based on measurement of neutrons from both (1) fast- and (2) thermal-neutron-induced fission events. In technique #1, the change in the average neutron energy is a signature that is determined using the so-called ring ratio method, according to which two detector rings are positioned close to and far from the sample, respectively, to measure the increase of the average neutron energy due to the increased number of fast-neutron-induced fission events and, in turn, the Pu mass in the ingot. We call this technique, fast-neutron energy multiplication (FNEM). In technique #2, which is well known as Passive Neutron Albedo Reactivity (PNAR), a neutron population's changes resulting from thermal-neutron-induced fission events due to the presence or absence of a cadmium (Cd) liner in the sample's cavity wall, and reflected in the Cd ratio, is the signature that is measured. In the present study, it was considered that the use of a hybrid, FNEM×PNAR technique would significantly enhance the signature of a Pu mass. Therefore, the performance of such a system was investigated for different detector parameters in order to determine the optimal geometry. The performance was additionally evaluated by MCNP6 Monte Carlo simulations for different U/TRU compositions reflecting different burnups (BU), initial enrichments (IE), and cooling times (CT) to estimate its performance in real situations. Copyright © 2015 Elsevier Ltd. All

  20. Suitability of 239+240Pu and 137Cs as tracers for soil erosion assessment in mountain grasslands.

    PubMed

    Alewell, Christine; Meusburger, Katrin; Juretzko, Gregor; Mabit, Lionel; Ketterer, Michael E

    2014-05-01

    Anthropogenic radionuclides have been distributed globally due to nuclear weapons testing, nuclear accidents, nuclear weapons fabrication, and nuclear fuel reprocessing. While the negative consequences of this radioactive contamination are self-evident, the ubiquitous fallout radionuclides (FRNs) distribution form the basis for the use as tracers in ecological studies, namely for soil erosion assessment. Soil erosion is a major threat to mountain ecosystems worldwide. We compare the suitability of the anthropogenic FRNs, 137Cs and 239+240Pu as soil erosion tracers in two alpine valleys of Switzerland (Urseren Valley, Canton Uri, Central Swiss Alps and Val Piora, Ticino, Southern Alps). We sampled reference and potentially erosive sites in transects along both valleys. 137Cs measurements of soil samples were performed with a Li-drifted Germanium detector and 239+240Pu with ICP-MS. Our data indicates a heterogeneous deposition of the 137Cs, since most of the fallout origins from the Chernobyl April/May 1986 accident, when large parts of the European Alps were still snow-covered. In contrast, 239+240Pu fallout originated mainly from 1950s to 1960s atmospheric nuclear weapons tests, resulting in a more homogenous distribution and thus seems to be a more suitable tracer in mountainous grasslands. Soil erosion assessment using 239+240Pu as a tracer pointed to a huge dynamic and high heterogeneity of erosive processes (between sedimentation of 1.9 and 7 t ha(-1) yr(-1) and erosion of 0.2-16.4 t ha(-1) yr(-1) in the Urseren Valley and sedimentation of 0.4-20.3 t ha(-1) yr(-1) and erosion of 0.1-16.4 t ha(-1) yr(-1) at Val Piora). Our study represents a novel and successful application of 239+240Pu as a tracer of soil erosion in a mountain environment. Copyright © 2013 The Authors. Published by Elsevier Ltd.. All rights reserved.

  1. Spatial and temporal distribution of Pu in the Northwest Pacific Ocean using modern coral archives.

    PubMed

    Lindahl, Patric; Andersen, Morten B; Keith-Roach, Miranda; Worsfold, Paul; Hyeong, Kiseong; Choi, Min-Seok; Lee, Sang-Hoon

    2012-04-01

    Historical (239)Pu activity concentrations and (240)Pu/(239)Pu atom ratios were determined in skeletons of dated modern corals collected from three locations (Chuuk Lagoon, Ishigaki Island and Iki Island) to identify spatial and temporal variations in Pu inputs to the Northwest Pacific Ocean. The main Pu source in the Northwest Pacific is fallout from atmospheric nuclear weapons testing which consists of global fallout and close-in fallout from the former US Pacific Proving Grounds (PPG) in the Marshall Islands. PPG close-in fallout dominated the Pu input in the 1950s, as was observed with higher (240)Pu/(239)Pu atom ratios (>0.30) at the Ishigaki site. Specific fallout Pu contamination from the Nagasaki atomic bomb and the Ivy Mike thermonuclear detonation at the PPG were identified at Ishigaki Island from the (240)Pu/(239)Pu atom ratios of 0.07 and 0.46, respectively. During the 1960s and 1970s, global fallout was the major Pu source to the Northwest Pacific with over 60% contribution to the total Pu. After the cessation of the atmospheric nuclear tests, the PPG again dominated the Pu input due to the continuous transport of remobilised Pu from the Marshall Islands along the North Equatorial Current and the subsequent Kuroshio Current. The Pu contributions from the PPG in recent coral bands (1984 onwards) varied over time with average estimated PPG contributions between 54% and 72% depending on location. Copyright © 2011 Elsevier Ltd. All rights reserved.

  2. Neutronic double heterogeneity effect in particle dispersed type inert matrix fuels

    NASA Astrophysics Data System (ADS)

    Akie, H.; Takano, H.

    2006-06-01

    Rock-like oxide (ROX) fuel concept is studied in Japan for effective plutonium burning in light water reactors (LWRs). ROX is a heterogeneous fuel, where Pu containing yttria stabilized zirconia (YSZ) particles are dispersed in spinel matrix, and similar to the high temperature gas cooled reactor (HTR) fuel. The effect of such a 'double' heterogeneity (fuel, structure and coolant heterogeneity in reactor core, plus fuel heterogeneity) on HTR neutronic characteristics is important, while the effect was not taken into account in the ROX fueled LWR neutronics calculations. Here, this double heterogeneity effect is estimated for ROX fueled LWR, and compared with the Pu containing YSZ particle fueled HTR. As a result, the heterogeneity effect was negligible in the ROX-LWR system, while it is notable in YSZ-HTR system. The volume fraction of YSZ particle in the fuel region is one of the important parameter to cause the difference.

  3. Ferro- and antiferro-magnetism in (Np, Pu)BC

    SciTech Connect

    Klimczuk, T.; Kozub, A. L.; Griveau, J.-C.; Colineau, E.; Wastin, F.; Falmbigl, M.; Rogl, P.

    2015-04-01

    Two new transuranium metal boron carbides, NpBC and PuBC, have been synthesized. Rietveld refinements of powder XRD patterns of (Np,Pu)BC confirmed in both cases isotypism with the structure type of UBC. Temperature dependent magnetic susceptibility data reveal antiferromagnetic ordering for PuBC below T{sub N} = 44 K, whereas ferromagnetic ordering was found for NpBC below T{sub C} = 61 K. Heat capacity measurements prove the bulk character of the observed magnetic transition for both compounds. The total energy electronic band structure calculations support formation of the ferromagnetic ground state for NpBC and the antiferromagnetic ground state for PuBC.

  4. Radiation Damage Effects in Candidate Titanates for Pu Disposition: Pyrochlore

    SciTech Connect

    Strachan, Denis M; Scheele, Randall D; Buck, Edgar C; Icenhower, Jonathan P; Kozelisky, Anne E; Sell, Rachel L; Elovich, Robert J; Buchmiller, William C

    2005-10-15

    Laboratory experiments on titanate ceramics were performed to verify whether certain assumptions are valid regarding the swelling, chemical durability, and microcracking that might occur as 239Pu decays. Titanate ceramics are the material of choice for the immobilization of surplus weapons-grade Pu. The short-lived isotope, 238Pu, was incorporated into the ceramic formulation to accelerate the effects of radiation induced damage. We report on the effects of this damage on the density (volumetric swelling <6%), crystal structure of pyrochlore-bearing specimens (amorphous after about 21018 α/g), and dissolution (no change from fully the crystalline specimen). Even though the specimens became amorphous during the tests, there was no evidence for microcracking in the photomicrographs from the scanning electron microscope. Thus, although pyrochlore is susceptible to radiation-induced damage, the material remains chemically and physically viable as a material for immobilizing surplus weapons-grade Pu.

  5. PU.1 and CEBPA expression in acute myeloid leukemia.

    PubMed

    D'Alò, Francesco; Di Ruscio, Annalisa; Guidi, Francesco; Fabiani, Emiliano; Greco, Mariangela; Rumi, Carlo; Hohaus, Stefan; Voso, Maria Teresa; Leone, Giuseppe

    2008-09-01

    Alterations of the transcription factors CCAAT/enhancer binding protein alpha (CEBPA) and PU.1 have been described in acute myeloid leukemia (AML). We studied CEBPA and PU.1 mRNA levels by real-time RT-PCR in 109 primary AML samples, compared with normal bone marrow and peripheral blood cells. Low PU.1 levels were observed in monoblastic leukemias, while low CEBPA levels were associated with leukopenia at diagnosis and lack of expression of differentiation antigens CD33 and CD11c. We conclude that down-regulation of CEBPA and PU.1 is not a general feature of primary AML, but appears to be restricted to distinct AML subtypes.

  6. Experimental Bench-marking of Pu Electronic Structure

    SciTech Connect

    Lawrence Livermore National Laboratory

    2007-07-31

    Our plan is to do Ce (as a Pu surrogate) this year and be ready to do Pu next year. The Fano (Spin-resolved Photoelectron Spectroscopy) measurements are essential to testing electron correlation in the occupied 5f states. BIS (Bremstrahlung Isochromat Spectroscopy or high energy Inverse Photoelectron Spectroscopy) experiments are crucial to a quantitative determination of the 5f unoccupied density of states (5f-UDOS). The 5f UDOS is the key to differentiation between a myriad of models of 5f electronic structure. During this time, we will work to converge to a solution for the Pu safety issues, with the plan to implement these in the next FY. Acceleration of this schedule and implementation of the safety plan in this FY will require a very significant increase in funding. Ultimately, results from the Pu experiments will be fed into calculations performed by P. Soderlind, A. Landa, and others.

  7. On the electronic configuration in Pu: spectroscopy and theory

    SciTech Connect

    Tobin, J G; Soderlind, P; Landa, A; Moore, K T; Schwartz, A J; Chung, B W; Wall, M; Wills, J M; Eriksson, O; Haire, R; Kutepov, A L

    2006-10-11

    Photoelectron spectroscopy, synchrotron-radiation-based x-ray absorption, electron energy-loss spectroscopy, and density-functional calculations within the mixed-level and magnetic models, together with canonical band theory have been used to study the electron configuration in Pu. These methods suggest a 5f{sup n} configuration for Pu of 5 {le} n < 6, with n {ne} 6, contrary to what has recently been suggested in several publications. We show that the n = 6 picture is inconsistent with the usual interpretation of photoemission and x-ray absorption spectra. Instead, these spectra support the traditional conjecture of a 5f{sup 5} configuration in Pu as is obtained by density-functional theory. We further argue, based on 5f-band filling, that an n = 6 hypothesis is incompatible with the position of Pu in the actinide series and its monoclinic ground-state phase.

  8. [239Pu and chromosomal aberrations in human peripheral blood lymphocytes].

    PubMed

    Okladnikova, N D; Osovets, S V; Kudriavtseva, T I

    2009-01-01

    The genome status in somatic cells was assessed using the chromosomal aberration (CA) test in peripheral blood lymphocytes from 194 plutonium workers exposed to occupational radiation mainly from low-transportable compounds of airborne 230Pu. Pu body burden at the time of cytogenetic study varied from values close to the method sensitivity to values multiply exceeding the permissible level. Standard (routine) methods of peripheral blood lymphocytes cultivation were applied. Chromatid- and chromosomal-type structural changes were estimated. Aberrations were estimated per 100 examined metaphase cells. The quantitative relationship between the CA frequency and Pu body burden and the absorbed dose to the lung was found. Mathematical processing of results was carried out based on the phenomenological model. The results were shown as theoretical and experimental curves. The threshold of the CA yield was 0.43 +/- 0.03 kBq (Pu body burden) and 6.12 +/- 1.20 cGy (absorbed dose to the lung).

  9. [ECG indices in dogs after inhalation of 239Pu].

    PubMed

    Karpova, V N

    1985-11-01

    Dogs of both sexes aged 2 to 4 were subjected to inhalation inoculation with polymer 239Pu or submicron 239PuO2 aerosols in amounts close to acute, subacute and chronically effective ones. ECG was recorded in standard, amplified and single leads (V3). All calculations were done by lead II. Signs of the right heart overburdening were noted in the presence of the P-pulmonale complex, deep S1 wave or cardiac electrical axis of SI-SII-SIII type. Signs of the right heart overburdening were revealed after inhalation of polimer 239Pu (70%). The absence of similar changes in damage caused by 239Pu could be attributed to its fast resorption from the lungs resulting in more moderate lesion of the respiratory organs.

  10. Phase Characteristics of a U-20Pu-3Am-2Np-15Zr Metallic Alloy Containing Rare Earths

    SciTech Connect

    Douglas E. Burkes; J. Rory Kennedy; Thomas Hartmann; Cynthia A. Papesch

    2009-12-01

    Metallic fuel alloys consisting of uranium, plutonium, and zirconium with minor additions of americium and neptunium are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. The current irradiation test series design, designated AFC2, includes minor additions of rare earth elements to simulate expected fission product carry-over from the electrochemical molten salt reprocessing technique. The metal fuel alloys have been fabricated by an arc casting technique. The as-cast fuel alloys have been investigated for phase and thermal properties, specifically, enthalpies of transition, transition temperatures, and room temperature phase characteristics. Results and observations related to these characteristics for the “fresh” fuel alloys are provided. The alloy compositions are based on a U-20Pu-3Am-2Np-15Zr alloy, along with additions of 1 and 1.5 wt% RE (at the expense of U) where RE denotes rare earth alloy of cerium, lanthanum, praseodymium and neodymium). Phase behavior and associated transitions have been compared to available U-Pu-Zr ternary diagrams with acceptable agreement. Enthalpies of transition were deconvoluted from heating and cooling thermal traces for relatively reliable values. The rare earth additions to the base alloy have a minimal influence on the room temperature phases present, but the room temperature phases present slightly impacted the enthalpies of transition and transition temperatures.

  11. Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR

    SciTech Connect

    G. S. Chang; Hongbin Zhang

    2009-09-01

    One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.

  12. METHOD OF PREPARING A CERAMIC FUEL ELEMENT

    DOEpatents

    Ross, W.T.; Bloomster, C.H.; Bardsley, R.E.

    1963-09-01

    A method is described for preparing a fuel element from -325 mesh PuO/ sub 2/ and -20 mesh UO/sub 2/, and the steps of screening --325 mesh UO/sub 2/ from the -20 mesh UO/sub 2/, mixing PuO/sub 2/ with the --325 mesh UO/sub 2/, blending this mixture with sufficient --20 mesh UO/sub 2/ to obtain the desired composition, introducing the blend into a metal tube, repeating the procedure until the tube is full, and vibrating the tube to compact the powder are included. (AEC)

  13. An examination of the potential fission-bomb weaponizability of nuclides other than 235U and 239Pu

    NASA Astrophysics Data System (ADS)

    Reed, B. Cameron

    2017-01-01

    Long-lived fissionable isotopes other than uranium-235 and plutonium-239 are examined for possible use in fission weapons. A few other isotopes are potentially weaponizable and in some cases have been tried or their criticality experimentally demonstrated. In most cases, however, promising isotopes are either extremely rare, difficult to produce in quantity, or hazardous to handle. Some isotopes can serve to boost the yield of fission weapons, but 235U and 239Pu are likely to remain the only practical primary fuels for nuclear weapons. In view of this, and the fact that this analysis gives no engineering details on the design of nuclear weapons, this paper will be of no assistance to putative bomb-makers; rather, my purpose is to clarify the physics similarities between 235U and 239Pu that make them suitable candidates for fission weapons.

  14. Nuclear resonance fluorescence excitations near 2 MeV in {sup 235}U and {sup 239}Pu

    SciTech Connect

    Bertozzi, W.; Korbly, S. E.; Ledoux, R. J.; Park, W. H.; Caggiano, J. A.; Hensley, W. K.; Warren, G. A.; Johnson, M. S.; McNabb, D. P.; Norman, E. B.

    2008-10-15

    A search for nuclear resonance fluorescence excitations in {sup 235}U and {sup 239}Pu within the energy range of 1.0- to 2.5-MeV was performed using a 4-MeV continuous bremsstrahlung source at the High Voltage Research Laboratory at the Massachusetts Institute of Technology. Measurements utilizing high purity Ge detectors at backward angles identified nine photopeaks in {sup 235}U and 12 photopeaks in {sup 239}Pu in this energy range. These resonances provide unique signatures that allow the materials to be nonintrusively detected in a variety of environments including fuel cells, waste drums, vehicles, and containers. The presence and properties of these states may prove useful in understanding the mechanisms for mixing low-lying collective dipole excitations with other states at low excitations in heavy nuclei.

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

    SciTech Connect

    Finn, P. A.

    1998-04-15

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

  16. Probing Actinide Electronic Structure through Pu Cluster Calculations

    SciTech Connect

    Ryzhkov, Mickhail V.; Mirmelstein, Alexei; Yu, Sung-Woo; Chung, Brandon W.; Tobin, James G.

    2013-02-26

    The calculations for the electronic structure of clusters of plutonium have been performed, within the framework of the relativistic discrete-variational method. Moreover, these theoretical results and those calculated earlier for related systems have been compared to spectroscopic data produced in the experimental investigations of bulk systems, including photoelectron spectroscopy. Observation of the changes in the Pu electronic structure as a function of size provides powerful insight for aspects of bulk Pu electronic structure.

  17. Probing Actinide Electronic Structure through Pu Cluster Calculations

    DOE PAGES

    Ryzhkov, Mickhail V.; Mirmelstein, Alexei; Yu, Sung-Woo; ...

    2013-02-26

    The calculations for the electronic structure of clusters of plutonium have been performed, within the framework of the relativistic discrete-variational method. Moreover, these theoretical results and those calculated earlier for related systems have been compared to spectroscopic data produced in the experimental investigations of bulk systems, including photoelectron spectroscopy. Observation of the changes in the Pu electronic structure as a function of size provides powerful insight for aspects of bulk Pu electronic structure.

  18. Historical changes in 239Pu and 240Pu sources in sedimentary records in the East China Sea: Implications for provenance and transportation

    NASA Astrophysics Data System (ADS)

    Wang, Jinlong; Baskaran, Mark; Hou, Xiaolin; Du, Jinzhou; Zhang, Jing

    2017-05-01

    Concentrations and isotopic compositions of plutonium (Pu) are widely used for its source identification and to determine transport processes of Pu-associated particulate matter and water. We investigated the concentrations of 239Pu and 240Pu and their ratios in a number of sediment samples from the East China Sea (ECS) collected in the summer of 2013 (August 6-28). The 239+240Pu activity concentrations in surface sediment samples were found to range between 0.048 and 0.492 Bq kg-1 and the 240Pu/239Pu atom ratios showed a similar trend as that of the 239, 240Pu activities; the Pu atom ratios ranged from 0.158 to 0.297 and were mostly higher than the mean global fallout value of 0.18. The 239, 240Pu inventories in the ECS varied widely, from 2 to 807 Bq m-2, with the highest values commonly found in the coastal areas. In the Yangtze Estuary, the mean 239+240Pu activity concentration is close to the estimated value of the suspended material from the Yangtze River catchment (0.18 Bq kg-1), and the 240Pu/239Pu atom ratio was found to be ∼0.18, which indicates that the Yangtze River input is the dominant source of Pu for this area. The total annual Yangtze River input of 239+240Pu was estimated to be 2.4 ×1010 Bq, which is small compared to the total amount of 239+240Pu buried, 3.1 ×1013 Bq in the whole ECS. The Pacific Proving Ground input appears to be the dominant source of Pu to the ECS, accounting for 45%-52% of the total inventory. The fractional amount of 239+240Pu scavenged from the total 239+240Pu transported by the Kuroshio Current (KC) and Taiwan Warm Current (TWC) into ECS sediments is estimated to be ∼10%. Our study shows that the 240Pu/239Pu atom ratio is useful not only to obtain a better insight of the biogeochemistry influenced by the KC, but also to trace the long-range transport of other particle-reactive species. Besides, the sedimentation rates obtained based on the penetration depths of 239+240Pu and vertical profiles of excess 210Pb agree

  19. IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

    SciTech Connect

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

    This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis, the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  20. PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

    SciTech Connect

    S. T. Khericha

    2007-04-01

    The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.

  1. Complexation Behavior of the Tri-n-butyl Phosphate Ligand with Pu(IV) and Zr(IV): A Computational Study.

    PubMed

    Gopakumar, Gopinadhanpillai; Sreenivasulu, B; Suresh, A; Brahmmananda Rao, C V S; Sivaraman, N; Joseph, M; Anoop, Anakuthil

    2016-06-23

    Tri-n-butyl phosphate (TBP), used as the extractant in nuclear fuel reprocessing, shows superior extraction abilities for Pu(IV) over a large number of fission products including Zr(IV). We have applied density functional theory (DFT) calculations to explain this selectivity by investigating differences in electronic structures of Pu(NO3)4·2TBP and Zr(NO3)4·2TBP complexes. On the basis of our quantum chemical calculations, we have established the lowest energy electronic states for both complexes; the quintet is the ground state for the former, whereas the latter exists in the singlet spin state. The calculated structural parameters for the optimized geometry of the plutonium complex are in agreement with the experimental results. Atoms in Molecules analysis revealed a considerable amount of ionic character to M-O{TBP} and M-O{NO3} bonds. Additionally, we have also investigated the extraction behavior of TBP for metal nitrates and have estimated the extraction energies to be -73.1 and -57.6 kcal/mol for Pu(IV) and Zr(IV), respectively. The large extraction energy of Pu(IV) system is in agreement with the observed selectivity in the extraction of Pu.

  2. Method for photochemical reduction of uranyl nitrate by tri-N-butyl phosphate and application of this method to nuclear fuel reprocessing

    DOEpatents

    De Poorter, Gerald L.; Rofer-De Poorter, Cheryl K.

    1978-01-01

    Uranyl ion in solution in tri-n-butyl phosphate is readily photochemically reduced to U(IV). The product U(IV) may effectively be used in the Purex process for treating spent nuclear fuels to reduce Pu(IV) to Pu(III). The Pu(III) is readily separated from uranium in solution in the tri-n-butyl phosphate by an aqueous strip.

  3. Study of neutron-deficient isotopes of Fl in the 239Pu, 240Pu + 48Ca reactions

    NASA Astrophysics Data System (ADS)

    Voinov, A. A.; Utyonkov, V. K.; Brewer, N. T.; Oganessian, Yu Ts; Rykaczewski, K. P.; Abdullin, F. Sh; Dmitriev, S. N.; Grzywacz, R. K.; Itkis, M. G.; Miernik, K.; Polyakov, A. N.; Roberto, J. B.; Sagaidak, R. N.; Shirokovsky, I. V.; Shumeiko, M. V.; Tsyganov, Yu S.; Subbotin, V. G.; Sukhov, A. M.; Sabelnikov, A. V.; Vostokin, G. K.; Hamilton, J. H.; Stoyer, M. A.; Strauss, S. Y.

    2016-07-01

    The results of the experiments aimed at the synthesis of Fl isotopes in the 239Pu + 48Ca and 240Pu + 48Ca reactions are presented. The experiment was performed using the Dubna gas-filled recoil separator at the U400 cyclotron. In the 239Pu+48Ca experiment one decay of spontaneously fissioning 284Fl was detected at 245-MeV beam energy. In the 240Pu+48Ca experiment three decay chains of 285Fl were detected at 245 MeV and four decays were assigned to 284Fl at the higher 48Ca beam energy of 250 MeV. The α-decay energy of 285Fl was measured for the first time and decay properties of its descendants 281Cn, 277Ds, 273Hs, 269Sg, and 265Rf were determined more precisely. The cross section of the 239Pu(48Ca,3n)284Fl reaction was observed to be about 20 times lower than those predicted by theoretical models and 50 times less than the value measured in the 244Pu+48Ca reaction. The cross sections of the 240Pu(48Ca,4-3n)284,285Fl at both 48Ca energies are similar and exceed that observed in the reaction with lighter isotope 239Pu by a factor of 10. The decay properties of the synthesized nuclei and their production cross sections indicate rapid decrease of stability of superheavy nuclei with departing from the neutron number N=184 predicted to be the next magic number.

  4. Nuclear Fuel Reprocessing

    SciTech Connect

    Harold F. McFarlane; Terry Todd

    2013-11-01

    Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore. Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor

  5. Investigation of the oxidation states of Pu isotopes in a hydrochloric acid solution.

    PubMed

    Lee, M H; Kim, J Y; Kim, W H; Jung, E C; Jee, K Y

    2008-12-01

    The characteristics of the oxidation states of Pu in a hydrochloric acid solution were investigated and the results were applied to a separating of Pu isotopes from IAEA reference soils. The oxidation states of Pu(III) and Pu(IV) were prepared by adding hydroxylamine hydrochloride and sodium nitrite to a Pu stock solution, respectively. Also, the oxidation state of Pu(VI) was adjusted with concentrated HNO(3) and HClO(4). The stability of the various oxidation states of plutonium in a HCl solution with elapsed time after preparation were found to be in the following order: Pu(III) approximately Pu(VI)>Pu(IV)>Pu(V). The chemical recoveries of Pu(IV) in a 9M HCl solution with an anion exchange resin were similar to those of Pu(VI). This method for the determination of Pu isotopes with an anion exchange resin in a 9M HCl medium was applied to IAEA reference soils where the activity concentrations of (239,240)Pu and (238)Pu in IAEA-375 and IAEA-326 were consistent with the reference values reported by the IAEA.

  6. Fueling systems

    SciTech Connect

    Gorker, G.E.

    1987-01-01

    This report deals with concepts of the Tiber II tokamak reactor fueling systems. Contained in this report are the fuel injection requirement data, startup fueling requirements, intermediate range fueling requirements, power range fueling requirements and research and development considerations. (LSR)

  7. Scavenged (239)Pu, (240)Pu, and (241)Am from snowfalls in the atmosphere settling on Mt. Zugspitze in 2014, 2015 and 2016.

    PubMed

    Gückel, Katharina; Shinonaga, Taeko; Christl, Marcus; Tschiersch, Jochen

    2017-09-19

    Concentrations of (239)Pu, (240)Pu, and (241)Am, and atomic ratio of (240)Pu/(239)Pu in freshly fallen snow on Mt. Zugspitze collected in 2014, 2015 and 2016 were determined by accelerator mass spectrometry (AMS). For the sub-femtogram (10(-15) g) - level of Pu and Am analysis, a chemical separation procedure combined with AMS was improved and an excellent overall efficiency of about 10(-4) was achieved. The concentration of (239)Pu ranges from 75 ± 13 ag/kg to 2823 ± 84 ag/kg, of (240)Pu from 20.6 ± 5.2 to 601 ± 21 ag/kg, and of (241)Am was found in the range of 16.7 ± 5.0-218.8 ± 8.9 ag/kg. Atomic ratios of (240)Pu/(239)Pu for most samples are comparable to the fallout in middle Europe. One exceptional sample shows a higher Pu concentration. High airborne dust concentration, wind directions, high Cs concentrations and the activity ratio of (239+240)Pu/(137)Cs lead to the conclusion that the sample was influenced by Pu in Saharan dust transported to Mt. Zugspitze.

  8. UO2 and PuO2 utilization in high temperature engineering test reactor with helium coolant

    NASA Astrophysics Data System (ADS)

    Waris, Abdul; Aji, Indarta K.; Novitrian, Pramuditya, Syeilendra; Su'ud, Zaki

    2016-03-01

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO2 fuel. In this study, we have evaluated the use of UO2 and PuO2 in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of 235U in loaded fuel is 18.0% or above.

  9. UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

    SciTech Connect

    Waris, Abdul Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki; Aji, Indarta K.

    2016-03-11

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

  10. A study of accelerated radiation damage effects in PuO2 and gadolinia-stabilized cubic zirconia, Zr0.79Gd0.14Pu0.07O1.93, doped with 238Pu

    NASA Astrophysics Data System (ADS)

    Burakov, B. E.; Yagovkina, M. A.

    2015-12-01

    Polycrystalline samples of cubic zirconia, Zr0.79Gd0.14Pu0.07O1.93, doped with approximately 9.9 wt.% 238Pu, and PuO2 containing 11.0 wt. % 238Pu (and main isotope is 239Pu) have been repeatedly studied during many years by X-ray diffraction analysis. At a temperature of 25 °C the unit-cell parameter of PuO2 increases depending on accumulated dose, and is accompanied by decrease of coherent scattering region (CSR). Self-irradiation of Zr0.79Gd0.14Pu0.07O1.93 is accompanied with repeated change of unit-cell parameter and CSR.

  11. Characterization of Nuclear Fuel using Multivariate Statistical Analysis

    SciTech Connect

    Robel, M; Robel, M; Robel, M; Kristo, M J; Kristo, M J

    2007-11-27

    Various combinations of reactor type and fuel composition have been characterized using principle components analysis (PCA) of the concentrations of 9 U and Pu isotopes in the 10 fuel as a function of burnup. The use of PCA allows the reduction of the 9-dimensional data (isotopic concentrations) into a 3-dimensional approximation, giving a visual representation of the changes in nuclear fuel composition with burnup. Real-world variation in the concentrations of {sup 234}U and {sup 236}U in the fresh (unirradiated) fuel was accounted for. The effects of reprocessing were also simulated. The results suggest that, 15 even after reprocessing, Pu isotopes can be used to determine both the type of reactor and the initial fuel composition with good discrimination. Finally, partial least squares discriminant analysis (PSLDA) was investigated as a substitute for PCA. Our results suggest that PLSDA is a better tool for this application where separation between known classes is most important.

  12. Static electric dipole polarizabilities of tri- and tetravalent U, Np, and Pu ions.

    PubMed

    Parmar, Payal; Peterson, Kirk A; Clark, Aurora E

    2013-11-21

    High-quality static electric dipole polarizabilities have been determined for the ground states of the hard-sphere cations of U, Np, and Pu in the III and IV oxidation states. The polarizabilities have been calculated using the numerical finite field technique in a four-component relativistic framework. Methods including Fock-space coupled cluster (FSCC) and Kramers-restricted configuration interaction (KRCI) have been performed in order to account for electron correlation effects. Comparisons between polarizabilities calculated using Dirac-Hartree-Fock (DHF), FSCC, and KRCI methods have been made using both triple- and quadruple-ζ basis sets for U(4+). In addition to the ground state, this study also reports the polarizability data for the first two excited states of U(3+/4+), Np(3+/4+), and Pu(3+/4+) ions at different levels of theory. The values reported in this work are the most accurate to date calculations for the dipole polarizabilities of the hard-sphere tri- and tetravalent actinide ions and may serve as reference values, aiding in the calculation of various electronic and response properties (for example, intermolecular forces, optical properties, etc.) relevant to the nuclear fuel cycle and material science applications.

  13. Static Electric Dipole Polarizabilities of Tri- and Tetravalent U, Np, and Pu Ions

    SciTech Connect

    Parmar, Payal; Peterson, Kirk A.; Clark, Aurora E.

    2013-11-21

    High-quality static electric dipole polarizabilities have been determined for the ground states of the hard-sphere cations of U, Np, and Pu in the III and IV oxidation states. The polarizabilities have been calculated using the numerical finite field technique in a four-component relativistic framework. Methods including Fock-space coupled cluster (FSCC) and Kramers-restricted configuration interaction (KRCI) have been performed in order to account for electron correlation effects. Comparisons between polarizabilities calculated using Dirac-Hartree-Fock (DHF), FSCC, and KRCI methods have been made using both triple- and quadruple-ζ basis sets for U⁴⁺. In addition to the ground state, this study also reports the polarizability data for the first two excited states of U3+/4+, Np3+/4+, and Pu3+/4+ ions at different levels of theory. The values reported in this work are the most accurate to date calculations for the dipole polarizabilities of the hard-sphere tri- and tetravalent actinide ions and may serve as reference values, aiding in the calculation of various electronic and response properties (for example, intermolecular forces, optical properties, etc.) relevant to the nuclear fuel cycle and material science applications.

  14. {sup 238}PuO{sub 2} heat sources: An enabling technology for space exploration

    SciTech Connect

    George, T.G.

    1998-12-31

    Surprisingly, during the 35 yr that {sup 238}PuO{sub 2}-fueled radioisotope power systems have been demonstrated and continuously improved in terms of safety, reliability, and efficiency, US capabilities to produce {sup 238}PuO{sub 2} have significantly decreased. At the same time, progress in the efficiency and longevity of chemical and solar power systems has reduced the suite of potential applications for radioisotope power systems remain a viable option for deep space exploration and planetary missions that must survive hostile operating environments. The enabling aspect of radioisotope power supplies for deep space exploration missions can best be illustrated by a comparative analysis of power supply options available for the National Aeronautics and Space Administration`s most recent planetary explorer, the Cassini mission to Saturn. A comparative evaluation of currently available power supply options for the Cassini mission, demonstrates the significant mass and maneuverability penalties that would result from the use of chemical or solar energy sources in place of the RTGs and heater units. Absent additional nuclear power options, such as a spacecraft reactor, radioisotope power sources were clearly the only viable option for providing electrical and thermal power to the Cassini spacecraft.

  15. Determination of plutonium content in high burnup pressurized water reactor fuel samples and its use for isotope correlations for isotopic composition of plutonium.

    PubMed

    Joe, Kihsoo; Jeon, Young-Shin; Han, Sun-Ho; Lee, Chang-Heon; Ha, Yeong-Keong; Song, Kyuseok

    2012-06-01

    The content of plutonium isotopes in high burnup pressurized water reactor fuel samples was examined using both alpha spectrometry and mass spectrometry after anion exchange separation. The measured values were compared with results calculated by the ORIGEN-2 code. On average, the ratios (m/c) of the measured values (m) over the calculated values (c) were 1.22±0.16 for (238)Pu, 1.02±0.14 for (239)Pu, 1.08±0.06 for (240)Pu, 1.06±0.16 for (241)Pu, and 1.13±0.08 for (242)Pu. Using the Pu data obtained in this work, correlations were derived between the alpha activity ratios of (238)Pu/((239)Pu+(240)Pu), the alpha specific activities of Pu, and the atom % abundances of the Pu isotopes. Using these correlations, the atom % abundances of the plutonium isotopes in the target samples were calculated. These calculated results agreed within a range from 2 to 8% of the experimentally derived values according to the isotopes of plutonium.

  16. Spent fuel temperature and age determination from the analysis of uranium and plutonium isotopics

    SciTech Connect

    Scott, Mark R; Eccleston, George W; Bedell, Jeffrey J; Lockard, Chanelle M

    2009-01-01

    The capability to determine the age (time since irradiation) of spent fuel can be useful for verification and safeguards. While the age of spent fuel can be determined based on measurements of short-lived fission products, these measurements are not routinely done nor generally reported. As an alternative, age can also be determined if the uranium (U) and plutonium (Pu) isotopic values are available. Uranium isotopics are not strongly affected by fuel temperature, and bumup is determined from the {sup 235}U and {sup 236}U isotopic values. Age is calculated after estimating the {sup 241}Pu at the end of irradiation while accounting for the fuel temperature, which is determined from {sup 239}Pu or {sup 240}Pu. Burnup and age determinations are calibrated to reactor models that provide uranium and plutonium isotopics over the range of fuel irradiation. The reactor model must contain sufficient fidelity on details of the reactor type, fuel burnup, irradiation history, initial fuel enrichment and fuel temperature to obtain accurate isotopic calculations. If the latter four are unknown, they can be derived from the uranium and plutonium isotopics. Fuel temperature has a significant affect on the production of plutonium isotopics; therefore, one group cross section reactor models, such as ORIGEN, cannot be used for these calculations. Multi-group cross section set codes, such as Oak Ridge National Laboratory's TRITON code, must be used.

  17. Pu and 137Cs in the Yangtze River estuary sediments: distribution and source identification.

    PubMed

    Liu, Zhiyong; Zheng, Jian; Pan, Shaoming; Dong, Wei; Yamada, Masatoshi; Aono, Tatsuo; Guo, Qiuju

    2011-03-01

    Pu isotopes and (137)Cs were analyzed using sector field ICP-MS and γ spectrometry, respectively, in surface sediment and core sediment samples from the Yangtze River estuary. (239+240)Pu activity and (240)Pu/(239)Pu atom ratios (>0.18) shows a generally increasing trend from land to sea and from north to south in the estuary. This spatial distribution pattern indicates that the Pacific Proving Grounds (PPG) source Pu transported by ocean currents was intensively scavenged into the suspended sediment under favorable conditions, and mixed with riverine sediment as the water circulated in the estuary. This process is the main control for the distribution of Pu in the estuary. Moreover, Pu is also an important indicator for monitoring the changes of environmental radioactivity in the estuary as the river basin is currently the site of extensive human activities and the sea level is rising because of global climate changes. For core sediment samples the maximum peak of (239+240)Pu activity was observed at a depth of 172 cm. The sedimentation rate was estimated on the basis of the Pu maximum deposition peak in 1963-1964 to be 4.1 cm/a. The contributions of the PPG close-in fallout Pu (44%) and the riverine Pu (45%) in Yangtze River estuary sediments are equally important for the total Pu deposition in the estuary, which challenges the current hypothesis that the riverine Pu input was the major source of Pu budget in this area.

  18. Thorium-Based Transmuter Fuels for Light Water Reactors

    SciTech Connect

    J. Stephen Herring; P. E. MacDonald; K. Weaver

    2004-04-01

    A light water reactor (LWR) fuel cycle is proposed where the reactor core mainly consists of standard uranium-dioxide (UO2) fuel rods with typical 235U enrichment, along with thoria-urania (ThO2-UO2) or yttria-stablized zirconia fertile-free fuel rods containing the plutonium and minor actinides typical of 30-yr old UO2 fuel in 1/9 to 1/3 of the positions. The goals of this mono-recycling strategy or "twice through fuel cycle" are to transmute the great majority of the long lived actinides in existing LWRs and to discharge a fuel form that is a very robust waste form and whose isotopic content is very proliferation resistant. The incorporation of plutonium into a ThO2 or yttria-stablized zirconia fertile-free matrix results in the consumption of already-separated plutonium without breeding significant additional 239Pu. The minor actinides (i.e., neptunium, americium, curium, berkelium, californium, etc.) are also included in the ThO2 or fertile-free transmuter fuel rods to further reduce the overall long-term radiotoxicity of the fuel cycle. Our analyses have shown that thorium-based or fertile-free fuels can reduce the amount of 239Pu needing further transmutation or going to a repository by ~90%. Also, thorium-based fuels produce a mixture of plutonium isotopes high in 238Pu. Because of the high decay heat and spontaneous neutron generation of 238Pu, this isotope provides intrinsic proliferation resistance.

  19. Thorium-Based Transmuter Fuels for Light Water Reactors

    SciTech Connect

    Herring, J. Stephen; MacDonald, Philip E.; Weaver, Kevan D.

    2004-07-15

    A light water reactor (LWR) fuel cycle is proposed where the reactor core mainly consists of standard uranium-dioxide (UO{sub 2}) fuel rods with typical {sup 235}U enrichment, along with thoria-urania (ThO{sub 2}-UO{sub 2}) or yttria-stablized zirconia fertile-free fuel rods containing the plutonium and minor actinides typical of 30-yr old UO{sub 2} fuel in 1/9 to 1/3 of the positions. The goals of this mono-recycling strategy or 'twice through fuel cycle' are to transmute the great majority of the long lived actinides in existing LWRs and to discharge a fuel form that is a very robust waste form and whose isotopic content is very proliferation resistant. The incorporation of plutonium into a ThO{sub 2} or yttria-stablized zirconia fertile-free matrix results in the consumption of already-separated plutonium without breeding significant additional {sup 239}Pu. The minor actinides (i.e., neptunium, americium, curium, berkelium, californium, etc.) are also included in the ThO{sub 2} or fertile-free transmuter fuel rods to further reduce the overall long-term radiotoxicity of the fuel cycle. Our analyses have shown that thorium-based or fertile-free fuels can reduce the amount of {sup 239}Pu needing further transmutation or going to a repository by {approx}90%. Also, thorium-based fuels produce a mixture of plutonium isotopes high in {sup 238}Pu. Because of the high decay heat and spontaneous neutron generation of {sup 238}Pu, this isotope provides intrinsic proliferation resistance.

  20. Quantitative Fissile Assay In Used Fuel Using LSDS System

    NASA Astrophysics Data System (ADS)

    Lee, YongDeok; Jeon, Ju Young; Park, Chang-Je

    2017-09-01

    A quantitative assay of isotopic fissile materials (U235, Pu239, Pu241) was done at Korea Atomic Energy Research Institute (KAERI), using lead slowing down spectrometer (LSDS). The optimum design of LSDS was performed based on economics, easy maintenance and assay effectiveness. LSDS system consists of spectrometer, neutron source, detection and control. LSDS system induces fissile fission and fast neutrons are collected at fission chamber. The detected signal has a direct relation to the mass of existing fissile isotopes. Many current commercial assay technologies have a limitation in direct application on isotopic fissile assay of spent fuel, except chemical analysis. In the designed system, the fissile assay model was setup and the correction factor for self-shield was obtained. The isotopic fissile content assay was performed by changing the content of Pu239. Based on the fuel rod, the isotopic content was consistent with 2% uncertainty for Pu239. By applying the covering (neutron absorber), the effective shielding was obtained and the activation was calculated on the target. From the assay evaluation, LSDS technique is very powerful and direct to analyze the isotopic fissile content. LSDS is applicable for nuclear fuel cycle and spent fuel management for safety and economics. Additionally, an accurate fissile content will contribute to the international transparency and credibility on spent fuel.

  1. Out-of-pile chemical compatibility of hyperstoichiometric (Pu 0.7U 0.3)C with stainless steel cladding and sodium coolant

    NASA Astrophysics Data System (ADS)

    Ganguly, C.; Sengupta, A. K.

    1988-09-01

    Chemical compatibility experiments of the hitherto unknown fuel (Pu 0.7U 0.3)C 1+xwith sodium coolant and SS 316 (20% cold-worked) cladding were carried out at 973 K for 1000 h for its out-of-pile 'proof testing'. Any possible chemical interaction was assessed by metallographic examination and microhardness measurements of SS 316 cladding specimens. Hyperstoichiometric (Pu 0.7U 0.3)C containing upto 0.7% oxygen and 20% mixed sesquicarbide (M 2C 3) did not interact at all with sodium and caused insignificant carburization of the SS 316 cladding. Mixed carbide pellets containing high 'O' (~ 1%) and high M 2C 3(~ 60%) caused clad carburization to a depth of around 90 μm. These experiments generated valuable informations in support of choosing plutonium rich mixed carbide as the driver fuel for the fast breeder test reactor (FBTR).

  2. The I-Pu-Xe age of the Moon

    NASA Technical Reports Server (NTRS)

    Swindle, T. D.; Caffee, M. W.; Hohenberg, C. M.

    1984-01-01

    The Rb-Sr analyses of some lunar samples which indicate that the Moon is close to the age of primitive meteorites are only reliable to within about 100 m.y. A potentially more precise chronometer is the I-Pu-Xe system. I129 has a 17 m.y. halflife and decays to Xe129; Pu244, with an 82 m.y. halflife, produces Xe131 to Xe136 in fission. The I129/Pu244 ratio has a halflife of 21 m.y. Xenon retention for the Earth could have begun as late as the event that gave birth to the Moon. For the Moon, it is hard to imagine that xenon retention could have begun before re-accretion of the fissioned (and initially dispersed?) material, particularly if that material got hot enough to account for the depletion of the volatile elements. Thus, if fission model are correct, xenon retention in the Earth certainly began no later than in the Moon, and possibly began earlier. Therefore, the I-Pu-Xe system is only marginally consistent with a fission origin. If further study confirms that the I/U ratio of the Moon is .01 or less, or if gas-rich lunar highland breccias with higher ratios of I129 to Pu244 are found, it would be difficult to explain the results in an earth-fission model of lunar origin.

  3. Neutron Capture Cross Section Measurement on $^{238}$Pu at DANCE

    SciTech Connect

    Chyzh, A; Wu, C Y

    2011-02-14

    The proposed neutron capture measurement for {sup 238}Pu was carried out in Nov-Dec, 2010, using the DANCE array at LANSCE, LANL. The total beam-on-target time is about 14 days plus additional 5 days for the background measurement. The target was prepared at LLNL with the new electrplating cell capable of plating the {sup 238}Pu isotope simultaneously on both sides of the 3-{micro}m thick Ti backing foil. A total mass of 395 {micro}g with an activity of 6.8 mCi was deposited onto the area of 7 mm in diameter. The {sup 238}Pu sample was enriched to 99.35%. The target was covered by 1.4 {micro}m double-side aluminized mylar and then inserted into a specially designed vacuum-tight container, shown in Fig. 1, for the {sup 238}Pu containment. The container was tested for leaks in the vacuum chamber at LLNL. An identical container without {sup 238}Pu was made as well and used as a blank for the background measurement.

  4. Soft tissue tumors induced by monomeric {sup 239}Pu

    SciTech Connect

    Lloyd, R.D.; Angus, W.; Taylor, G.N.; Miller, S.C.

    1995-10-01

    Individual records of soft tissue tumor occurrence (lifetime incidence) among 236 beagles injected with {sup 239}Pu citrate as young adults and 131 comparable control beagles given no radioactivity enabled us to analyze the possible effects on soft tissue tumor induction resulting from internal exposure to {sup 239}Pu. A significant trend was identified in the proportion of animals having malignant liver tumors with increasing radiation dose from {sup 239}. There was also a significant difference in the relative numbers of both malignant liver tumors (18.1 expected, 66 observed). Malignant tumors of the mouth, pancreas, and skin were more frequent among controls than among the dogs given {sup 239}Pu as well as tumors (malignant plus benign) of the mouth, pancreas, testis, and vagina. For all other tumor sites or types, there was no significant difference for both malignant and all (malignant plus benign) tumors. Mammary tumor occurrence appeared not to be associated with {sup 239}Pu incorporation. We conclude that the only soft-tissue neoplasia induced by the intake of {sup 239}Pu directly into blood is probably a liver tumor. 20 refs., 6 tabs.

  5. Pulmonary Function in Flight (PuFF) Experiment

    NASA Technical Reports Server (NTRS)

    2003-01-01

    In this International Space Station (ISS) onboard photo, Expedition Six Science Officer Donald R. Pettit works to set up the Pulmonary Function in Flight (PuFF) experiment hardware in the Destiny Laboratory. Expedition Six is the fourth and final crew to perform the PuFF experiment. The PuFF experiment was developed to better understand what effects long term exposure to microgravity may have on the lungs. The focus is on measuring changes in the everness of gas exchange in the lungs, and on detecting changes in respiratory muscle strength. It allows astronauts to measure blood flow through the lungs, the ability of the lung to take up oxygen, and lung volumes. Each PuFF session includes five lung function tests, which involve breathing only cabin air. For each planned extravehicular (EVA) activity, a crew member performs a PuFF test within one week prior to the EVA. Following the EVA, those crew members perform another test to document the effect of exposure of the lungs to the low-pressure environment of the space suits. This experiment utilizes the Gas Analyzer System for Metabolic Analysis Physiology, or GASMAP, located in the Human Research Facility (HRF), along with a variety of other Puff equipment including a manual breathing valve, flow meter, pressure-flow module, pressure and volume calibration syringes, and disposable mouth pieces.

  6. Measurements Conducted on an Unknown Object Labeled Pu-239

    SciTech Connect

    Hoteling, Nathan

    2013-11-18

    Measurements were carried out on 12 November 2013 to determine whether Pu-239 was present on an object discovered in a plastic bag with label “Pu-­239 6 uCi.” Following initial survey measurements to verify that the object was not leaking or contaminated, spectra were collected with a High Purity Germanium (HPGe) detector with object positioned in two different configurations. Analysis of the spectra did not yield any direct evidence of Pu-­239. From the measured spectra, minimum detectable activity (MDA) was determined to be approximately 2 uCi for the gamma-­ray measurements. Although there was no direct evidence of Pu-239, a peak at 60 keV characteristic of Am-­241 decay was observed. Since it is very likely that Am-­241 would be present in aged plutonium samples, this was interpreted as indirect evidence for the presence of plutonium on the object. Analysis of this peak led to an estimated Pu-­239 activity of 0.02–0.04 uCi, or <1x10-6 grams.

  7. Pulmonary Function in Flight (PuFF) Experiment

    NASA Technical Reports Server (NTRS)

    2003-01-01

    In this International Space Station (ISS) onboard photo, Expedition Six Science Officer Donald R. Pettit works to set up the Pulmonary Function in Flight (PuFF) experiment hardware in the Destiny Laboratory. Expedition Six is the fourth and final crew to perform the PuFF experiment. The PuFF experiment was developed to better understand what effects long term exposure to microgravity may have on the lungs. The focus is on measuring changes in the everness of gas exchange in the lungs, and on detecting changes in respiratory muscle strength. It allows astronauts to measure blood flow through the lungs, the ability of the lung to take up oxygen, and lung volumes. Each PuFF session includes five lung function tests, which involve breathing only cabin air. For each planned extravehicular (EVA) activity, a crew member performs a PuFF test within one week prior to the EVA. Following the EVA, those crew members perform another test to document the effect of exposure of the lungs to the low-pressure environment of the space suits. This experiment utilizes the Gas Analyzer System for Metabolic Analysis Physiology, or GASMAP, located in the Human Research Facility (HRF), along with a variety of other Puff equipment including a manual breathing valve, flow meter, pressure-flow module, pressure and volume calibration syringes, and disposable mouth pieces.

  8. Combined effects of Fe(II) and oxidizing radiolysis products on UO2 and PuO2 dissolution in a system containing solid UO2 and PuO2

    NASA Astrophysics Data System (ADS)

    Amme, Marcus; Pehrman, Reijo; Deutsch, Rudolf; Roth, Olivia; Jonsson, Mats

    2012-11-01

    The stability of UO2 spent nuclear fuel in an oxygen-free geological repository depends on the absence of oxidizing reaction partners in the near field. This work investigates the reactions between the products of water radiolysis by alpha radiation and Fe(II) an the effect on UO2 dissolution. Solid 238PuO2 powder and UO2 pellet were allowed to react in Fe(II) solution in oxygen-free batch reactor tests and kinetics of the subsequent redox reactions were measured. Depending on the concentration of Fe(II) (tests with 10-5 and 10-4 mol L-1 were made), the induced redox reactions took place between 20 and 400 h. Dissolved uranium concentrations went first through a minimum caused by reduction, followed by a maximum caused by radiolytic oxidation, and eventually reached another minimum, probably due to sorption on precipitated Fe(III). Plutonium concentrations were decreasing steadily after going through a maximum about 70 h from the start of the experiments. The results show that in the presence of the strong alpha-radiolytic field induced by the presence of solid 238Pu, the behavior of the system is largely governed by Fe(II) as it controls the H2O2 concentration, reduces U(VI) in solution and drives the Fenton reaction leading to the oxidation of Pu(IV).

  9. The inflow of 238Pu and (239+240)Pu from the Odra and Pomeranian rivers catchments area to the Baltic Sea.

    PubMed

    Strumińska-Parulska, Dagmara I; Skwarzec, Bogdan; Tuszkowska, Agnieszka

    2012-11-01

    The aim of the work was to estimate plutonium inflow from the Odra River catchments area to the Baltic Sea. The highest activities of (238)Pu and (239+240)Pu were observed in a winter and a spring season. The highest annual surface inflow of (239+240)Pu from the Odra River watershed was observed for a mountain tributary the Bóbr (1230 Bq km(-2) year(-1)). The annual inflow of (238)Pu and (239+240)Pu to the Baltic Sea was estimated at 9.51 MBq and 45.86 MBq respectively and the highest plutonium surface runoff was observed for the Bóbr drainage.

  10. Characterization of Pu concentration and its isotopic composition in a reference fallout material.

    PubMed

    Zhang, Yongsan; Zheng, Jian; Yamada, Masatoshi; Wu, Fengchang; Igarashi, Yasuhito; Hirose, Katsumi

    2010-02-01

    Because there is no reference material for fallout plutonium isotope monitoring, preparation of such a material is necessary for quality control of fallout radionuclides analysis for atmospheric environmental studies. In this work, we report the characterization of Pu activity and its isotopic composition in a reference fallout material prepared by the Meteorological Research Institute (MRI), Japan. This material was prepared from samples collected at 14 stations throughout Japan in 1963-1979, with reference values of (137)Cs, (90)Sr and (239)(+)(240)Pu activities. We analyzed the activities of (239)(+)(240)Pu and (241)Pu, and the atom ratios of (240)Pu/(239)Pu and (241)Pu/(239)Pu using an isotope dilution sector-field inductively coupled plasma mass spectrometry (SF-ICP-MS). The (239)(+)(240)Pu activities in this fallout material using acid leaching and total digestion were 6.56+/-0.20 mBq/g and 6.79+/-0.16 mBq/g, respectively. Atom ratios of (240)Pu/(239)Pu were 0.1915+/-0.0030 and 0.1922+/-0.0044, respectively. Both (240)Pu/(239)Pu and (241)Pu/(239)Pu atom ratios were slightly higher than those of global fallout, which could be attributed to the deposition of fallout radionuclides resulting from the Chinese nuclear weapons tests conducted in the 1970s. The dominant host phases of (239)(+)(240)Pu were found to be organic matter-sulfides (70%) with a relative high (240)Pu/(239)Pu atom ratio, and Fe-Mn oxides (19%) using a sequential extraction method.

  11. An evaluation of alternate production methods for Pu-238 general purpose heat source pellets

    SciTech Connect

    Mark Borland; Steve Frank

    2009-06-01

    For the past half century, the National Aeronautics and Space Administration (NASA) has used Radioisotope Thermoelectric Generators (RTG) to power deep space satellites. Fabricating heat sources for RTGs, specifically General Purpose Heat Sources (GPHSs), has remained essentially unchanged since their development in the 1970s. Meanwhile, 30 years of technological advancements have been made in the applicable fields of chemistry, manufacturing and control systems. This paper evaluates alternative processes that could be used to produce Pu 238 fueled heat sources. Specifically, this paper discusses the production of the plutonium-oxide granules, which are the input stream to the ceramic pressing and sintering processes. Alternate chemical processes are compared to current methods to determine if alternative fabrication processes could reduce the hazards, especially the production of respirable fines, while producing an equivalent GPHS product.

  12. Neutron radiation characteristics of plutonium dioxide fuel

    NASA Technical Reports Server (NTRS)

    Taherzadeh, M.

    1972-01-01

    The major sources of neutrons from plutonium dioxide nuclear fuel are considered in detail. These sources include spontaneous fission of several of the Pu isotopes, (alpha, n) reactions with low Z impurities in the fuel, and (alpha, n) reactions with O-18. For spontaneous fission neutrons a value of (1.95 + or - 0.07) X 1,000 n/s/g PuO2 is obtained. The neutron yield from (alpha, n) reactions with oxygen is calculated by integrating the reaction rate equation over all alpha-particle energies and all center-of-mass angles. The results indicate a neutron emission rate of (1.14 + or - 0.26) X 10,000 n/s/g PuO2. The neutron yield from (alpha, n) reactions with low Z impurities in the fuel is presented in tabular form for one part part per million of each impurity. The total neutron yield due to the combined effects of all the impurities depends upon the fractional weight concentration of each impurity. The total neutron flux emitted from a particular fuel geometry is estimated by adding the neutron yield due to the induced fission to the other neutron sources.

  13. Neutron Radiation Characteristics of Plutonium Dioxide Fuel

    NASA Technical Reports Server (NTRS)

    Taherzadeh, M.

    1972-01-01

    The major sources of neutrons from plutonium dioxide nuclear fuel are considered in detail. These sources include spontaneous fission of several of the Pu isotopes, reactions with low Z impurities in the fuel, and reactions with O-18. For spontaneous fission neutrons a value of (1.95 plus or minus 0.07) X 1,000 n/s/q PuO2 is obtained. The neutron yield from (alpha, neutron) reactions with oxygen is calculated by integrating the reaction rate equation over all alpha particle energies and all center-of-mass angles. The results indicate a neutron emission rate of (1.42 plus or minus 0.32) X 10,000 n/s/q PuO2. The neutron yield from (alpha, neutron) reactions with low Z impurities in the fuel is presented in tabular form for one part per million of each impurity. The total neutron flux emitted from a particular fuel geometry is estimated by adding the neutron yield due to the induced fission to the other neutron sources.

  14. ICE1 of Pyrus ussuriensis functions in cold tolerance by enhancing PuDREBa transcriptional levels through interacting with PuHHP1

    PubMed Central

    Huang, Xiaosan; Li, Kongqing; Jin, Cong; Zhang, Shaoling

    2015-01-01

    ICE1 transcription factor plays an important role in plant cold stress via regulating the expression of stress-responsive genes. In this study, a PuICE1 gene isolated from Pyrus ussuriensis was characterized for its function in cold tolerance. The expression levels of the PuICE1 were induced by cold, dehydration and salt, with the greatest induction under cold conditions. PuICE1 was localized in the nucleus and could bind specifically to the MYC element in the PuDREBa promoter. The PuICE1 fused to the GAL4 DNA-binding domain to have transcriptional activation activity. Ectopic expression of the PuICE1 in tomato conferred enhanced tolerance to cold stress at cold temperatures, less electrolyte leakage, less MDA content, higher chlorophyll content, higher survival rate, higher proline content, higher activities of enzymes. In additon, steady-state mRNA levels of six stress-responsive genes coding for either functional or regulatory genes were induced to higher levels in the transgenic lines by cold stress. Yeast two-hybrid, transient assay, split luciferase complementation and BiFC assays all revealed that PuHHP1 protein can physically interact with PuICE1. Taken together, these results demonstrated that PuICE1 plays a positive role in cold tolerance, which may be due to enhancement of PuDREBa transcriptional levels through interacting with the PuHHP1. PMID:26626798

  15. Formation of crystalline PuO2+x·nH2O nanoparticles upon sorption of Pu(V,VI) onto hematite

    NASA Astrophysics Data System (ADS)

    Romanchuk, Anna Yu.; Kalmykov, Stepan N.; Egorov, Alexander V.; Zubavichus, Yan V.; Shiryaev, Andrey A.; Batuk, Olga N.; Conradson, Steven D.; Pankratov, Denis A.; Presnyakov, Igor A.

    2013-11-01

    It has been recognized that natural aquatic colloids can readily sorb actinide elements, including plutonium, whose behavior is complicated by its multiple valence states and the possibility of redox reactions under environmental conditions. In this paper, the sorption and surface-mediated redox transformations of hexavalent plutonium on synthetic well-characterized hematite colloids are studied in a series of batch sorption experiments. The variation in the kinetics of the Pu-hematite interactions, Pu-L3-XAFS, and HRTEM over a broad range of total concentrations of Pu have been studied in an attempt to define the molecular-level speciation of Pu. The surface-mediated slow reduction of Pu(V/VI) results in the formation of crystalline nanoparticles of PuO2+x·nH2O approximately 1.5 nm in size at [Pu]tot ⩾ 10-9 M. This result is confirmed independently by HRTEM images of Pu-containing particles and through the identification in the EXAFS of a Pu neighbor shell at 3.8 Å. The formation of such nanoparticles potentially influences the colloid-mediated transport of Pu in the subsurface environment because of the very slow leaching of Pu from the hematite colloids.

  16. ICE1 of Pyrus ussuriensis functions in cold tolerance by enhancing PuDREBa transcriptional levels through interacting with PuHHP1

    NASA Astrophysics Data System (ADS)

    Huang, Xiaosan; Li, Kongqing; Jin, Cong; Zhang, Shaoling

    2015-12-01

    ICE1 transcription factor plays an important role in plant cold stress via regulating the expression of stress-responsive genes. In this study, a PuICE1 gene isolated from Pyrus ussuriensis was characterized for its function in cold tolerance. The expression levels of the PuICE1 were induced by cold, dehydration and salt, with the greatest induction under cold conditions. PuICE1 was localized in the nucleus and could bind specifically to the MYC element in the PuDREBa promoter. The PuICE1 fused to the GAL4 DNA-binding domain to have transcriptional activation activity. Ectopic expression of the PuICE1 in tomato conferred enhanced tolerance to cold stress at cold temperatures, less electrolyte leakage, less MDA content, higher chlorophyll content, higher survival rate, higher proline content, higher activities of enzymes. In additon, steady-state mRNA levels of six stress-responsive genes coding for either functional or regulatory genes were induced to higher levels in the transgenic lines by cold stress. Yeast two-hybrid, transient assay, split luciferase complementation and BiFC assays all revealed that PuHHP1 protein can physically interact with PuICE1. Taken together, these results demonstrated that PuICE1 plays a positive role in cold tolerance, which may be due to enhancement of PuDREBa transcriptional levels through interacting with the PuHHP1.

  17. U, Pu, and Am nuclear signatures of the Thule hydrogen bomb debris.

    PubMed

    Eriksson, Mats; Lindahl, Patric; Roos, Per; Dahlgaard, Henning; Holm, Elis

    2008-07-01

    This study concerns an arctic marine environment that was contaminated by actinide elements after a nuclear accident in 1968, the so-called Thule accident In this study we have analyzed five isolated hot particles as well as sediment samples containing particles from the weapon material for the determination of the nuclear fingerprint of the accident. We report that the fissile material in the hydrogen weapons involved in the Thule accident was a mixture of highly enriched uranium and weapon-grade plutonium and that the main fissile material was 235U (about 4 times more than the mass of 239Pu). In the five hot particles examined, the measured uranium atomic ratio was 235U/238U = 1.02 +/- 0.16 and the Pu-isotopic ratios were as follows: 24Pu/239Pu = 0.0551 +/- 0.0008 (atom ratio), 238Pu/239+240Pu = 0.0161 +/- 0.0005 (activity ratio), 241Pu/239+240Pu = 0.87 +/- 0.12 (activity ratio), and 241Am/ 239+240Pu = 0.169 +/- 0.005 (activity ratio) (reference date 2001-10-01). From the activity ratios of 241Pu/241Am, we estimated the time of production of this weapon material to be from the late 1950s to the early 1960s. The results from reanalyzed bulk sediment samples showed the presence of more than one Pu source involved in the accident, confirming earlier studies. The 238Pu/239+240PU activity ratio and the 240Pu/ 239Pu atomic ratio were divided into at least two Pu-isotopic ratio groups. For both Pu-isotopic ratios, one ratio group had identical ratios as the five hot particles described above and for the other groups the Pu isotopic ratios were lower (238Pu/ 239+240PU activity ratio approximately 0.01 and the 240Pu/P239Pu atomic ratio 0.03). On the studied particles we observed that the U/Pu ratio decreased as a function of the time these particles were present in the sediment. We hypothesis that the decrease in the ratio is due to a preferential leaching of U relative to Pu from the particle matrix.

  18. Radiation dose aspects in the handling of emerging nuclear fuels.

    PubMed

    Nicolaou, G

    2014-12-01

    The occupational annual dose levels, encountered at fabrication of emerging nuclear fuels, have been studied. Emerging fuels for the single and multiple recycling of Pu and MA have resulted in considerably higher gamma and neutron doses in comparison with commercial fuels. The occupational dose limit is exceeded at fabrication by a single fuel rod in all fuel cases with (241)Am and Cm isotopes present in their composition. In the absence of these isotopes, 2-4 adjacent fuel rods are sufficient to exceed the limit. Self-shielding within the fuel reduces significantly only the gamma dose that would have been delivered otherwise. Hence, only the first row of fuel rods in an assembly contributes to the dose, whereas in the case of neutrons, all fuel rods contribute.

  19. 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.

  20. Radiation Damage Effects in Candidate Titanates for Pu Disposition: Zirconolite

    SciTech Connect

    Strachan, Denis M.; Scheele, Randall D.; Buck, Edgar C.; Kozelisky, Anne E.; Sell, Rachel L.; Elovich, Robert J.; Buchmiller, William C.

    2008-01-15

    Specimens of titanate ceramics containing approximately 10 mass% 238Pu were tested to determine the long-term effects of radiation-induced damage from the α decay of 239Pu that would have been disposed of in the nuclear-waste repository at Yucca Mountain. These tests provided information on the changes in bulk properties such as dimensions, densities, and chemical durability. Although these materials become amorphous at low doses, the specimens remained physically strong. Even after the radiation-induced swelling saturated, the specimens remained physically intact with no evidence for microcracking. Thus, in combination with results reported previously on similar materials, the material remains a physically viable material for the disposition of surplus weapons-grade Pu.

  1. Transuranic concentrations in reef and pelagic fish from the Marshall Islands. [/sup 239/Pu, /sup 240/Pu

    SciTech Connect

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

    1980-09-01

    Concentrations of /sup 239 + 240/Pu are reported in tissues of several species of reef and pelagic fish caught at 14 different atolls in the northern Marshall Islands. Several regularities that are species dependent are evident in the distribution of /sup 239 + 240/Pu among different body tissues. Concentrations in liver always exceeded those in bone and concentrations were lowest in the muscle of all fish analyzed. A progressive discrimination against /sup 239 + 240/Pu was observed at successive trophic levels at all atolls except Bikini and Enewetak, where it was difficult to conclude if any real difference exists between the average concentration factor for /sup 239 + 240/Pu among all fish, which include bottom feeding and grazing herbivores, bottom feeding carnivores, and pelagic carnivores from different atoll locations. The average concentration of /sup 239 + 240/Pu in the muscle of surgeonfish from Bikini and Enewetak was not significantly different from the average concentrations determined in these fish at the other, lesser contaminated atolls. Concentrations among all 3rd, 4th, and 5th trophic level species are highest at Bikini where higher environmental concentrations are found. The reasons for the anomalously low concentrations in herbivores from Bikini and Enewetak are not known.

  2. Deducing the 236Pu(n,f) and 237Pu(n,f) cross sections via the surrogate ratio method

    NASA Astrophysics Data System (ADS)

    Hughes, R. O.; Beausang, C. W.; Ross, T. J.; Burke, J. T.; Casperson, R. J.; McCleskey, M.; Cooper, N.; Escher, J. E.; Gell, K. B.; Good, E.; Humby, P.; Saastimoinen, A.; Tarlow, T. D.

    2013-10-01

    The short half-lives associated with certain minor actinide nuclei that are relevant to stockpile stewardship pursuits and the development of next-generation nuclear reactors make direct neutron measurements very challenging. In certain cases, a stable beam and target ``surrogate reaction'' can be used in lieu of the neutron-induced reaction, and the (n,f) cross section can then be deduced indirectly. Agreement between surrogate and direct measurements for (n,f) cross sections in actinide nuclei is usually within 10%. The present work reports on the measurement of the 236Pu(n,f) and 237Pu(n,f) cross sections via 239Pu(p,tf) and 239Pu(p,df) surrogate reactions, respectively. The experiment was performed at the Texas A&M University Cyclotron Facility using a 28.5 MeV proton beam to bombard 239Pu and 235U targets. Outgoing light ions were detected in coincidence with fission fragments using the STAR-LiTe detector array. Results of the analysis will be presented. This work was supported by DoE Grant Numbers: DE-FG52-09NA29454 and DE-FG02-05ER41379 (Richmond), DE-AC52-07NA27344 (LLNL) and DE-FG52-09NA29467 (TAMU).

  3. Pu236(n,f), Pu237(n,f), and Pu238(n,f) cross sections deduced from (p,t), (p,d), and (p,p') surrogate reactions

    NASA Astrophysics Data System (ADS)

    Hughes, R. O.; Beausang, C. W.; Ross, T. J.; Burke, J. T.; Casperson, R. J.; Cooper, N.; Escher, J. E.; Gell, K.; Good, E.; Humby, P.; McCleskey, M.; Saastimoinen, A.; Tarlow, T. D.; Thompson, I. J.

    2014-07-01

    The Pu236(n,f), Pu237(n,f) and Pu238(n,f) cross sections have been inferred by utilizing the surrogate ratio method. Targets of Pu239 and U235 were bombarded with 28.5-MeV protons, and the light ion recoils, as well as fission fragments, were detected using the STARS detector array at the K150 Cyclotron at the Texas A&M cyclotron facility. The (p, tf) reaction on Pu239 and U235 targets was used to deduce the σ (Pu236(n ,f))/σ(U232(n,f)) ratio, and the Pu236(n,f) cross section was subsequently determined for En=0.5-7.5 MeV. Similarly, the (p,df) reaction on the same two targets was used to deduce the σ(Pu237(n ,f))/σ(U233(n,f)) ratio, and the Pu237(n,f) cross section was extracted in the energy range En=0.5-7 MeV. The Pu238(n,f) cross section was also deduced by utilizing the (p,p') reaction channel on the same targets. There is good agreement with the recent ENDF/B-VII.1 evaluated cross section data for Pu238(n,f) in the range En=0.5-10.5 MeV and for Pu237(n,f) in the range En=0.5-7 MeV; however, the Pu236(n,f) cross section deduced in the present work is higher than the evaluation between 2 and 7 MeV.

  4. Determination of Plutonium Activity Concentrations and 240Pu/239Pu Atom Ratios in Brown Algae (Fucus distichus) Collected from Amchitka Island, Alaska.

    SciTech Connect

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

    2005-05-02

    Plutonium-239 ({sup 239}Pu) and plutonium-240 ({sup 240}Pu) activity concentrations and {sup 240}Pu/{sup 239}Pu atom ratios are reported for Brown Algae (Fucus distichus) collected from the littoral zone of Amchitka Island (Alaska) and at a control site on the Alaskan peninsula. Plutonium isotope measurements were performed in replicate using Accelerator Mass Spectrometry (AMS). The average {sup 240}Pu/{sup 239}Pu atom ratio observed in dried Fucus d. collected from Amchitka Island was 0.227 {+-} 0.007 (n=5) and compares with the expected {sup 240}Pu/{sup 239}Pu atom ratio in integrated worldwide fallout deposition in the Northern Hemisphere of 0.1805 {+-} 0.0057 (Cooper et al., 2000). In general, the characteristically high {sup 240}Pu/{sup 239}Pu content of Fucus d. analyzed in this study appear to indicate the presence of a discernible basin-wide secondary source of plutonium entering the marine environment. Of interest to the study of plutonium source terms within the Pacific basin are reports of elevated {sup 240}Pu/{sup 239}Pu atom ratios in fallout debris from high-yield atmospheric nuclear tests conducted in the Marshall Islands during the 1950s (Diamond et al., 1960), the wide range of {sup 240}Pu/{sup 239}Pu atom ratio values (0.19 to 0.34) observed in sea water, sediments, coral and other environmental media from the North Pacific Ocean (Hirose et al., 1992; Buesseler, 1997) and updated estimates of the relative contributions of close-in and intermediate fallout deposition on oceanic inventories of radionuclidies, especially in the Northern Pacific Ocean (Hamilton, 2004).

  5. Economical Production of Pu-238: NIAC Phase I Final Report

    NASA Technical Reports Server (NTRS)

    Howe, Steven D.; Crawford, Douglas; Navarro, Jorge; O'Brien, Robert C.; Katalenich, Jeff; Ring, Terry

    2016-01-01

    All space exploration missions traveling beyond Jupiter must use radioisotopic power sources for electrical power. The best isotope to power these sources is plutonium-238 (Pu-238). The US supply of Pu-238 is almost exhausted and will be gone within the next decade. The Department of Energy has initiated a production program with a $10M allocation from NASA but the cost is estimated at over $100M to get to production levels. The Center for Space Nuclear Research (CSNR) has conceived of a potentially better process to produce Pu-238 earlier and for significantly less cost. Potentially, the front end capital costs could be provided by private industry such that the government only had to pay for the product produced. In the Phase I NIAC (NASA Innovative Advanced Concepts) grant, the CSNR has evaluated the feasibility of using a low power, commercially available nuclear reactor to produce 1.5 kg of Pu-238 per year. The impact on the neutronics of the reactor have been assessed, the amount of Neptunium target material estimated, and the production rates calculated. In addition, the size of the post-irradiation processing facility has been established. Finally, as the study progressed, a new method for fabricating the Pu-238 product into the form used for power sources has been identified to reduce the cost of the final product. In short, the concept appears to be viable, can produce the amount of Pu-238 needed to support the NASA missions, can be available within a few years, and will cost significantly less than the current DOE program.

  6. Complexation of Pu(IV) with the natural siderophore desferrioxamine B and the redox properties of Pu(IV)(siderophore) complexes.

    PubMed

    Boukhalfa, Hakim; Reilly, Sean D; Neu, Mary P

    2007-02-05

    The bioavailability and mobility of Pu species can be profoundly affected by siderophores and other oxygen-rich organic ligands. Pu(IV)(siderophore) complexes are generally soluble and may constitute with other soluble organo-Pu(IV) complexes the main fraction of soluble Pu(IV) in the environment. In order to understand the impact of siderophores on the behavior of Pu species, it is important to characterize the formation and redox behavior of Pu(siderophore) complexes. In this work, desferrioxamine B (DFO-B) was investigated for its capacity to bind Pu(IV) as a model siderophore and the properties of the complexes formed were characterized by optical spectroscopy measurements. In a 1:1 Pu(IV)/DFO-B ratio, the complexes Pu(IV)(H2DFO-B)4+, Pu(IV)(H1DFO-B)3+, Pu(IV)(DFO-B)2+, and Pu(IV)(DFO-B)(OH)+ form with corresponding thermodynamic stability constants log beta1,1,2 = 35.48, log beta1,1,1 = 34.87, log beta1,1,0 = 33.98, and log beta1,1,-1 = 27.33, respectively. In the presence of excess DFO-B, the complex Pu(IV)H2(DFO-B)22+ forms with the formation constant log beta2,1,2 = 62.30. The redox potential of the complex Pu(IV)H2(DFO-B)22+ was determined by cyclic voltammetry to be E1/2 = -0.509 V, and the redox potential of the complex Pu(IV)(DFO-B)2+ was estimated to be E1/2 = -0.269 V. The redox properties of Pu(IV)(DFO-B)2+ complexes indicate that Pu(III)(siderophore) complexes are more than 20 orders of magnitude less stable than their Pu(IV) analogues. This indicates that under reducing conditions, stable Pu(siderophore) complexes are unlikely to persist.

  7. Calculated lattice relaxation in Pu-Ga alloys

    SciTech Connect

    Becker, J.D.; Wills, J.M.; Cox, L.; Cooper, B.R.

    1997-12-01

    Hellman-Feynman forces on atoms surrounding the gallium site in a Pu{sub 31}Ga supercell are calculated with the full-potential LMTO method in the local density approximation. These forces are minimized by adjusting atomic positions using an iterative Broyden scheme. At equilibrium the nearest-neighbor shell of plutonium atoms relaxes inward by 1.04% of the initial theoretical bond length (2.86 {angstrom}). A similar calculation on a Pu{sub 32} supercell shows no relaxation.

  8. Thermal conductivity of (Np0.20Pu0.50Am0.25Cm0.05)O2-x solid solutions

    NASA Astrophysics Data System (ADS)

    Nishi, Tsuyoshi; Takano, Masahide; Akabori, Mitsuo; Arai, Yasuo

    2013-09-01

    recovery behavior by annealing. The thermal conductivity of (Np0.20Pu0.50Am0.25Cm0.05)O2-x was smaller than those of PuO2 and (Pu0.91Cm0.09)O2 mainly because of the oxygen vacancies as is seen other actinide dioxide, such as mixed oxide (MOX) fuels.

  9. Reductive Disslocation of Pu(IV) by Clostridium sp. Under Anaerobic Conditions

    SciTech Connect

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

    2008-01-01

    An anaerobic, gram positive, spore-forming bacterium Clostridium sp., common in soils and wastes, capable of reduction of Fe(III) to Fe(II), Mn(IV) to Mn(II), Tc(VII) to Tc(IV), and U(VI) to U(IV), reduced Pu(IV) to Pu(III). Addition of 242Pu (IV)-nitrate to the bacterial growth medium at pH 6.4 resulted in the precipitation of Pu as amorphous Pu(OH)4 due to hydrolysis and polymerization reactions. The Pu (1 x 10-5 M) had no effect upon growth of the bacterium as evidenced by glucose consumption; carbon dioxide and hydrogen production; a decrease in pH of the medium from 6.4 to 3.0 due to production of acetic and butyric acids from glucose fermentation; and a change in the Eh of the culture medium from +50 to -180 mV. Commensurate with bacterial growth, Pu was rapidly solubilized as evidenced by an increase in Pu concentration in solution which passed through a 0.03 {mu}m filtration. Selective solvent extraction of the culture by thenoyltrifluoroacetone (TTA) indicated the presence of a reduced Pu species in the soluble fraction. X-ray absorption near edge spectroscopic (XANES) analysis of Pu in the culture sample at the Pu LIII absorption edge (18.054 keV) showed a shift of -3 eV compared to a Pu(IV) standard indicating reduction of Pu(IV) to Pu(III). These results suggest that, although Pu generally exists as insoluble Pu(IV) in the environment, under appropriate conditions, anaerobic microbial activity could affect the long-term stability and mobility of Pu by its reductive dissolution.

  10. Accumulation of 242Pu by a macrophyte of the Yenisei River (Elodea canadensis) in laboratory experiments.

    PubMed

    Bolsunovsky, A; Bondareva, L; Sukhorukov, F; Melgunov, M

    2009-04-01

    The study addresses 242Pu accumulation by Elodea canadensis, one of the abundant species of submerged plants in the Yenisei River. 242Pu in water samples of the "Elodea-Yenisei River water" model system and in the biomass fractions was determined alpha-spectrometrically, following radiochemical recovery of 242Pu using 236Pu--a chemical yield tracer. The experiments on accumulation of 242Pu by Elodea biomass showed that the activity concentration of 242Pu can reach 21 +/- 2 Bq/g dry wt, with the concentration factor for 242Pu 13100 +/- 2100 L/kg dry wt. Results of chemical fractionation proved that during the first few hours of the experiment 242Pu contained in Elodea was mainly concentrated in the exchangeable and the adsorbed fractions of biomass (about 100%). As Elodea biomass accumulated 242Pu, the absolute amount of 242Pu in the exchangeable and the adsorbed fractions remained almost unchanged, although the portion of 242Pu tightly bound to biomass increased. At the end of the experiment, on day 7, 242Pu tightly bound to biomass (fractions of organics and mineral residue) constituted 43-63% (in different experiments) of the total 242Pu in the biomass.

  11. Humic acids facilitated microbial reduction of polymeric Pu(IV) under anaerobic conditions.

    PubMed

    Xie, Jinchuan; Liang, Wei; Lin, Jianfeng; Zhou, Xiaohua; Li, Mei

    2017-08-25

    Flavins and humic substances have been extensively studied with emphasis on their ability to transfer extracellular electrons to insoluble metal oxides. Nevertheless, whether the low-solubility Pu(IV) polymers are microbially reduced to aqueous Pu(III) remains uncertain. Experiments were conducted under anaerobic and slightly alkaline conditions to study the difference between humic acids and flavins to transport extracellular electrons to Pu(IV) polymers. Our study demonstrates that Shewanella putrefaciens was unable to directly reduce polymeric Pu(IV) with a notably low reduction rate (3.4×10(-12)mol/L Pu(III)aq within 144h). The relatively high redox potential of flavins reveals the thermodynamically unfavorable reduction: Eh(PuO2(am)/Pu(3+))Pu(IV) (2.1×10(-10)mol/L Pu(III)aq) 62 times more rapidly than the flavins. The driving force for electron transfer explains the observed reduction: Eh(HAox/HAred)PuO2(am)/Pu(3+)) when S. putrefaciens oxidized lactate and respired on the humic acids. In contrast, flavins were able to substantially reduce aqueous Pu(IV)-EDTA (1.9×10(-9)mol/L Pu(III)aq) because of the available driving force for electron transfer: ΔrGm=-F[Eh(PuL2(4-)/PuL2(5-))-Eh(o)'(FMN/FMNH2)]=-33.5kJ/mol is a result of Eh(PuL2(4-)/PuL2(5-))≫Eh(PuO2(am)/Pu(3+)), where L is the EDTA ligand. In the presence of humic acids, the reduction of Pu(IV)-EDTA exhibited the most rapid rate (2.2×10(-9)mol/L Pu(III)aq). This result further demonstrates that humic acids facilitated the extracellular electron transfer to polymeric and aqueous Pu(IV). Reductive solubilization of the polymers may enhance Pu mobility in the geosphere and hence increases risks to human health. Copyright © 2017 Elsevier B.V. All rights reserved.

  12. Evidence of extinct 244Pu in ancient terrestrial zircons

    NASA Astrophysics Data System (ADS)

    Harrison, T. M.; Turner, G.; Holland, G.; Gilmour, J. D.; Mojzsis, S. J.

    2003-04-01

    The Pu/U ratio of the early Earth is an important parameter in models of mantle evolution based on noble gas isotopes. Current estimates assume the Earth accreted with a chondritic Pu/U and are based on analyses of the chondrite St Severin and the achondrite Angra dos Reis. These estimates are poorly constrained, ranging from 0.004 to 0.008. On account of its short, 82 Ma, half-life, 244Pu was essentially extinct 3,900 Ma ago, and consequently there exists no reliable measurement of Pu/U for the Earth. The discovery of zircons dating from the period when 244Pu was "live" offers the first opportunity to measure the former terrestrial abundance of 244Pu directly. Xenon isotopes are produced by spontaneous fission and, in principle, are readily distinguishable from those produced by 238U-fission (e.g. 131Xe/136Xe = 0.24 and 0.08 respectively). However the expected levels of fission xenon in individual zircons, weighing 1 - 2 μg and containing 100 - 200 ppm U, are below, or at best comparable to, the Xe blank levels (˜10-15 ccSTP) typical of conventional noble gas mass spectrometers. In order to analyse these minute amounts of xenon we have made use of a uniquely sensitive instrument, developed in Manchester, based on the principle of laser resonance ionisation. RELAX (Refrigerator Enhanced Laser Analyser for Xenon) is capable of analysing samples of only a few thousand atoms, some two orders of magnitude smaller than conventional mass spectrometers. We have carried out preliminary analyses of 4 individual 4,150 Ma zircons and one 3,600 Ma zircon from Jack Hills, Western Australia, and obtained five clear fission spectra. All but one were essentially free from significant atmospheric blank (the average 130Xe blank was 3× 10-18 ccSTP, i.e. 80 atoms). The spectra of the older zircons clearly demonstrated the presence of varying amounts of 244Pu fission xenon. The highest 131Xe/136Xe, 0.136 ± 0.003, corresponds to an initial Pu/U ratio of 0.0057. The lower ratios

  13. Fuel clad chemical interactions in fast reactor MOX fuels

    NASA Astrophysics Data System (ADS)

    Viswanathan, R.

    2014-01-01

    Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements (oxygen, cesium, tellurium, iodine) in the clad-attack are discussed and many Fuel-Clad-Chemical-Interaction (FCCI) models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL (Hanford Engineering Development Laboratory) relation is recommended: d/μm = ({0.507 ṡ [B/(at.% fission)] ṡ (T/K-705) ṡ [(O/M)i-1.935]} + 20.5) for (O/M)i ⩽ 1.98. A new model is proposed for (O/M)i ⩾ 1.98: d/μm = [B/(at.% fission)] ṡ (T/K-800)0.5 ṡ [(O/M)i-1.94] ṡ [P/(W cm-1)]0.5. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, (O/M)i is the initial oxygen-to-(uranium + plutonium) ratio, and P is the linear power rating. For fuels with [n(Pu)/n(M = U + Pu)] > 0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.

  14. Analysis of thorium-salted fuels to improve uranium utilization in the once-through fuel cycle

    SciTech Connect

    Eschbach, E.A.; Merrill, E.T.; Prichard, A.W.

    1981-09-01

    Calculations and analyses indicate that no improvement can be achieved in uranium utilization for the once-through LWR fuel cycle over use of slightly enriched uranium by employing thorium distributed with uranium. The study included thorium additions: (1) slight amounts, (2) larger amounts, in either intimately mixed or in duplex pellets, (3) in spectrally shifted or not spectrally shifted reactors, and (4) in three- or five-year reactivity limited exposures. While thorium-uranium combinations improves the initial conversion ratio, the reactivity lifetime was not extended enough to override the additional uranium required. The effective fission cross-section of the bred /sup 233/U relative to /sup 239/Pu's in typical LWR neutron spectra is not large enough for /sup 233/U to make as great a contribution to end-of-life reactivity as /sup 239/Pu in a slightly enriched uranium fuel element. /sup 233/U's reactivity contribution relative to /sup 239/Pu's is lower in fuel configurations such as slightly enriched uranium LWR fuel loads. On the other hand, /sup 233/U's reactivity contribution appears more positive for reactors that involve lower average concentrations of thermal neutron absorbers. If /sup 238/U-thorium fuels reprocessed, the recovered /sup 233/U would increase uranium utilization, but may not reduce fuel cycle costs. The thorium-salted fuels exhibit substantially flatter reactivity characteristics with exposure time. Spectral shift helped the utilization of uranium and thorium.

  15. Microautoradiography in studies of Pu(V) sorption by trace and fracture minerals in tuff

    SciTech Connect

    Vaniman, D.; Furlano, A.; Chipera, S.; Thompson, J.; Triay, I.

    1996-12-31

    Microautoradiography was used to evaluate the mineralogic basis of Pu(V) retention by tuffs from Yucca Mountain, Nevada. Altered orthopyroxenes and oxide minerals are associated with high Pu retention but are limited to specific stratigraphic horizons. A weaker but more general association of Pu with smectite occurs in most samples. Thin-sections that cross fractures allow comparative studies of Pu retention by fracture-lining versus matrix minerals. Using Ag metal in emulsions as a measure of underlying Pu concentration, electron-microprobe analysis can quantify Pu retention along fracture walls and provide mineral/mineral Pu retention factors. For smectite-lined microfractures in zeolitized tuff, the smectite/clinoptilolite Pu retention factor is >80.

  16. Biogeochemical Cycling and Environmental Stability of Pu Relevant to Long-Term Stewardship of DOE Sites

    SciTech Connect

    Santschi, Peter H.

    2006-06-01

    The overall objective of this proposed research is to understand the biogeochemical cycling of Pu in environments of interest to long-term DOE stewardship issues. Central to Pu cycling (transport initiation to immobilization) is the role of microorganisms. The hypothesis underlying this proposal is that microbial activity is the causative agent in initiating the mobilization of Pu in near-surface environments: through the transformation of Pu associated with solid phases, production of extracellular polymeric substances (EPS) carrier phases, and the creation of microenvironments. Also, 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 Pu transport retardation.

  17. Biogeochemical Cycling and Environmental Stability of Pu Relevant to Long-Term Stewardship of DOE Sites

    SciTech Connect

    Francis, Arokiasamy J.; Santschi, Peter H.; Honeyman, Bruce D.

    2005-06-01

    The overall objective of this proposed research is to understand the biogeochemical cycling of Pu in environments of interest to long-term DOE stewardship issues. Central to Pu cycling (transport initiation to immobilization) is the role of microorganisms. The hypothesis underlying this proposal is that microbial activity is the causative agent in initiating the mobilization of Pu in near-surface environments: through the transformation of Pu associated with solid phases, production of extracellular polymeric substances (EPS) carrier phases, and the creation of microenvironments. Also, 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 Pu transport retardation.

  18. Biogeochemical Cycling and Environmental Stability of Pu Relevant to Long-Term Stewardship of DOE Sites

    SciTech Connect

    Honeyman, Bruce D.; Francis, A.J.; Gillow, Jeffrey B.; Dodge, Cleveland J.; Santschi, Peter H.; Chin-Chang Hung; Diaz, Angelique; Tinnacher, Ruth; Roberts, Kimberly; Schwehr, Kathy

    2006-04-05

    The overall objective of this research is to understand the biogeochemical cycling of Pu in environments of interest to long-term DOE stewardship issues. Central to Pu cycling (transport initiation and immobilization) is the role of microorganisms. The hypothesis underlying this work is that microbial activity is the causative agent in initiating the mobilization of Pu in near-surface environments: through the transformation of Pu associated with solid phases, production of extracellular polymeric substances (EPS) carrier phases and the creation of microenvironments. Also, 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 Pu transport retardation.

  19. Global fallout Pu recorded in lacustrine sediments in Lake Hongfeng, SW China.

    PubMed

    Zheng, Jian; Wu, Fengchang; Yamada, Masatoshi; Liao, Haiqing; Liu, Congqiang; Wan, Guojiang

    2008-03-01

    Studies on the distribution and isotope compositions of fallout Pu are important for source characterization of possible future non-fallout Pu contamination in aquatic environments, and useful for dating of recent sediments to understand the pollution history of environmental contaminants. We present the historical record of atmospheric Pu fallout reconstructed from a sediment core from Lake Hongfeng, China. The Pu activity profile was in agreement with the 137Cs profile. Inventories were 50.7 Bq m(-2) for 239+240Pu and 1586 Bq m(-2) for 137Cs. The average 240Pu/239Pu atom ratio was 0.185+/-0.009, indicating that Pu originated from global stratospheric fallout rather than from direct tropospheric or close-in fallout from the Chinese nuclear testing conducted in the 1970s. Our data suggested that Lake Hongfeng would be an ideal setting for monitoring atmospheric fallout and environmental changes in this region.

  20. Enhancing VVER Annular Proliferation Resistance Fuel with Minor Actinides

    SciTech Connect

    G. S. Chang

    2007-06-01

    Key aspects of the Global Nuclear Energy Partnership (GNEP) are to significantly advance the science and technology of nuclear energy systems and the Advanced Fuel Cycle (AFC) program. The merits of nuclear energy are the high-density energy, and low environmental impacts i.e. almost zero greenhouse gas emission. Planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current LWR as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The challenges are solving the energy needs of the world, protection against nuclear proliferation, the problem of nuclear waste, and the global environmental problem. To reduce the spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu and 240Pu isotopes ratio to enhance the proliferation resistance, (b) use of transuranic nuclides (237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope 238Pu /Pu ratio. For future advanced nuclear systems, the minor actinides are viewed more as a resource to be recycled, or transmuted to less hazardous and possibly more useful forms, rather than simply as a waste stream to be disposed of in expensive repository facilities. In this paper, a typical pressurized water reactor (PWR) VVER-1000 annular fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. We concluded that the concept of MARA, involves the use of transuranic nuclides (237Np and/or 241Am), can not only drastically

  1. Fossil fuels -- future fuels

    SciTech Connect

    1998-03-01

    Fossil fuels -- coal, oil, and natural gas -- built America`s historic economic strength. Today, coal supplies more than 55% of the electricity, oil more than 97% of the transportation needs, and natural gas 24% of the primary energy used in the US. Even taking into account increased use of renewable fuels and vastly improved powerplant efficiencies, 90% of national energy needs will still be met by fossil fuels in 2020. If advanced technologies that boost efficiency and environmental performance can be successfully developed and deployed, the US can continue to depend upon its rich resources of fossil fuels.

  2. 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.

  3. Post irradiation examination of thermal reactor fuels

    NASA Astrophysics Data System (ADS)

    Sah, D. N.; Viswanathan, U. K.; Ramadasan, E.; Unnikrishnan, K.; Anantharaman, S.

    2008-12-01

    The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (˜15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ˜11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

  4. Lead Slowing-Down Spectrometry Time Spectral Analysis for Spent Fuel Assay: FY11 Status Report

    SciTech Connect

    Kulisek, Jonathan A.; Anderson, Kevin K.; Bowyer, Sonya M.; Casella, Andrew M.; Gesh, Christopher J.; Warren, Glen A.

    2011-09-30

    Developing a method for the accurate, direct, and independent assay of the fissile isotopes in bulk materials (such as used fuel) from next-generation domestic nuclear fuel cycles is a goal of the Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign. To meet this goal, MPACT supports a multi-institutional collaboration, of which PNNL is a part, to study the feasibility of Lead Slowing Down Spectroscopy (LSDS). This technique is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic masses in used fuel with an uncertainty considerably lower than the approximately 10% typical of today's confirmatory assay methods. This document is a progress report for FY2011 PNNL analysis and algorithm development. Progress made by PNNL in FY2011 continues to indicate the promise of LSDS analysis and algorithms applied to used fuel. PNNL developed an empirical model based on calibration of the LSDS to responses generated from well-characterized used fuel. The empirical model, which accounts for self-shielding effects using empirical basis vectors calculated from the singular value decomposition (SVD) of a matrix containing the true self-shielding functions of the used fuel assembly models. The potential for the direct and independent assay of the sum of the masses of 239Pu and 241Pu to within approximately 3% over a wide used fuel parameter space was demonstrated. Also, in FY2011, PNNL continued to develop an analytical model. Such efforts included the addition of six more non-fissile absorbers in the analytical shielding function and the non-uniformity of the neutron flux across the LSDS assay chamber. A hybrid analytical-empirical approach was developed to determine the mass of total Pu (sum of the masses of 239Pu, 240Pu, and 241Pu), which is an important quantity in safeguards. Results using this hybrid method were of approximately the same accuracy as the pure

  5. Biophysical aspects of Am-241 and Pu-241 in the environment.

    PubMed

    Holm, E; Persson, R B

    1978-10-12

    Most of plutonium released by nuclear explosions is Pu-241 which decays to Am-241. We have studied the deposition of Pu-241 and Am-241 in lichens collected since 1958 in the central part of Sweden (62.3 degrees N, 12. 4 degrees E). Comparative studies with Pu-isotopes, Pu-239 + 240 and Pu-238 were also performed. In 1972 the total accumulated deposition of Pu-241 was 8 mCi/km2 of Pu-239 + 240 1 mCi/km2 and of Am-241 0.2 mCi/km2. About 80% of the Am-241 activity has been formed in situ from decay of Pu-241. The biological mean-residence time for all Pu-isotopes were about 6 years and for Am-241 4 years. The spatial distribution of Am-241 in the lichen carpet is quite different from that of Pu-241. The activity concentrations of Am-241 and Pu-241 have been studied in reindeer liver and bone. The average concentrations found were in liver 0.6 and 40 pCi/kg, in bone 0.2 and 6 pCi per kg for Am-241 and Pu-241 respectively. The activity content of Am-241 and Pu-241 in the Lapps due to their reindeer diet was estimated to be in liver 1.0 E-4 and 1.0 E-2 pCi/kg, in bone (3-9) E-5 and 1.0 E-2 pCi/kg for Am-241 and Pu-241 respectively. The estimated values for the fractions of ingested activity retained were in liver 7 E-6 and 14 E-6, in bone 20 E-6 for Am-241 and Pu-241 respectively. The fraction of ingested activity of Pu retained in reindeer liver is about 2-3 times higher than that of Am.

  6. XAFS and LIBD Investigation of the Formation and Structure of Pu(IV) Hydrolysis Products

    SciTech Connect

    Rothe, J.; Walther, C.; Denecke, M.A.; Fanghänel, Th.

    2010-11-16

    Pu(IV) oxyhydroxide colloid growth is investigated with XAFS and LIBD. From combined results a model of colloid formation is proposed, which leads to a face-centered cubic Pu sublattice having cation defects, as observed with EXAFS, and a linear dependency of log [Pu(IV)] on -log [H{sup +}] with slope -2, in accord with LIBD. The solubility for Pu(IV) measured with LIBD is close to the lower limit of the solubility curve from previously reported data.

  7. DENSITY-FUNCTIONAL STUDY OF Zr-BASED ACTINIDE ALLOYS: 2. U-Pu-Zr SYSTEM

    SciTech Connect

    Landa, A; Soderlind, P; Turchi, P; Vitos, L; Ruban, A

    2009-02-09

    Density-functional theory, previously used to describe phase equilibria in the U-Zr alloys [1], is applied to study ground state properties of the bcc U-Pu-Zr solid solutions. Calculated heats of formation of the Pu-U and Pu-Zr alloys are in a good agreement with CALPHAD assessments. We found that account for spin-orbit coupling is important for successful description of Pu-containing alloys.

  8. Ultra-trace determination of (90)Sr, (137)Cs, (238)Pu, (239)Pu, and (240)Pu by triple quadruple collision/reaction cell-ICP-MS/MS: Establishing a baseline for global fallout in Qatar soil and sediments.

    PubMed

    Amr, Mohamed A; Helal, Abdul-Fattah I; Al-Kinani, Athab T; Balakrishnan, Perumal

    2016-03-01

    The development of practical, fast, and reliable methods for the ultra-trace determination of anthropogenic radionuclides (90)Sr, (137)Cs, (238)Pu, (239)Pu, and (240)Pu by triple quadruple collision/reaction cell inductively coupled plasma mass spectrometry (CRC-ICP-MS/MS) were investigated in term of its accuracy and precision for producing reliable results. The radionuclides were extracted from 1 kg of the environmental soil samples by concentrated nitric and hydrochloric acids. The leachate solutions were measured directly by triple quadrupole CRC-ICP-MS/MS. For quality assurance, a chemical separation of the concerned radionuclides was conducted and then measured by single quadrupole-ICP-MS. The developed methods were next applied to measure the anthropogenic radionuclides (90)Sr, (137)Cs, (238)Pu, (239)Pu, and (240)Pu in soil samples collected throughout the State of Qatar. The average concentrations of (90)Sr, (137)Cs, (238)Pu, (239)Pu, and (240)Pu were 0.606 fg/g (3.364 Bq/kg), 0.619 fg/g (2.038 Bq/kg), 0.034 fg/g (0.0195 Bq/kg), 65.59 fg/g (0.150 Bq/kg), and 12.06 fg/g (0.103 Bq/kg), respectively.

  9. Fractionation of (137)Cs and Pu in natural peatland.

    PubMed

    Mihalík, Ján; Bartusková, Miluše; Hölgye, Zoltán; Ježková, Tereza; Henych, Ondřej

    2014-08-01

    High Cs-137 concentrations in plants growing on peatland inspired us to investigate the quantity of its bioavailable fraction in natural peat. Our investigation aims to: a) estimate the quantity of bioavailable Cs-137 and Pu present in peat, b) verify the similarity of Cs-137 and K-40 behaviours, and c) perform a quantification of Cs-137 and Pu transfer from peat to plants. We analysed the vertical distribution of Cs-137 and Pu isotopes in the peat and their concentrations in plants growing on these places. Bioavailability of radionuclides was investigated by sequential extraction. Sequential analyses revealed that it was the upper layer which contained the majority of Cs-137 in an available form while deeper layers retained Cs-137 in immobile fractions. We can conclude that 18% of all Cs-137 in the peat is still bioavailable. Despite of the low quantity of bioavailable fraction of Cs-137 its transfer factor reached extremely high values. In the case of Pu, 64% of its total amount was associated with fulvic/humic acids which resulted in the high transfer factor from peat to plants. 27 years after the Chernobyl nuclear accident, the significant part of radionuclides deposited in peatland is still bioavailable.

  10. Evaluation of the neutron cross sections for Pu-240

    SciTech Connect

    Weston, L.W.; Arthur, E.D.

    1987-04-01

    The present evaluation is proposed to supersede the ENDF/B-V, Revision 2 file for /sup 240/Pu. In this work, resonance parameters, cross sections, energy distributions, and angular distributions have been modified. These changes are outlined in detail and appropriate references included. 37 refs., 21 figs., 2 tabs.

  11. Selection and Evaluation of a new Pu Density Measurement Fluid

    SciTech Connect

    Dziewinska, Krystyna; Peters, Michael A; Martinez, Patrick P; Dziewinski, Jacek J; Pugmire, David L; Trujillo, Stephen M; La Verne, Jake A; Rajesh, P

    2009-01-01

    This paper summarizes efforts leading to selection of a new fluid for the determination of the density of large Pu parts. Based on an extended literature search, perfluorotributylamine (FC-43) was chosen for an experimental study. Plutonium coupon corrosion studies were performed by exposing Pu to deaerated and aerated solutions and measuring corrosion gravimetrically. Corrosion rates were determined. Samples of deaerated and aerated perfuluorotributylamine (FC-43) were also irradiated with {sup 60}Co gamma rays (96 Gy/min) to various doses. The samples were extracted with NaOH and analyzed by IC and showed the presence of F and Cl{sup -}. The G-values were established. In surface study experiments Pu coupons were exposed to deaerated and aerated solutions of FC-43 and analyzed by X-ray photoelectron spectroscopy (XPS). The XPS data indicate that there is no detectable surface effect caused by the new fluid. In conclusion the FC-43 was determined to be a very effective and practical fluid for Pu density measurements.

  12. Atomic Structure and Phase Transformations in Pu Alloys

    SciTech Connect

    Schwartz, A J; Cynn, H; Blobaum, K M; Wall, M A; Moore, K T; Evans, W J; Farber, D L; Jeffries, J R; Massalski, T B

    2008-04-28

    Plutonium and plutonium-based alloys containing Al or Ga exhibit numerous phases with crystal structures ranging from simple monoclinic to face-centered cubic. Only recently, however, has there been increased convergence in the actinides community on the details of the equilibrium form of the phase diagrams. Practically speaking, while the phase diagrams that represent the stability of the fcc {delta}-phase field at room temperature are generally applicable, it is also recognized that Pu and its alloys are never truly in thermodynamic equilibrium because of self-irradiation effects, primarily from the alpha decay of Pu isotopes. This article covers past and current research on several properties of Pu and Pu-(Al or Ga) alloys and their connections to the crystal structure and the microstructure. We review the consequences of radioactive decay, the recent advances in understanding the electronic structure, the current research on phase transformations and their relations to phase diagrams and phase stability, the nature of the isothermal martensitic {delta} {yields} {alpha}{prime} transformation, and the pressure-induced transformations in the {delta}-phase alloys. New data are also presented on the structures and phase transformations observed in these materials following the application of pressure, including the formation of transition phases.

  13. PU Vulpeculae - The outburst of a symbiotic nova

    NASA Astrophysics Data System (ADS)

    Vogel, M.; Nussbaumer, H.

    1992-06-01

    We report the full history of PU Vulpeculae from outburst to 1991 as seen in its ultraviolet emission. We show that PU Vul is a symbiotic nova, which went into outburst in 1977 after a nova-like thermonuclear event. The outbursting object went first into an F supergiant phase. The spectrum evolved between 1979 and 1989 from F-type into that of Ao. In 1990 PU Vul entered the nebular phase, showing a rich emission line spectrum in the UV and in the optical. Between 1979 and 1983-1985 the luminosity of the outbursting object increased by approximately a factor of 2 against 2600 solar luminosities in 1979. It subsequently decreased to reach in 1989 approximately the same value as in 1979. During 1980 the lightcurve went through a minimum and the spectral appearance changed. We interpret this as an eclipse of the outbursting star by the M giant companion. Based on IUE observations we discuss the early nebular phase of PU Vul, and we show that the UV is still dominated by the outbursting component, which in 1991 has reached a temperature of 40,000 K.

  14. Polymerization of Pu(IV) in aqueous nitric acid solutions

    SciTech Connect

    Toth, L.M.; Friedman, H.A.; Osborne, M.M.

    1980-10-01

    The polymerization of Pu(IV) in aqueous nitric acid solutions has been studied spectrophotometrically both to establish the influence of large UO{sub 2}(NO{sub 3}){sub 2} concentrations on the polymerization rates and, more generally, to review the influence of the major parameters on the polymer reaction. Typically, experiments have been performed at 50{sup 0}C and with 0.05 M Pu in nitric acid solutions that vary in acidity from 0.07 to 0.4 M. An induction period usually precedes the polymer growth stage during which time nucleation of primary hydrolysis products occurs. Uranyl nitrate retards the polymerization reaction by approximately 35% in spite of the counteracting influence of the nitrate ions associated with this solute. The rate of polymer formation, expressed as d(percent polymer)/dt, has been shown to depend on the total plutonium concentration in reactions where the Pu(IV) concentration remained constant; and it is therefore suggested that the polymer reaction rate is not first order with respect to the concentration of plutonium as was previously thought. It has been shown further that accurate acid determinations on stock reagents are essential in order to obtain reliable polymerization experiments. Satisfactory procedures for these analyses did not exist, so appropriate modifications to the iodate precipitation methods were developed. The most ideal plutonium reagent material has been shown to be crystalline Pu(IV) nitrate because it can be added directly to acid solutions without the occurrence of unintentional hydrolysis reactions.

  15. Determination of plutonium in spent nuclear fuel using high resolution X-ray

    SciTech Connect

    McIntosh, Kathryn G.; Reilly, Sean D.; Havrilla, George J.

    2015-05-30

    Characterization of Pu is an essential aspect of safeguards operations at nuclear fuel reprocessing facilities. A novel analysis technique called hiRX (high resolution X-ray) has been developed for the direct measurement of Pu in spent nuclear fuel dissolver solutions. hiRX is based on monochromatic wavelength dispersive X-ray fluorescence (MWDXRF), which provides enhanced sensitivity and specificity compared with conventional XRF techniques. A breadboard setup of the hiRX instrument was calibrated using spiked surrogate spent fuel (SSF) standards prepared as dried residues. Samples of actual spent fuel were utilized to evaluate the performance of the hiRX. The direct detection of just 39 ng of Pu is demonstrated. Initial quantitative results, with error of 4–27% and precision of 2% relative standard deviation (RSD), were obtained for spent fuel samples. The limit of detection for Pu (100 s) within an excitation spot of 200 μm diameter was 375 pg. This study demonstrates the potential for the hiRX technique to be utilized for the rapid, accurate, and precise determination of Pu. Moreover, the results highlight the analytical capability of hiRX for other applications requiring sensitive and selective nondestructive analyses.

  16. Determination of plutonium in spent nuclear fuel using high resolution X-ray

    DOE PAGES

    McIntosh, Kathryn G.; Reilly, Sean D.; Havrilla, George J.

    2015-05-30

    Characterization of Pu is an essential aspect of safeguards operations at nuclear fuel reprocessing facilities. A novel analysis technique called hiRX (high resolution X-ray) has been developed for the direct measurement of Pu in spent nuclear fuel dissolver solutions. hiRX is based on monochromatic wavelength dispersive X-ray fluorescence (MWDXRF), which provides enhanced sensitivity and specificity compared with conventional XRF techniques. A breadboard setup of the hiRX instrument was calibrated using spiked surrogate spent fuel (SSF) standards prepared as dried residues. Samples of actual spent fuel were utilized to evaluate the performance of the hiRX. The direct detection of just 39more » ng of Pu is demonstrated. Initial quantitative results, with error of 4–27% and precision of 2% relative standard deviation (RSD), were obtained for spent fuel samples. The limit of detection for Pu (100 s) within an excitation spot of 200 μm diameter was 375 pg. This study demonstrates the potential for the hiRX technique to be utilized for the rapid, accurate, and precise determination of Pu. Moreover, the results highlight the analytical capability of hiRX for other applications requiring sensitive and selective nondestructive analyses.« less

  17. Irradiated Nuclear Fuel Management: Resource Versus Waste

    SciTech Connect

    Nash, Kenneth L.; Lumetta, Gregg J.; Vienna, John D.

    2013-01-01

    Management of irradiated fuel is an important component of commercial nuclear power production. Although it is broadly agreed that the disposition of some fraction of the fuel in geological repositories will be necessary, there is a range of options that can be considered that affect exactly what fraction of material will be disposed in that manner. Furthermore, until geological repositories are available to accept commercial irradiated fuel, these materials must be safely stored. Temporary storage of irradiated fuel has traditionally been conducted in storage pools, and this is still true for freshly discharged fuel. Criticality control technologies have led to greater efficiencies in packing of irradiated fuel into storage pools. With continued delays in establishing permanent repositories, utilities have begun to move some of the irradiated fuel inventory into dry storage. Fuel cycle options being considered worldwide include the once-through fuel cycle, limited recycle in which U and Pu are recycled back to power reactors as mixed oxide fuel, and advance partitioning and transmutation schemes designed to reduce the long term hazards associated with geological disposal from millions of years to a few hundred years. Each of these options introduces specific challenges in terms of the waste forms required to safely immobilize the hazardous components of irradiated fuel.

  18. Neutronic Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 - Volume 4, Part 2--Saxton Plutonium Program Critical Experiments

    SciTech Connect

    Abdurrahman, NM

    2000-10-12

    Critical experiments with water-moderated, single-region PuO{sub 2}-UO{sub 2} or UO{sub 2}, and multiple-region PuO{sub 2}-UO{sub 2}- and UO{sub 2}-fueled cores were performed at the CRX reactor critical facility at the Westinghouse Reactor Evaluation Center (WREC) at Waltz Mill, Pennsylvania in 1965 [1]. These critical experiments were part of the Saxton Plutonium Program. The mixed oxide (MOX) fuel used in these critical experiments and then loaded in the Saxton reactor contained 6.6 wt% PuO{sub 2} in a mixture of PuO{sub 2} and natural UO{sub 2}. The Pu metal had the following isotopic mass percentages: 90.50% {sup 239}Pu; 8.57% {sup 239}Pu; 0.89% {sup 240}Pu; and 0.04% {sup 241}Pu. The purpose of these critical experiments was to verify the nuclear design of Saxton partial plutonium cores while obtaining parameters of fundamental significance such as buckling, control rod worth, soluble poison worth, flux, power peaking, relative pin power, and power sharing factors of MOX and UO{sub 2} lattices. For comparison purposes, the core was also loaded with uranium dioxide fuel rods only. This series is covered by experiments beginning with the designation SX.

  19. Development of a microanalytical energy dispersive X-ray fluorescence method for compositional characterization of (U, Pu)O2 samples

    NASA Astrophysics Data System (ADS)

    Dhara, Sangita; Sanjay Kumar, S.; Jayachandran, Kavitha; Kamat, J. V.; Kumar, Ashok; Radhakrishna, J.; Misra, N. L.

    2017-05-01

    Elemental compositional characterization of (U, Pu)O2 samples is an important aspect for quality control of nuclear fuels. A microanalytical Energy Dispersive X-ray Fluorescence (EDXRF) method has been developed for the determination of plutonium and uranium. The method involves sample dissolution, addition of internal standard(s) into the sample solution, taking about 50 μg of the sample on absorbent sheets, drying and sealing the specimens in such a manner that no loose radioactive particle comes out, during the analysis. Such approach does not require putting the instrument inside the glove box. These specimens were presented for the EDXRF measurements using Rh Kα as the excitation source. The amount of uranium and plutonium in samples were calculated using two different internal standards Ga and Y. The Pu weight percent (wt%) was calculated with respect to (U + Pu) in the samples. The results obtained were in good agreement with the biamperometric analysis results of the samples. The average precision observed was within 1% (1σ, n = 4) and the deviation of the EDXRF determined values from expected values was within 2%. The detection limits obtained for Pu and U were 4 μg.

  20. Modelling the thermal conductivity of (UxTh1-x)O2 and (UxPu1-x)O2

    DOE PAGES

    Cooper, M. W. D.; Middleburgh, S. C.; Grimes, R. W.

    2015-07-15

    The degradation of thermal conductivity due to the non-uniform cation lattice of (UxTh1-x)O2 and (UxPu1-x)O2 solid solutions has been investigated by molecular dynamics, using the non-equilibrium method, from 300 to 2000 K. Degradation of thermal conductivity is predicted in (UxTh1-x)O2 and (UxPu1-x)O2 as compositions deviate from the pure end members: UO2, PuO2 and ThO2. The reduction in thermal conductivity is most apparent at low temperatures where phonon-defect scattering dominates over phonon-phonon interactions. The effect is greater for (UxTh1-x)O2 than UxPu1-x)O2 due to the greater mismatch in cation size. Parameters for an analytical expressions have been developed that describe the predictedmore » thermal conductivities over the full temperature and compositional ranges. Finally, these expressions may be used in higher level fuel performance codes.« less

  1. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOEpatents

    Flint, O.

    1961-01-10

    Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

  2. ENHANCING ADVANCED CANDU PROLIFERATION RESISTANCE FUEL WITH MINOR ACTINIDES

    SciTech Connect

    Gray S. Chang

    2010-05-01

    The advanced nuclear system will significantly advance the science and technology of nuclear energy systems and to enhance the spent fuel proliferation resistance. Minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs can play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In this work, an Advanced CANDU Reactor (ACR) fuel unit lattice cell model with 43 UO2 fuel rods will be used to investigate the effectiveness of a Minor Actinide Reduction Approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. The main MARA objective is to increase the 238Pu / Pu isotope ratio by using the transuranic nuclides (237Np and 241Am) in the high burnup fuel and thereby increase the proliferation resistance even for a very low fuel burnup. As a result, MARA is a very effective approach to enhance the proliferation resistance for the on power refueling ACR system nuclear fuel. The MA transmutation characteristics at different MA loadings were compared and their impact on neutronics criticality assessed. The concept of MARA, significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance.

  3. Fuel Cycle System Analysis Handbook

    SciTech Connect

    Steven J. Piet; Brent W. Dixon; Dirk Gombert; Edward A. Hoffman; Gretchen E. Matthern; Kent A. Williams

    2009-06-01

    This Handbook aims to improve understanding and communication regarding nuclear fuel cycle options. It is intended to assist DOE, Campaign Managers, and other presenters prepare presentations and reports. When looking for information, check here. The Handbook generally includes few details of how calculations were performed, which can be found by consulting references provided to the reader. The Handbook emphasizes results in the form of graphics and diagrams, with only enough text to explain the graphic, to ensure that the messages associated with the graphic is clear, and to explain key assumptions and methods that cause the graphed results. Some of the material is new and is not found in previous reports, for example: (1) Section 3 has system-level mass flow diagrams for 0-tier (once-through), 1-tier (UOX to CR=0.50 fast reactor), and 2-tier (UOX to MOX-Pu to CR=0.50 fast reactor) scenarios - at both static and dynamic equilibrium. (2) To help inform fast reactor transuranic (TRU) conversion ratio and uranium supply behavior, section 5 provides the sustainable fast reactor growth rate as a function of TRU conversion ratio. (3) To help clarify the difference in recycling Pu, NpPu, NpPuAm, and all-TRU, section 5 provides mass fraction, gamma, and neutron emission for those four cases for MOX, heterogeneous LWR IMF (assemblies mixing IMF and UOX pins), and a CR=0.50 fast reactor. There are data for the first 10 LWR recycle passes and equilibrium. (4) Section 6 provides information on the cycle length, planned and unplanned outages, and TRU enrichment as a function of fast reactor TRU conversion ratio, as well as the dilution of TRU feedstock by uranium in making fast reactor fuel. (The recovered uranium is considered to be more pure than recovered TRU.) The latter parameter impacts the required TRU impurity limits specified by the Fuels Campaign. (5) Section 7 provides flows for an 800-tonne UOX separation plant. (6) To complement 'tornado' economic uncertainty

  4. Simulation of radiation driven fission gas diffusion in UO2, ThO2 and PuO2

    DOE PAGES

    Cooper, Michael William D.; Stanek, Christopher Richard; Turnbull, James Anthony; ...

    2016-12-01

    Below 1000 K it is thought that fission gas diffusion in nuclear fuel during irradiation occurs through atomic mixing due to radiation damage. Here we present a molecular dynamics (MD) study of Xe, Kr, Th, U, Pu and O diffusion due to irradiation. It is concluded that the ballistic phase does not sufficiently account for the experimentally observed diffusion. Thermal spike simulations are used to confirm that electronic stopping remedies the discrepancy with experiment and the predicted diffusivities lie within the scatter of the experimental data. Here, our results predict that the diffusion coefficients are ordered such that D*0 >more » D*Kr > D*Xe > D*U. For all species >98.5% of diffusivity is accounted for by electronic stopping. Fission gas diffusivity was not predicted to vary significantly between ThO2, UO2 and PuO2, indicating that this process would not change greatly for mixed oxide fuels.« less

  5. Simulation of radiation driven fission gas diffusion in UO2, ThO2 and PuO2

    DOE PAGES

    Cooper, Michael William D.; Stanek, Christopher Richard; Turnbull, James Anthony; ...

    2016-12-01

    Below 1000 K it is thought that fission gas diffusion in nuclear fuel during irradiation occurs through atomic mixing due to radiation damage. Here we present a molecular dynamics (MD) study of Xe, Kr, Th, U, Pu and O diffusion due to irradiation. It is concluded that the ballistic phase does not sufficiently account for the experimentally observed diffusion. Thermal spike simulations are used to confirm that electronic stopping remedies the discrepancy with experiment and the predicted diffusivities lie within the scatter of the experimental data. Here, our results predict that the diffusion coefficients are ordered such that D*0 >more » D*Kr > D*Xe > D*U. For all species >98.5% of diffusivity is accounted for by electronic stopping. Fission gas diffusivity was not predicted to vary significantly between ThO2, UO2 and PuO2, indicating that this process would not change greatly for mixed oxide fuels.« less

  6. Effects of Aging on PuO2∙xH2O Particle Size in Alkaline Solution

    SciTech Connect

    Delegard, Calvin H.

    2013-05-01

    Between 1944 and 1989, 54.5 metric tons of the United States’ weapons-grade plutonium and an additional 12.9 metric tons of fuel-grade plutonium were produced and separated from irradiated fuel at the Hanford Site. Acidic high-activity wastes containing around 600 kg of plutonium were made alkaline and discharged to underground storage tanks from separations, isolation, and recycle processes to yield average plutonium concentration of about 0.003 grams per liter (or ~0.0002 wt%) in the ~200 million liter tank waste volume. The plutonium is largely associated with low-solubility metal hydroxide/oxide sludges where its low concentration and intimate mixture with neutron-absorbing elements (e.g., iron) are credited in nuclear criticality safety. However, concerns have been expressed that plutonium, in the form of plutonium hydrous oxide, PuO2∙xH2O, could undergo sufficient crystal growth through dissolution and reprecipitation in the alkaline tank waste to potentially become separable from neutron absorbing constituents by settling or sedimentation. Thermodynamic considerations and laboratory studies of systems chemically analogous to tank waste show that the plutonium formed in the alkaline tank waste by precipitation through neutralization from acid solution probably entered as 2–4-nm PuO2∙xH2O crystallite particles that, because of their low solubility and opposition from radiolytic processes, grow from that point at exceedingly slow rates, thus posing no risk of physical segregation.

  7. A theoretical study of the ground state and lowest excited states of PuO0/+/+2 and PuO20/+/+2

    SciTech Connect

    Gibson, John K.; La Macchia, Giovanni; Infante, Ivan; Gagliardi, Laura; Raab, Juraj

    2008-12-08

    The ground and excited states of neutral and cationic PuO and PuO2 have been studied with multiconfigurational quantum chemical methods followed by second order perturbation theory, the CASSCF/CASPT2 method. Scalar relativistic effects and spin-orbit coupling have been included in the treatment. As literature values for the ionization energy of PuO2 are in the wide range of ~;;6.6 eV to ~;;10.1 eV, a central goal of the computations was to resolve these discrepancies; the theoretical results indicate that the ionization energy is near the lower end of this range. The calculated ionization energies for PuO, PuO+ and PuO2+ are in good agreement with the experimental values.

  8. Determination of (235)U, (239)Pu, (240)Pu, and (241)Am in a nuclear bomb particle using a position-sensitive α-γ coincidence technique.

    PubMed

    Peräjärvi, Kari A; Ihantola, Sakari; Pöllänen, Roy C; Toivonen, Harri I; Turunen, Jani A

    2011-02-15

    A nuclear bomb particle containing 1.6 ng of Pu was investigated nondestructively with a position-sensitive α detector and a broad-energy HPGe γ-ray detector. An event-mode data acquisition system was used to record the data. α-γ coincidence counting was shown to be well suited to nondestructive isotope ratio determination. Because of the very small background, the 51.6 keV γ rays of (239)Pu and the 45.2 keV γ rays of (240)Pu were identified, which enabled isotopic ratio calculations. In the present work, the (239)Pu/((239)Pu+(240)Pu) atom ratio was determined to be 0.950 ± 0.010. The uncertainties were much smaller than in the previous more conventional nondestructive studies on this particle. Obtained results are also in good agreement with the data from the destructive mass spectrometric studies obtained previously by other investigators.

  9. Formation of plutonium(IV) colloid by the alpha-reduction of aqueous solutions of Pu(V) and Pu(VI)

    SciTech Connect

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

    1985-12-31

    We describe concentration changes caused by chemical and alpha-induced radiolytic reactions in various oxidation state pure solutions of Pu(VI), Pu(V), or Pu(IV) colloid or mixtures of these oxidation states at pH values > 1 for a period of nearly two years. The rates of approach to steady-states and the resulting experimental concentration quotient values were determined in order to find the conditions under which equilibrium in 2PuO{sub 2}{sup +} + PuO{sub 2}{sup 2+} + PuO/sub 2(coll)/ reaction might be attained and to learn about the underlying reactions. Computer calculations were used to compare the data with the results required from proposed reaction schemes.

  10. Spent nuclear fuel recycling with plasma reduction and etching

    DOEpatents

    Kim, Yong Ho

    2012-06-05

    A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF6 and PuF4. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF6 (gas) that is formed to decompose back to PuF4 (solid), and b) to maintain stability of the UF6. Uranium (in the form of gaseous UF6) is easily extracted and separated from the plutonium (in the form of solid PuF4). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF6 (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF4 or NF3 as the fluorine sources instead of F2 or HF.

  11. Application of fully ceramic microencapsulated fuels in light water reactors

    SciTech Connect

    Gentry, C.; George, N.; Maldonado, I.; Godfrey, A.; Terrani, K.; Gehin, J.

    2012-07-01

    This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

  12. Variations in Pu isotopic composition in soils from the Spitsbergen (Norway): Three potential pollution sources of the Arctic region.

    PubMed

    Łokas, E; Anczkiewicz, R; Kierepko, R; Mietelski, J W

    2017-07-01

    Although the polar regions have not been industrialised, numerous contaminants originating from human activity are detectable in the Arctic environment. This study reports evidence of (240)Pu/(239)Pu atomic ratios in the tundra and initial soils from different parts of west and central Spitsbergen and recognizes possible environmental inputs of non-global fallout Pu. The average atomic ratio of (240)Pu/(239)Pu equal to 0.179 (ranging between 0.129 and 0.201) in tundra soils are comparable to the characteristic ratio for global fallout (0.180). However, the (240)Pu/(239)Pu atomic ratios in the initial soils from proglacial zone of glaciers change within wide range between 0.1281 and 0.234 with the mean value of 0.169. By combining alpha and mass spectrometry, the three-sources model was used to identify the Pu sources in initial soils. Our study indicated that the main source of Pu is nuclear tests and that a second source with lower Pu ratio may come from weapons grade Pu (unexploded weapons grade Pu ie. material from bomb which didn't undergo nuclear explosions for example for security tests). Additionally, we found samples with high (238)Pu/(239+240)Pu activity ratios and with typical global fallout (240)Pu/(239)Pu atomic ratios, which are associated with separate sources of pure (238)Pu from the SNAP-9A satellite burn up in the atmosphere. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. Release of Pu isotopes from the Fukushima Daiichi Nuclear Power Plant accident to the marine environment was negligible.

    PubMed

    Bu, Wenting; Fukuda, Miho; Zheng, Jian; Aono, Tatsuo; Ishimaru, Takashi; Kanda, Jota; Yang, Guosheng; Tagami, Keiko; Uchida, Shigeo; Guo, Qiuju; Yamada, Masatoshi

    2014-08-19

    Atmospheric deposition of Pu isotopes from the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident has been observed in the terrestrial environment around the FDNPP site; however, their deposition in the marine environment has not been studied. The possible contamination of Pu in the marine environment has attracted great scientific and public concern. To fully understand this possible contamination of Pu isotopes from the FDNPP accident to the marine environment, we collected marine sediment core samples within the 30 km zone around the FDNPP site in the western North Pacific about two years after the accident. Pu isotopes ((239)Pu, (240)Pu, and (241)Pu) and radiocesium isotopes ((134)Cs and (137)Cs) in the samples were determined. The high activities of radiocesium and the (134)Cs/(137)Cs activity ratios with values around 1 (decay corrected to 15 March 2011) suggested that these samples were contaminated by the FDNPP accident-released radionuclides. However, the activities of (239+240)Pu and (241)Pu were low compared with the background level before the FDNPP accident. The Pu atom ratios ((240)Pu/(239)Pu and (241)Pu/(239)Pu) suggested that global fallout and the pacific proving ground (PPG) close-in fallout are the main sources for Pu contamination in the marine sediments. As Pu isotopes are particle-reactive and they can be easily incorporated with the marine sediments, we concluded that the release of Pu isotopes from the FDNPP accident to the marine environment was negligible.

  14. 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.

  15. Synthesis, structural and thermal studies of Pu(VO3)4: A new vanadate in PuO2-V2O5 system

    NASA Astrophysics Data System (ADS)

    Dahale, N. D.; Achary, S. N.; Sali, S. K.; Keskar, Meera; Phatak, R.; Kannan, S.

    2012-08-01

    A new phase Pu(VO3)4 in PuO2-V2O5 system was prepared by heating an appropriate amount of pre-dried PuO2 and V2O5 at 640 °C. The structural and thermal properties of Pu(VO3)4 were investigated by powder X-ray diffraction and thermogravimetry (TG). Structural analyses revealed that Pu(VO3)4 crystallizes in tetragonal (space group: I41/a) lattice with unit cell parameters, a = 8.4528(2) Å and c = 28.6644(9) Å and is isostructural to metavanadates of thorium and neptunium. Structure shows crystallographically two distinct vanadium atoms V(1) and V(2), one plutonium atom (Pu1) and seven oxygen atoms in the unit cell of Pu(VO3)4. Vanadium atoms form VO4 tetrahedra while plutonium forms PuO8 polyhedra with oxygen atoms. The VO4 units are linked together by sharing apex oxygen atoms and form infinite spiral chains of (VO3)n. Along c-axis, the plutonium atoms are sandwiched between the two layers of (VO3)n chains. TG of Pu(VO3)4 showed thermal stability of the compound up to 700 °C.

  16. Structural, magnetic, electronic and optical properties of PuC and PuC0.75: A hybrid density functional study

    NASA Astrophysics Data System (ADS)

    Yang, Rong; Tang, Bin; Gao, Tao; Ao, BingYun

    2016-05-01

    We perform first principles calculations to investigate the structural, magnetic, electronic and optical properties of PuC and PuC0.75. Furthermore, we examine the influence of carbon non-stoichiometry on plutonium monocarbide. For the treatment of strongly correlated electrons, the hybrid density functionals like PBE0, Fock-0.25 are used and we compare the results with the generalized gradient approximation (GGA), local density approximation (LDA), LDA + U and experimental ones. The optimized lattice constant a0 = 4.961 Å for PuC in the Fock-0.25 scheme is the most close to the experimental data. The ground states of PuC and PuC0.75 are found to be anti-ferromagnetic. Our results indicate that additional removal of a C atom make lattice contract and new DOS peak appear in the near-Fermi region. We also compute and compare the optical properties of PuC and PuC0.75. The difference in optical properties between PuC and PuC0.75 should also be the influence of carbon vacancies.

  17. Partitioning of /sup 238/Pu, /sup 239/Pu and /sup 241/Am in skeleton and liver of U. S. Transuranium Registry autopsy cases

    SciTech Connect

    Kathren, R.L.; McInroy, J.F.; Reichert, M.M.; Swint, M.J.

    1988-02-01

    The content of /sup 238/Pu, /sup 239/Pu and /sup 241/Am in the liver and skeleton was estimated from radiochemical analysis of human liver and bone samples obtained at autopsy from former actinide workers whose occupational histories were suggestive of chronic inhalation exposures, with minor skin contamination and wounds documented in a few individuals. For times estimated to be several years to a few decades post intake, 75.8 +/- 15.3% of the total /sup 241/Am in the skeleton and liver was found in the skeleton (25 cases) as compared with 63.4 +/- 24.1% for /sup 238/Pu (36 cases) and 53.2 +/- 18.2% for /sup 239/Pu (43 cases). These differences are significant at the 95% confidence level. Of these cases, 34 included data on both /sup 238/Pu and /sup 239/Pu and were divided into high and low activity subgroups. The difference in the fractionation of the two Pu isotopes was apparent only in the low activity subgroup, suggesting that the difference observed between the Pu isotopes may be an artifact of the data. The different partitioning of these three nuclides suggests that the ALIs for /sup 238/Pu and /sup 241/Am may be high by about 25-50% if only the dose to bone is considered and may be high by 12-13%, based on the weighted committed dose equivalent in target organs or tissues.

  18. AFS-2 FLOWSHEET MODIFICATIONS TO ADDRESS THE INGROWTH OF PU(VI) DURING METAL DISSOLUTION

    SciTech Connect

    Crapse, K.; Rudisill, T.; O'Rourke, P.; Kyser, E.

    2014-07-02

    In support of the Alternate Feed Stock Two (AFS-2) PuO{sub 2} production campaign, Savannah River National Laboratory (SRNL) conducted a series of experiments concluding that dissolving Pu metal at 95°C using a 6–10 M HNO{sub 3} solution containing 0.05–0.2 M KF and 0–2 g/L B could reduce the oxidation of Pu(IV) to Pu(VI) as compared to dissolving Pu metal under the same conditions but at or near the boiling temperature. This flowsheet was demonstrated by conducting Pu metal dissolutions at 95°C to ensure that PuO{sub 2} solids were not formed during the dissolution. These dissolution parameters can be used for dissolving both Aqueous Polishing (AP) and MOX Process (MP) specification materials. Preceding the studies reported herein, two batches of Pu metal were dissolved in the H-Canyon 6.1D dissolver to prepare feed solution for the AFS-2 PuO{sub 2} production campaign. While in storage, UV-visible spectra obtained from an at-line spectrophotometer indicated the presence of Pu(VI). Analysis of the solutions also showed the presence of Fe, Ni, and Cr. Oxidation of Pu(IV) produced during metal dissolution to Pu(VI) is a concern for anion exchange purification. Anion exchange requires Pu in the +4 oxidation state for formation of the anionic plutonium(IV) hexanitrato complex which absorbs onto the resin. The presence of Pu(VI) in the anion feed solution would require a valence adjustment step to prevent losses. In addition, the presence of Cr(VI) would result in absorption of chromate ion onto the resin and could limit the purification of Pu from Cr which may challenge the purity specification of the final PuO{sub 2} product. Initial experiments were performed to quantify the rate of oxidation of Pu(IV) to Pu(VI) (presumed to be facilitated by Cr(VI)) as functions of the HNO{sub 3} concentration and temperature in simulated dissolution solutions containing Cr, Fe, and Ni. In these simulated Pu dissolutions studies, lowering the temperature from near boiling

  19. AMS of natural 236U and 239Pu produced in uranium ores

    NASA Astrophysics Data System (ADS)

    Wilcken, K. M.; Barrows, T. T.; Fifield, L. K.; Tims, S. G.; Steier, P.

    2007-06-01

    The rare isotopes 236U and 239Pu are produced naturally by neutron capture in uranium ores. Here we measure 236U and 239Pu by accelerator mass spectrometry (AMS) in the same ore samples for the first time. To ensure efficient extraction of both elements and isotopic equilibrium between the 239Pu in the ore and a 242Pu spike, we developed a new sample preparation protocol. AMS has clear advantages over previous methods because it achieves better discrimination against molecular interferences with higher sensitivity and shorter counting times. Measurements of 236U and 239Pu hold considerable promise as proxy indicators of neutron flux and uranium concentration.

  20. Vertical distribution of (241)Pu in the southern Baltic Sea sediments.

    PubMed

    Strumińska-Parulska, Dagmara I

    2014-12-15

    The vertical distribution of plutonium (241)Pu in marine sediments can assist in determining the deposition history and sedimentation process of analyzed regions. In addition, (241)Pu/(239+240)Pu activity ratio could be used as a sensitive fingerprint for radioactive source identification. The present preliminary studies on vertical distribution of (241)Pu in sediments from four regions of the southern Baltic Sea are presented. The distribution of (241)Pu was not uniform and depended on sediment geomorphology and depth as well as location. The highest concentrations of plutonium were found in the surface layers of all analyzed sediments and originated from the Chernobyl accident.

  1. Production of {sup 238}PuO{sub 2} heat sources for the Cassini mission

    SciTech Connect

    George, T.G.; Foltyn, E.M.

    1998-01-01

    NASA{close_quote}s Cassini mission to Saturn, scheduled to launch in October, 1997, is perhaps the most ambitious interplanetary explorer ever constructed. Electric power for the spacecraft{close_quote}s science instruments and on-board computers will be provided by three radioisotope thermoelectric generators (RTGs) powered by 216 {sup 238}PuO{sub 2}-fueled General-Purpose Heat Source (GPHS) capsules. In addition, critical equipment and instruments on the spacecraft and Huygens probe will be warmed by 128 Light-Weight Radioisotope Heater Units (LWRHUs). Fabrication and assembly of the GPHS capsules and LWRHU heat sources was performed at Los Alamos National Laboratory (LANL) between January 1994 and September 1996. During this production campaign, LANL pressed and sintered 315 GPHS fuel pellets and 181 LWRHU pellets. By October 1996, NMT-9 had delivered a total of 235 GPHS capsules to EG&G Mound Applied Technologies (EG&G MAT) in Miamisburg, Ohio. EG&G MAT conditioned the capsules for use, loaded the capsules into the Cassini RTGs, tested the RTGs, and coordinated transportation to Kennedy Space Center (KSC). LANL also fabricated and assembled a total of 180 LWRHUs. The LWRHUs required for the Cassini spacecraft were shipped to KSC in mid-1997. {copyright} {ital 1998 American Institute of Physics.}

  2. Anaerobic Biotransformation and Mobility of Pu and of Pu-EDTA

    SciTech Connect

    Xun, Luying

    2009-11-20

    The enhanced mobility of radionuclides by co-disposed chelating agent, ethylenediaminetetraacetate (EDTA), is likely to occur only under anaerobic conditions. Our extensive effort to enrich and isolate anaerobic EDTA-degrading bacteria has failed. Others has tried and also failed. To explain the lack of anaerobic biodegradation of EDTA, we proposed that EDTA has to be transported into the cells for metabolism. A failure of uptake may contribute to the lack of EDTA degradation under anaerobic conditions. We demonstrated that an aerobic EDTA-degrading bacterium strain BNC1 uses an ABC-type transporter system to uptake EDTA. The system has a periplasmic binding protein that bind EDTA and then interacts with membrane proteins to transport EDTA into the cell at the expense of ATP. The bind protein EppA binds only free EDTA with a Kd of 25 nM. The low Kd value indicates high affinity. However, the Kd value of Ni-EDTA is 2.4 x 10^(-10) nM, indicating much stronger stability. Since Ni and other trace metals are essential for anaerobic respiration, we conclude that the added EDTA sequestrates all trace metals and making anaerobic respiration impossible. Thus, the data explain the lack of anaerobic enrichment cultures for EDTA degradation. Although we did not obtain an EDTA degrading culture under anaerobic conditions, our finding may promote the use of certain metals that forms more stable metal-EDTA complexes than Pu(III)-EDTA to prevent the enhanced mobility. Further, our data explain why EDTA is the most dominant organic pollutant in surface waters, due to the lack of degradation of certain metal-EDTA complexes.

  3. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997

    SciTech Connect

    Akkurt, H

    2001-01-11

    In 1967, a series of critical experiments were conducted at the Westinghouse Reactor Evaluation Center (WREC) using mixed-oxide (MOX) PuO{sub 2}-UO{sub 2} and/or UO{sub 2} fuels in various lattices and configurations . These experiments were performed under the joint sponsorship of the Empire State Atomic Development Associates (ESADA) plutonium program and Westinghouse . The purpose of these experiments was to develop experimental data to validate analytical methods used in the design of a plutonium-bearing replacement fuel for water reactors. Three different fuels were used during the experimental program: two MOX fuels and a low-enriched UO{sub 2} fuel. The MOX fuels were distinguished by their {sup 240}Pu content: 8 wt% {sup 240}Pu and 24 wt% {sup 240}Pu. Both MOX fuels contained 2.0 wt % PuO{sub 2} in natural UO{sub 2} . The UO{sub 2} fuel with 2.72 wt % enrichment was used for comparison with the plutonium data and for use in multiregion experiments.

  4. Redox state of plutonium in irradiated mixed oxide fuels

    NASA Astrophysics Data System (ADS)

    Degueldre, C.; Pin, S.; Poonoosamy, J.; Kulik, D. A.

    2014-03-01

    Nowadays, MOX fuels are used in about 20 nuclear power plants around the world. After irradiation, plutonium co-exists with uranium oxide. Due to the redox sensitive nature of UO2 other plutonium oxides than PuO2 potentially present in the fuel may interact with the matrix. The aim of this study is to determine which plutonium species are present in heterogeneous and homogeneous MOX. The results provided by X-ray Absorption Near Edge Spectroscopy (XANES) for non-irradiated as well as irradiated (center and periphery) homogeneous MOX fuel were published earlier and are completed by Extended X-ray Fine Structure (EXAFS) analysis in this work. The EXAFS signals have been extracted using the ATHENA code and the analyses were carried using EXCURE98 as performed earlier for an analogous element. EXAFS shows that plutonium redox state remains tetravalent in the solid solution and that the minor fraction of trivalent Pu must be below 10%. Independently, the study of homogeneous MOX was also approached by thermodynamics of solid solution of (U,Pu)O2. Such solid solutions were modeled using the Gibbs Energy Minimisation (GEM)-Selektor code (developed at LES, NES, PSI) supported by the literature data on such solid solutions. A comparative study was performed showing which plutonium oxides in their respective mole fractions are more likely to occur in (U,Pu)O2. In the modeling, these oxides were set as ideal and non-ideal solid solutions, as well as separate pure phases. Pu exists mainly as PuO2 in the case of separate phases, but can exist under its reduced forms, PuO1.61 and PuO1.5 in minor fraction i.e. ~15% in ideal solid solution (unlikely) and ~10% in non-ideal solid solution (likely) and at temperature around 1300 K. This combined thermodynamic and EXAFS studies confirm independently the results obtained so far by Pu XANES for the same MOX samples.

  5. Surface-mediated formation of Pu(IV) nanoparticles at the muscovite-electrolyte interface.

    PubMed

    Schmidt, Moritz; Lee, Sang Soo; Wilson, Richard E; Knope, Karah E; Bellucci, Francesco; Eng, Peter J; Stubbs, Joanne E; Soderholm, L; Fenter, P

    2013-12-17

    The formation of Pu(IV)-oxo-nanoparticles from Pu(III) solutions by a surface-enhanced redox/polymerization reaction at the muscovite (001) basal plane is reported, with a continuous increase in plutonium coverage observed in situ over several hours. The sorbed Pu extends >70 Å from the surface with a maximum concentration at 10.5 Å and a total coverage of >9 Pu atoms per unit cell area of muscovite (0.77 μg Pu/cm(2)) (determined independently by in situ resonant anomalous X-ray reflectivity and by ex-situ alpha-spectrometry). The presence of discrete nanoparticles is confirmed by high resolution atomic force microscopy. We propose that the formation of these Pu(IV) nanoparticles from an otherwise stable Pu(III) solution can be explained by the combination of a highly concentrated interfacial Pu-ion species, the Pu(III)-Pu(IV) redox equilibrium, and the strong proclivity of tetravalent Pu to hydrolyze and form polymeric species. These results are the first direct observation of such behavior of plutonium on a naturally occurring mineral, providing insights into understanding the environmental transport of plutonium and other contaminants capable of similar redox/polymerization reactions.

  6. PU.1 is a potent tumor suppressor in classical Hodgkin lymphoma cells.

    PubMed

    Yuki, Hiromichi; Ueno, Shikiko; Tatetsu, Hiro; Niiro, Hiroaki; Iino, Tadafumi; Endo, Shinya; Kawano, Yawara; Komohara, Yoshihiro; Takeya, Motohiro; Hata, Hiroyuki; Okada, Seiji; Watanabe, Toshiki; Akashi, Koichi; Mitsuya, Hiroaki; Okuno, Yutaka

    2013-02-07

    PU.1 has previously been shown to be down-regulated in classical Hodgkin lymphoma (cHL) cells via promoter methylation. We performed bisulfite sequencing and proved that the promoter region and the -17 kb upstream regulatory element of the PU.1 gene were highly methylated. To evaluate whether down-regulation of PU.1 is essential for the growth of cHL cells, we conditionally expressed PU.1 in 2 cHL cell lines, L428 and KM-H2. Overexpression of PU.1 induced complete growth arrest and apoptosis in both cell lines. Furthermore, in a Hodgkin lymphoma tumor xenograft model using L428 and KM-H2 cell lines, overexpression of PU.1 led to tumor regression or stable disease. Lentiviral transduction of PU.1 into primary cHL cells also induced apoptosis. DNA microarray analysis revealed that among genes related to cell cycle and apoptosis, p21 (CDKN1A) was highly up-regulated in L428 cells after PU.1 induction. Stable knockdown of p21 rescued PU.1-induced growth arrest in L428 cells, suggesting that the growth arrest and apoptosis observed are at least partially dependent on p21 up-regulation. These data strongly suggest that PU.1 is a potent tumor suppressor in cHL and that induction of PU.1 with demethylation agents and/or histone deacetylase inhibitors is worth exploring as a possible therapeutic option for patients with cHL.

  7. Plutonium isotopes in settling particles: transport and scavenging of Pu in the western Northwest Pacific.

    PubMed

    Zheng, Jian; Yamada, Masatoshi

    2006-07-01

    We examined the vertical distributions of 239+240Pu activity and 240Pu/239Pu atom ratio in settling particles and quantified the particulate 239+240Pu fluxes in the water column in the western Northwest Pacific. Settling particle samples were collected using sediment traps. Plutonium isotopes were analyzed using a sector field high-resolution ICP-MS. To the best of our knowledge, this is the first time that both Pu activity and Pu isotope ratio data have been obtained for settling particles in the Pacific Ocean. The high (>0.18) 240Pu/239Pu atom ratios in settling particles indicate that plutonium from the Pacific Proving Grounds (PPG) source in the central Pacific is transported toward the western Northwest Pacific. Evidence indicates that Pu scavenging onto the settling particles is strongly dependent upon the bulk mass flux. The results suggest that advective lateral transport of dissolved Pu from the open ocean to the ocean margin and removal of Pu into the margin sediments by particle scavenging is a common phenomenon in the Pacific Ocean. Plutonium can be considered as a useful tracer to study the transport and fate of other contaminants that readily adsorb to particles in marine environments.

  8. PLANTS AS BIO-MONITORS FOR 137CS, 238PU, 239, 240PU AND 40K AT THE SAVANNAH RIVER SITE

    SciTech Connect

    Caldwell, E.; Duff, M.; Ferguson, C.

    2010-12-16

    The nuclear fuel cycle generates a considerable amount of radioactive waste, which often includes nuclear fission products, such as strontium-90 ({sup 90}Sr) and cesium-137 ({sup 137}Cs), and actinides such as uranium (U) and plutonium (Pu). When released into the environment, large quantities of these radionuclides can present considerable problems to man and biota due to their radioactive nature and, in some cases as with the actinides, their chemical toxicity. Radionuclides are expected to decay at a known rate. Yet, research has shown the rate of elimination from an ecosystem to differ from the decay rate due to physical, chemical and biological processes that remove the contaminant or reduce its biological availability. Knowledge regarding the rate by which a contaminant is eliminated from an ecosystem (ecological half-life) is important for evaluating the duration and potential severity of risk. To better understand a contaminants impact on an environment, consideration should be given to plants. As primary producers, they represent an important mode of contamination transfer from sediments and soils into the food chain. Contaminants that are chemically and/or physically sequestered in a media are less likely to be bio-available to plants and therefore an ecosystem.

  9. Positive feedback between PU.1 and the cell cycle controls myeloid differentiation.

    PubMed

    Kueh, Hao Yuan; Champhekar, Ameya; Champhekhar, Ameya; Nutt, Stephen L; Elowitz, Michael B; Rothenberg, Ellen V

    2013-08-09

    Regulatory gene circuits with positive-feedback loops control stem cell differentiation, but several mechanisms can contribute to positive feedback. Here, we dissect feedback mechanisms through which the transcription factor PU.1 controls lymphoid and myeloid differentiation. Quantitative live-cell imaging revealed that developing B cells decrease PU.1 levels by reducing PU.1 transcription, whereas developing macrophages increase PU.1 levels by lengthening their cell cycles, which causes stable PU.1 accumulation. Exogenous PU.1 expression in progenitors increases endogenous PU.1 levels by inducing cell cycle lengthening, implying positive feedback between a regulatory factor and the cell cycle. Mathematical modeling showed that this cell cycle-coupled feedback architecture effectively stabilizes a slow-dividing differentiated state. These results show that cell cycle duration functions as an integral part of a positive autoregulatory circuit to control cell fate.

  10. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    SciTech Connect

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-07-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at {approx}2.4, {approx}7 and {approx}11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of {approx}7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10{sup 15} n/cm{sup 2}/s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between {approx}410 deg. C and {approx}645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

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

    NASA Astrophysics Data System (ADS)

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

    2016-04-01

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

  12. Characterization of Pu concentration and its isotopic composition in soils of Gansu in northwestern China.

    PubMed

    Zheng, Jian; Yamada, Masatoshi; Wu, Fengchang; Liao, Haiqing

    2009-01-01

    The total 239+240Pu activities and 240Pu/239Pu atom ratios in surface soil samples (0-5cm) in the Kumtag Desert in western Gansu Province, and in a soil core sample in Lanzhou were investigated using a sector-field ICP-MS. In the surface soil samples, 239+240Pu activities in fine particles (<150microm) were 1.3-2.1 times of those in coarse particles (150microm-1mm) which ranged from 0.005 to 0.157mBq/g. Atom ratios of 240Pu/239Pu in the surface soils ranged from 0.168 to 0.192 with a mean of 0.182+/-0.008. The mean ratio was similar to the typical global fallout value although the Kumtag Desert was believed to have received close-in fallout derived from Chinese nuclear weapons tests mainly conducted in the 1970s. Furthermore, the mean 240Pu/239Pu atom ratio observed in the soil core sample in Lanzhou was similar to the typical global fallout value. In the soil core sample, 239+240Pu activities in the various layers ranged from 0.012 to 0.23mBq/g, and the inventory of 239+240Pu (32.4Bq/m2, 0-23cm) was slightly lower than that expected from global fallout (42Bq/m2) at the same latitude. Rapid downward migration of Pu isotopes was observed in Lanzhou soil core sample layers. The contribution of the 10-cm deep top layers of surface soils to total inventory was only 17%, while the contribution of deeper layers (10-23cm) was as high as 83%. The 239+240Pu activity levels and 240Pu/239Pu atom ratios in soils in Gansu Province, China are similar to those in atmospheric deposition samples collected in the spring in recent years in Japan.

  13. Thermal conductivity of heterogeneous LWR MOX fuels

    NASA Astrophysics Data System (ADS)

    Staicu, D.; Barker, M.

    2013-11-01

    It is generally observed that the thermal conductivity of LWR MOX fuel is lower than that of pure UO2. For MOX, the degradation is usually only interpreted as an effect of the substitution of U atoms by Pu. This hypothesis is however in contradiction with the observations of Duriez and Philiponneau showing that the thermal conductivity of MOX is independent of the Pu content in the ranges 3-15 and 15-30 wt.% PuO2 respectively. Attributing this degradation to Pu only implies that stoichiometric heterogeneous MOX can be obtained, while we show that any heterogeneity in the plutonium distribution in the sample introduces a variation in the local stoichiometry which in turn has a strong impact on the thermal conductivity. A model quantifying this effect is obtained and a new set of experimental results for homogeneous and heterogeneous MOX fuels is presented and used to validate the proposed model. In irradiated fuels, this effect is predicted to disappear early during irradiation. The 3, 6 and 10 wt.% Pu samples have a similar thermal conductivity. Comparison of the results for this homogeneous microstructure with MIMAS (heterogeneous) fuel of the same composition showed no difference for the Pu contents of 3, 5.9, 6, 7.87 and 10 wt.%. A small increase of the thermal conductivity was obtained for 15 wt.% Pu. This increase is of about 6% when compared to the average of the values obtained for 3, 6 and 10 wt.% Pu. For comparison purposes, Duriez also measured the thermal conductivity of FBR MOX with 21.4 wt.% Pu with O/M = 1.982 and a density close to 95% TD and found a value in good agreement with the estimation obtained using the formula of Philipponneau [8] for FBR MOX, and significantly lower than his results corresponding to the range 3-15 wt.% Pu. This difference in thermal conductivity is of about 20%, i.e. higher than the measurement uncertainties.Thus, a significant difference was observed between FBR and PWR MOX fuels, but was not explained. This difference

  14. Plutonium activities and 240Pu/ 239Pu atom ratios in sediment cores from the east China sea and Okinawa Trough: Sources and inventories

    NASA Astrophysics Data System (ADS)

    Wang, Zhong-liang; Yamada, Masatoshi

    2005-05-01

    Plutonium concentrations and 240Pu/ 239Pu atom ratios in the East China Sea and Okinawa Trough sediment cores were determined by isotope dilution inductively coupled plasma mass spectrometry after separation using ion-exchange chromatography. The results showed that 240Pu/ 239Pu atom ratios in the East China Sea and Okinawa Trough sediments, ranging from 0.21 to 0.33, were much higher than the reported value of global fallout (0.18). The highest 240Pu/ 239Pu ratios (0.32-0.33) were observed in the deepest Okinawa Trough sediment samples. These ratios suggested the US nuclear weapons tests in the early 1950s at the Pacific Proving Grounds in the Marshall Islands were a major source of plutonium in the East China Sea and Okinawa Trough sediments, in addition to the global fallout source. It was proposed that close-in fallout plutonium was delivered from the Pacific Proving Grounds test sites via early direct tropospheric fallout and transportation by the North Pacific Equatorial Circulation system and Kuroshio Current into the Okinawa Trough and East China Sea. The total 239 + 240 Pu inventories in the cores were about 150-200% of that expected from direct global fallout; about 46-67% of the total inventories were delivered from the Pacific Proving Grounds. Much higher 239 + 240 Pu inventories were observed in the East China Sea sediments than in sediments of the Okinawa Trough, because in the open oceans, part of the 239 + 240 Pu was still retained in the water column, and continued Pu scavenging was higher over the margin than the trough. According to the vertical distributions of 239 + 240 Pu activities and 240Pu/ 239Pu atom ratios in these cores, it was concluded that sediment mixing was the dominant process in controlling profiles of plutonium in this area. Faster mixing in the coastal samples has homogenized the entire 240Pu/ 239Pu ratio record today; slightly slower mixing and less scavenging in the Okinawa Trough have left the surface sediment ratios closer

  15. Fabrication of high exposure nuclear fuel pellets

    DOEpatents

    Frederickson, James R.

    1987-01-01

    A method is disclosed for making a fuel pellet for a nuclear reactor. A mixture is prepared of PuO.sub.2 and UO.sub.2 powders, where the mixture contains at least about 30% PuO.sub.2, and where at least about 12% of the Pu is the Pu.sup.240 isotope. To this mixture is added about 0.3 to about 5% of a binder having a melting point of at least about 250.degree. F. The mixture is pressed to form a slug and the slug is granulated. Up to about 4.7% of a lubricant having a melting point of at least about 330.degree. F. is added to the granulated slug. Both the binder and the lubricant are selected from a group consisting of polyvinyl carboxylate, polyvinyl alcohol, naturally occurring high molecular weight cellulosic polymers, chemically modified high molecular weight cellulosic polymers, and mixtures thereof. The mixture is pressed to form a pellet and the pellet is sintered.

  16. Differential die-away technique for determination of the fissile contents in spent fuel assembly

    SciTech Connect

    Lee, Tachoon; Menlove, Howard O; Swinhoe, Nartyn T; Tobin, Stephen J

    2010-01-01

    Monte Carlo simulations were performed for the differential die-away (DDA) technique to quantify its capability to measure the fissile contents in spent fuel assemblies of 64 different cases in terms of initial enrichment, burnup, and cooling time. The DDA count rate varies according to the contents of fissile isotopes such as {sup 235}U, {sup 239}Pu, and {sup 241}Pu contained in the spent fuel assembly. The effective {sup 239}Pu concept was introduced to quantify the total fissile mass of spent fuel by weighting the relative signal contributions of {sup 235}U and {sup 241}Pu compared to that of {sup 239}Pu. The Monte Carlo simulation results show that the count rate of the DDA instrument for a spent fuel assembly of 4% initial enrichment, 45 GWD/MTU burnup, and 5 year cooling time is {approx} 9.8 x 10{sup 4} counts per second (c/s) with the 100-Hz repeated interrogation pattern of 0 to 10 {micro}s interrogation, 0.2 ms to 1 ms counting time, and 1 x 10{sup 9} n/s neutron source. The {sup 244}Cm neutron background count rate for this counting time scheme is {approx} 1 x 10{sup 4} c/s, and thus the signal to background ratio is {approx}10.

  17. Antioxidant phenolic compounds from Pu-erh tea.

    PubMed

    Zhang, Hai Ming; Wang, Cheng Fang; Shen, Sheng Min; Wang, Gang Li; Liu, Peng; Liu, Zi Mu; Wang, Yong Yan; Du, Shu Shan; Liu, Zhi Long; Deng, Zhi Wei

    2012-11-27

    Eight compounds were isolated from the water extract of Pu-erh tea and their structures were elucidated by NMR and MS as gallic acid (1), (+)-catechin (2), (−)-epicatechin (3), (−)-epicatechin-3-O-gallate (4), (−)-epigallocatechin-3-O-gallate (5), (−)-epiafzelechin- 3-O-gallate (6), kaempferol (7), and quercetin (8). Their in vitro antioxidant activities were assessed by the DPPH and ABTS scavenging methods with microplate assays. The relative order of DPPH scavenging capacity for these compounds was compound 8 > compound 7 > compound 1 > compound 6 > compound 4 ≈ compound 5 > compound 2 > VC (reference) > compound 3, and that of ABTS scavenging capacity was compound 1 > compound 2 > compound 7 ≈ compound 8 > compound 6 > compound 5 > compound 4 > VC (reference) > compound 3. The results showed that these phenolic compounds contributed to the antioxidant activity of Pu-erh tea.

  18. First results from PuMa: The Dutch Pulsar machine

    NASA Astrophysics Data System (ADS)

    Stappers, B. W.; Ramachandran, R.; Kouwenhoven, M.; Voute, L.

    1998-12-01

    A new pulsar machine, called PuMa, has been developed in the Netherlands and has recently been installed at the Westerbork Synthesis Radio Telescope. PuMa takes advantage of high speed DSPs to carry out FFTs of the incoming data rather than the traditional filter approach to sampling the bandpass. This enables the formation of very fine frequency channels thus greatly improving the ability to correct for dispersion across the bandpass. The system is also very versatile allowing the approriate number channels to be chosen for a particular experiment. As the data are sampled at 20 MHz very high time resolution can also be acheived. It is also possible to record 10 MHz of raw bandwidth which can be coherently dedispersed off-line. I will present the initial results from a number of projects which we are undertaking, including a globular cluster survey, high resolution pulse studies and polarimetry.

  19. Np and Pu Sorption to Manganese Oxide Minerals

    SciTech Connect

    Zhao, P; Johnson, M R; Roberts, S K; Zavarin, M

    2005-08-30

    Manganese oxide minerals are a significant component of the fracture lining mineralogy at Yucca Mountain (Carlos et al., 1993) and within the tuff-confining unit at Yucca Flat (Prothro, 1998), Pahute Mesa (Drellack et al., 1997), and other locations at the Nevada Test Site (NTS). Radionuclide sorption to manganese oxide minerals was not included in recent Lawrence Livermore National Laboratory (LLNL) hydrologic source term (HST) models which attempt to predict the migration behavior of radionuclides away from underground nuclear tests. However, experiments performed for the Yucca Mountain Program suggest that these minerals may control much of the retardation of certain radionuclides, particularly Np and Pu (Triay et al., 1991; Duff et al., 1999). As a result, recent HST model results may significantly overpredict radionuclide transport away from underground nuclear tests. The sorption model used in HST calculations performed at LLNL includes sorption to iron oxide, calcite, zeolite, smectite, and mica minerals (Zavarin and Bruton 2004a; 2004b). For the majority of radiologic source term (RST) radionuclides, we believe that this accounts for the dominant sorption processes controlling transport. However, for the case of Np, sorption is rather weak to all but the iron and manganese oxides (Figure 1). Thus, we can expect to significantly reduce predicted Np transport by accounting for Np sorption to manganese oxides. Similarly, Pu has been shown to be predominantly associated with manganese oxides in Yucca Mountain fractured tuffs (Duff et al., 1999). Recent results on colloid-facilitated Pu transport (Kersting and Reimus, 2003) also suggest that manganese oxide coatings on fracture surfaces may compete with colloids for Pu, thus reducing the effects of colloid-facilitated Pu transport (Figure 1b). The available data suggest that it is important to incorporate Np and Pu sorption to manganese oxides in reactive transport models. However, few data are available for

  20. Ultrasound-assisted reductive dissolution of CeO2 and PuO2 in the presence of Ti particles.

    PubMed

    Beaudoux, Xavier; Virot, Matthieu; Chave, Tony; Leturcq, Gilles; Jouan, Gauthier; Venault, Laurent; Moisy, Philippe; Nikitenko, Sergey I

    2016-06-07

    PuO2 is considered an important material for current and future nuclear fuel; however it is a very refractive compound towards dissolution. Among other techniques, its reprocessing can be performed via complexing dissolution in concentrated and boiling nitric acid containing hydrofluoric acid, or via oxidant dissolution in the presence of reagents with redox couples having high potentials such as Ce(iv)/Ce(iii), or Ag(ii)/Ag(i). Reductive dissolution can be performed under softer conditions and is considered an alternative to these methods which may suffer from several drawbacks (corrosion, effluent management, compatibility with nuclear waste disposal, etc.). In this study, a sonochemical and reductive approach is investigated for PuO2 dissolution under relatively mild conditions. At the first stage, the experiments are performed with CeO2 as an inactive surrogate for PuO2. The quantitative dissolution of both oxides can be achieved under ultrasound (20 kHz, 0.35-0.70 W mL(-1)) in 0.5 M HNO3/0.1 M [N2H5NO3]/2 M HCOOH sparged with Ar at 33-35 °C in the presence of Ti particles as a generating source of reductive species. Ultrasound enables the depassivation of the Ti surface (usually strongly passivated in nitric solutions) through acoustic cavitation which then allows further generation of the intermediate Ti(iii) reductive species. Dissolution rates and yields can be further increased with the injection of dilute fluoride aliquots (NH4F or HF) in the sonicated solution to favor Ti chemical depassivation. The rapid and complete dissolution of PuO2 under selected conditions is accompanied by Pu(iii) accumulation in solution.

  1. High temperature investigation of the solid/liquid transition in the PuO2-UO2-ZrO2 system

    NASA Astrophysics Data System (ADS)

    Quaini, A.; Guéneau, C.; Gossé, S.; Sundman, B.; Manara, D.; Smith, A. L.; Bottomley, D.; Lajarge, P.; Ernstberger, M.; Hodaj, F.

    2015-12-01

    The solid/liquid transitions in the quaternary U-Pu-Zr-O system are of great interest for the analysis of core meltdown accidents in Pressurised Water Reactors (PWR) fuelled with uranium-dioxide and MOX. During a severe accident the Zr-based cladding can become completely oxidised due to the interaction with the oxide fuel and the water coolant. In this framework, the present analysis is focused on the pseudo-ternary system UO2-PuO2-ZrO2. The melting/solidification behaviour of five pseudo-ternary and one pseudo-binary ((PuO2)0.50(ZrO2)0.50) compositions have been investigated experimentally by a laser heating method under pre-set atmospheres. The effects of an oxidising or reducing atmosphere on the observed melting/freezing temperatures, as well as the amount of UO2 in the sample, have been clearly identified for the different compositions. The oxygen-to-metal ratio is a key parameter affecting the melting/freezing temperature because of incongruent vaporisation effects. In parallel, a detailed thermodynamic model for the UO2-PuO2-ZrO2 system has been developed using the CALPHAD method, and thermodynamic calculations have been performed to interpret the present laser heating results, as well as the high temperature behaviour of the cubic (Pu,U,Zr)O2±x-c mixed oxide phase. A good agreement was obtained between the calculated and experimental data points. This work enables an improved understanding of the major factors relevant to severe accident in nuclear reactors.

  2. Slow Neutron Velocity Spectrometer Transmission Studies Of Pu

    DOE R&D Accomplishments Database

    Havens, W. W. Jr.; Melkonian, E.; Rainwater, L. J.; Levin, M.

    1951-05-28

    The slow neutron transmission of several samples of Pu has been investigated with the Columbia Neutron Velocity Spectrometer. Data are presented in two groups, those covering the energy region from 0 to 6 ev, and those covering the region above 6 ev. Below 6 ev the resolution was relatively good, and a detailed study of the cross section variation was made. Work above 6 ev consisted of merely locating levels and obtaining a rough idea of their strengths.

  3. Environmental and radiological safety studies: interaction of /sup 238/PuO/sub 2/ heat sources with terrestrial and aquatic environments. Progress report, April 1-June 30, 1981

    SciTech Connect

    Matlack, G.M.; Patterson, J.H.

    1981-09-01

    The containers for /sup 238/PuO/sub 2/ heat sources in radioisotope thermoelectric generators are designed with large safety factors to ensure they will withstand reentry from orbit and impact with the earth and safely contain the nuclear fuel until it is recovered. Existing designs have proved more than adequately safe, but the Space and Terrestrial Division of the Department of Energy Office of Advanced Nuclear Systems and Projects continually seeks more information about the heat sources to improve their safety. The work discussed here includes studies of the effects on the heat source of terrestrial and aquatic environments to obtain data for design of even safer systems. This report includes data from environmental chamber experiments that simulate terrestrial conditions, experiments to measure PuO/sub 2/ dissolution rates, soil column experiments to measure sorption of plutonium by soils, and several aquatic experiments.

  4. Steady-state fuel behavior modeling of nitride fuels in FRAPCON-EP

    NASA Astrophysics Data System (ADS)

    Feng, Bo; Karahan, Aydın; Kazimi, Mujid S.

    2012-08-01

    Fuel material properties and mechanistic fission gas models in FRAPCON-EP were updated to model the steady-state behavior of high-porosity nitride fuel operating at temperatures below half of the melting point. The fuel thermal conductivity and fuel thermal expansion models were updated with correlations for UN and (U,Pu)N fuels. Hot-pressing of the as-fabricated porosity was modeled as a function of the hydrostatic pressure and creep rate. The solid fission product swelling was assumed to increase linearly with burnup. Fission gas swelling constitutive models were updated to appropriately capture the intragranular gas bubble evolution in nitride fuel. Intergranular gas swelling was neglected due to the assumed high porosity of the fuel. The fission gas release behavior was modeled by fitting the fission gas diffusion coefficient in UN to FRAPCON's default fission gas release model. This fitted gas diffusion coefficient reflects the effects of porosity, burnup, operating temperature, fission rate, and bubble sink strength. Fission gas release and fuel swelling benchmarks against irradiation data were performed. The updated code was applied to UN fuel in typical PWR geometry and operating conditions, with an extended cycle length of 24 months. The results show that swelling of the nitride fuel up to 60 MWd/kg burnup did not lead to excessive straining of the cladding. Furthermore, this study showed that a porous (>15% porosity) nitride fuel pellet could achieve a much higher margin to failure from the cladding collapse and grid-to-rod fretting.

  5. Alternative Fuels

    EPA Pesticide Factsheets

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  6. PuMa, the first fully digital pulsar machine.

    NASA Astrophysics Data System (ADS)

    van Haren, P. C.; Voute, J. L. L.; Beijaard, T. D.; Driesens, D.; Kouwenhoven, M. L. A.; Langerak, J. J.

    2000-04-01

    Pulsars are neutron stars, rapidly rotating remains of supernova explosions, emitting bundles of broadband electromagnetic radiation. To carry out pulsar observations, two hurdles have to be overcome. Typically, the signal-to-noise ratio is poor, requiring long observations and large bandwidths. Next there is dispersion, causing the pulsating signals to smear out and calls for narrow signal bands. PuMa, the first Dutch pulsar machine, uses digital signal processing to split the incoming signal in up to thousands of narrow bands. The processor based design also increases flexibility as it allows different observational modes by loading the appropriate software into the signal processors. In total 192 SHARC processors (ADSP 21062) deliver the processing capacity. For PuMa a general purpose 6-processor SHARC board was developed, optimized for concurrent use of data busses. Other parts are commercially available components and all is joined in a VME environment. Mid 1998 PuMa was installed at the Westerbork Synthesis Radio Telescope in the Netherlands and its commissioning is completed.

  7. Pu-238 assay performance with the Canberra IQ3 system

    SciTech Connect

    Booth, L.; Gillespie, B.; Seaman, G.

    1997-11-01

    Canberra Industries has recently completed a demonstration project at the Westinghouse Savannah River Site (WSRC) to characterize 55-gallon drums containing Pu-238 contaminated waste. The goal of this project was to detect and quantify Pu-238 contaminated waste. The goal of this project was to detect and quantify Pu-238 waste to detection limits of less than 50 nCi/g using gamma assay techniques. This would permit reclassification of these drums from transuranic (TRU) waste to low-level waste (LLW). The instrument used for this assay was a Canberra IQ3 high sensitivity gamma assay system, mounted in a trailer. The results of the measurements demonstrate achievement of detection levels as low as 1 nCi/g for low density waste drums, and good correlation with known concentrations in several test drums. In addition, the data demonstrates significant advantages for using large area low-energy germanium detectors for achieving the lowest possible MDAs for gamma rays in the 80-250 keV range. 1 fig., 2 tabs.

  8. Fuel pin

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

    1989-10-03

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  9. Fuel pin

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

    1989-01-01

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  10. Fuel pin

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

    1987-11-24

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  11. Fuel pin

    SciTech Connect

    Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

    1987-11-24

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  12. Determination of Pu isotopes and 241Am in a reference fallout material using SF-ICP-MS.

    PubMed

    Zheng, Jian; Zhang, Yongsan; Yamada, Masatoshi; Wu, Fengchang; Igarashi, Yasuhito; Hirose, Katsumi

    2011-07-01

    This paper reports on the characterisation of activities of Pu and (241)Am, and Pu isotopic composition in a reference fallout material prepared by the Meteorological Research Institute (MRI), Japan, from samples collected at 14 stations throughout Japan in 1963-1979. The acid leaching and total digestion were used to compare whether there is difference in Pu and (241)Am activities and Pu isotopic composition between these two methods. The results of activities of (239+240)Pu and (241)Pu, and Pu isotopic composition have been reported in the previous work (Sci. Total Environ. 2010, 408, 1139-1144). In this study, the (241)Am activity and (241)Am/(239+240)Pu activity ratio in the reference fallout material are reported, and the usefulness of Pu atom ratios and (241)Am/(239+240)Pu activity ratio for source identification is discussed.

  13. Retention of 239Pu in the mouse lung and estimation of consequent dose following inhalation of sized 239PuO2.

    PubMed

    Morgan, A; Black, A; Moores, S R; Lambert, B E

    1984-08-01

    The retention of 239Pu in the lungs of SAS/4 mice following inhalation exposure to sized 239PuO2 particles is described. When the initial alveolar deposition (IAD) was less than 200 Bq, retention of 239Pu could be described by a two-component exponential expression, about 90% being cleared with a half-time of about 40 days and the remainder with a half-time of about 150 days. Similar amounts of 239Pu were retained up to 3 months with IADs greater than 800 Bq, but clearance was impaired thereafter, the half-time of the second component increasing to about 720 days. The retention of 239Pu was independent of the particle size of the administered 239PuO2. Studies of the retention of 239Pu by individual lobes indicated that there were intrinsic interlobar differences which were enhanced at higher IADs. Lung clearance was also studied by the measurement of 239Pu in feces excreted by groups of mice in the period immediately prior to sacrifice. The estimation of radiation dose to lung is discussed.

  14. Retention of /sup 239/Pu in the mouse lung and estimation of consequent dose following inhalation of sized /sup 239/PuO/sub 2/

    SciTech Connect

    Morgan, A.; Black, A.; Moores, S.R.; Lambert, B.E.

    1984-08-01

    The retention of /sup 239/Pu in the lungs of SAS/4 mice following inhalation exposure to sized /sup 239/PuO/sub 2/ particles is described. When the initial alveolar deposition (IAD) was <200 Bq, retention of /sup 239/Pu could be described by a two-component exponential expression, about 90% being cleared with a half-time of about 40 days and the remainder with a half-time of about 150 days. Similar amounts of /sup 239/Pu were retained up to 3 months with IADs>800 Bq, but clearance was impaired thereafter, the half-time of the second component increasing to about 720 days. The retention of /sup 239/Pu was independent of the particle size of the administered /sup 239/PuO/sub 2/. Studies of the retention of /sup 239/Pu by individual lobes indicated that there were intrinsic interlobar differences which were enhanced at higher IADs. Lung clearance was also studied by the measurement of /sup 239/Pu in feces excreted by groups of mice in the period immediately prior to sacrifice. The estimation of radiation dose to lung is discussed.

  15. Carcinogenesis From Inhaled (PuO2)-Pu-239 in Beagles: Evidence for Radiation Homeostasis at Low Doses?

    SciTech Connect

    Fisher, Darrell R.; Weller, Richard E.

    2010-09-01

    From the early 1970s to the late 1980s, Pacific Northwest National Laboratory conducted life-span studies in beagle dogs on the biological effects of inhaled plutonium (239PuO2, 238PuO2, and 239Pu[NO3]4) to help predict risks associated with accidental intakes in workers. Years later, the purpose of the present follow-up study is to reassess the dose-response relationship for lung cancer induction in the 239PuO2 dogs compared to controls, with particular focus on the dose-response at low lung doses. A 239PuO2 aerosol (2.3 μm AMAD, 1.9 μm GSD) was administered to six groups of 20 young (18-month old) beagle dogs (10 males and 10 females) by inhalation at six different activity levels, as previously described in Laboratory reports. Control dogs were sham-exposed. In dose level 1, initial pulmonary lung depositions were 130 ± 48 Bq (3.5 ± 1.3 nCi), corresponding to 1 Bq g-1 lung tissue (0.029 ± 0.001 nCi g-1. Groups 2 through 6 received initial lung depositions (mean values) of 760, 2724, 10345, 37900, and 200000 Bq (22, 79, 300, 1100, and 5800 nCi) 239PuO2, respectively. For each dog, the absorbed dose to lungs was calculated from the initial lung burden and the final lung burden at time of death and lung mass, assuming a single, long-term retention function. Insoluble plutonium oxide exhibited long retention times in the lungs. Increased dose-dependent mortality due to lung cancer (bronchiolar-alveolar carcinoma, adenocarcinoma, epidermoid carcinoma) and radiation pneumonitis (highest exposures group) was observed in dogs exposed to 239PuO2. Calculated lung doses ranged from a few cGy in early-sacrificed dogs to 7764 cGy in dogs that experienced early deaths from radiation pneumonitis. Data were regrouped by lifetime lung dose and plotted as a function of lung tumor incidence. Lung tumor incidence in controls and zero-dose exposed dogs was 18% (5/28). However, no lung tumors were observed in 16 dogs with the lowest lung doses (8 to 22 cGy, mean 14.4 ± 7.6 c

  16. Safeguards Considerations for Thorium Fuel Cycles

    DOE PAGES

    Worrall, Louise G.; Worrall, Andrew; Flanagan, George F.; ...

    2016-04-21

    We report that by around 2025, thorium-based fuel cycles are likely to be deployed internationally. States such as China and India are pursuing research, development, and deployment pathways toward a number of commercial-scale thorium fuel cycles, and they are already building test reactors and the associated fuel cycle infrastructure. In the future, the potential exists for these emerging programs to sell, export, and deploy thorium fuel cycle technology in other states. Without technically adequate international safeguards protocols and measures in place, any future potential clandestine misuse of these fuel cycles could go undetected, compromising the deterrent value of these protocolsmore » and measures. The development of safeguards approaches for thorium-based fuel cycles is therefore a matter of some urgency. Yet, the focus of the international safeguards community remains mainly on safeguarding conventional 235U- and 239Pu-based fuel cycles while the safeguards challenges of thorium-uranium fuel cycles remain largely uninvestigated. This raises the following question: Is the International Atomic Energy Agency and international safeguards system ready for thorium fuel cycles? Furthermore, is the safeguards technology of today sufficiently mature to meet the verification challenges posed by thorium-based fuel cycles? In defining these and other related research questions, the objectives of this paper are to identify key safeguards considerations for thorium-based fuel cycles and to call for an early dialogue between the international safeguards and the nuclear fuel cycle communities to prepare for the potential safeguards challenges associated with these fuel cycles. In this paper, it is concluded that directed research and development programs are required to meet the identified safeguards challenges and to take timely action in preparation for the international deployment of thorium fuel cycles.« less

  17. Safeguards Considerations for Thorium Fuel Cycles

    SciTech Connect

    Worrall, Louise G.; Worrall, Andrew; Flanagan, George F.; Croft, Steven

    2016-04-21

    We report that by around 2025, thorium-based fuel cycles are likely to be deployed internationally. States such as China and India are pursuing research, development, and deployment pathways toward a number of commercial-scale thorium fuel cycles, and they are already building test reactors and the associated fuel cycle infrastructure. In the future, the potential exists for these emerging programs to sell, export, and deploy thorium fuel cycle technology in other states. Without technically adequate international safeguards protocols and measures in place, any future potential clandestine misuse of these fuel cycles could go undetected, compromising the deterrent value of these protocols and measures. The development of safeguards approaches for thorium-based fuel cycles is therefore a matter of some urgency. Yet, the focus of the international safeguards community remains mainly on safeguarding conventional 235U- and 239Pu-based fuel cycles while the safeguards challenges of thorium-uranium fuel cycles remain largely uninvestigated. This raises the following question: Is the International Atomic Energy Agency and international safeguards system ready for thorium fuel cycles? Furthermore, is the safeguards technology of today sufficiently mature to meet the verification challenges posed by thorium-based fuel cycles? In defining these and other related research questions, the objectives of this paper are to identify key safeguards considerations for thorium-based fuel cycles and to call for an early dialogue between the international safeguards and the nuclear fuel cycle communities to prepare for the potential safeguards challenges associated with these fuel cycles. In this paper, it is concluded that directed research and development programs are required to meet the identified safeguards challenges and to take timely action in preparation for the international deployment of thorium fuel cycles.

  18. Safeguards Considerations for Thorium Fuel Cycles

    SciTech Connect

    Worrall, Louise G.; Worrall, Andrew; Flanagan, George F.; Croft, Steven

    2016-04-21

    We report that by around 2025, thorium-based fuel cycles are likely to be deployed internationally. States such as China and India are pursuing research, development, and deployment pathways toward a number of commercial-scale thorium fuel cycles, and they are already building test reactors and the associated fuel cycle infrastructure. In the future, the potential exists for these emerging programs to sell, export, and deploy thorium fuel cycle technology in other states. Without technically adequate international safeguards protocols and measures in place, any future potential clandestine misuse of these fuel cycles could go undetected, compromising the deterrent value of these protocols and measures. The development of safeguards approaches for thorium-based fuel cycles is therefore a matter of some urgency. Yet, the focus of the international safeguards community remains mainly on safeguarding conventional 235U- and 239Pu-based fuel cycles while the safeguards challenges of thorium-uranium fuel cycles remain largely uninvestigated. This raises the following question: Is the International Atomic Energy Agency and international safeguards system ready for thorium fuel cycles? Furthermore, is the safeguards technology of today sufficiently mature to meet the verification challenges posed by thorium-based fuel cycles? In defining these and other related research questions, the objectives of this paper are to identify key safeguards considerations for thorium-based fuel cycles and to call for an early dialogue between the international safeguards and the nuclear fuel cycle communities to prepare for the potential safeguards challenges associated with these fuel cycles. In this paper, it is concluded that directed research and development programs are required to meet the identified safeguards challenges and to take timely action in preparation for the international deployment of thorium fuel cycles.

  19. The hematopoietic transcription factor PU.1 regulates RANK gene expression in myeloid progenitors

    SciTech Connect

    Kwon, Oh Hyung; Lee, Chong-Kil; Lee, Young Ik; Paik, Sang-Gi; Lee, Hyun-Jun . E-mail: hjlee7@kribb.re.kr

    2005-09-23

    Osteoclasts are bone resorbing cells of hematopoietic origin. The hematopoietic transcription factor PU.1 is critical for osteoclastogenesis; however, the molecular mechanisms of PU.1-regulated osteoclastogenesis have not been explored. Here, we present evidence that the receptor activator of nuclear factor {kappa}B (RANK) gene that has been shown to be crucial for osteoclastogenesis is a transcriptional target of PU.1. The PU.1 {sup -/-} progenitor cells failed to express the RANK gene and reconstitution of PU.1 in these cells induced RANK expression. Treatment of the PU.1 reconstituted cells with M-CSF and RANKL further augmented the RANK gene expression. To explore the regulatory mechanism of the RANK gene expression by PU.1, we have cloned the human RANK promoter. Transient transfection assays have revealed that the 2.2-kb RANK promoter was functional in a monocyte line RAW264.7, whereas co-transfection of PU.1 transactivated the RANK promoter in HeLa cells. Taken together, these results suggest that PU.1 regulates the RANK gene transcription and this may represent one of the key roles of PU.1 in osteoclast differentiation.

  20. Pu-erh Tea Inhibits Tumor Cell Growth by Down-Regulating Mutant p53

    PubMed Central

    Zhao, Lanjun; Jia, Shuting; Tang, Wenru; Sheng, Jun; Luo, Ying

    2011-01-01

    Pu-erh tea is a kind of fermented tea with the incorporation of microorganisms’ metabolites. Unlike green tea, the chemical characteristics and bioactivities of Pu-erh tea are still not well understood. Using water extracts of Pu-erh tea, we analyzed the tumor cell growth inhibition activities on several genetically engineered mouse tumor cell lines. We found that at the concentration that did not affect wild type mouse embryo fibroblasts (MEFs) growth, Pu-erh tea extracts could inhibit tumor cell growth by down-regulated S phase and cause G1 or G2 arrest. Further study showed that Pu-erh tea extracts down-regulated the expression of mutant p53 in tumor cells at the protein level as well as mRNA level. The same concentration of Pu-erh tea solution did not cause p53 stabilization or activation of its downstream pathways in wild type cells. We also found that Pu-erh tea treatment could slightly down-regulate both HSP70 and HSP90 protein levels in tumor cells. These data revealed the action of Pu-erh tea on tumor cells and provided the possible mechanism for Pu-erh tea action, which explained its selectivity in inhibiting tumor cells without affecting wild type cells. Our data sheds light on the application of Pu-erh tea as an anti-tumor agent with low side effects. PMID:22174618

  1. Pu-erh tea inhibits tumor cell growth by down-regulating mutant p53.

    PubMed

    Zhao, Lanjun; Jia, Shuting; Tang, Wenru; Sheng, Jun; Luo, Ying

    2011-01-01

    Pu-erh tea is a kind of fermented tea with the incorporation of microorganisms' metabolites. Unlike green tea, the chemical characteristics and bioactivities of Pu-erh tea are still not well understood. Using water extracts of Pu-erh tea, we analyzed the tumor cell growth inhibition activities on several genetically engineered mouse tumor cell lines. We found that at the concentration that did not affect wild type mouse embryo fibroblasts (MEFs) growth, Pu-erh tea extracts could inhibit tumor cell growth by down-regulated S phase and cause G1 or G2 arrest. Further study showed that Pu-erh tea extracts down-regulated the expression of mutant p53 in tumor cells at the protein level as well as mRNA level. The same concentration of Pu-erh tea solution did not cause p53 stabilization or activation of its downstream pathways in wild type cells. We also found that Pu-erh tea treatment could slightly down-regulate both HSP70 and HSP90 protein levels in tumor cells. These data revealed the action of Pu-erh tea on tumor cells and provided the possible mechanism for Pu-erh tea action, which explained its selectivity in inhibiting tumor cells without affecting wild type cells. Our data sheds light on the application of Pu-erh tea as an anti-tumor agent with low side effects.

  2. A DFT+U study of Pu immobilization in Gd2Zr2O7

    NASA Astrophysics Data System (ADS)

    Zhao, F. A.; Xiao, H. Y.; Jiang, M.; Liu, Z. J.; Zu, X. T.

    2015-12-01

    The solubility of Pu in Gd2Zr2O7 has been investigated by the density functional theory plus Hubbard U correction. It is found that the formation of PuGdZr2O7, Gd2PuZrO7 and Gd2Pu1.5Zr0.5O7 are exothermic, whereas Pu0.5Gd1.5Zr2O7, Pu1.5Gd0.5Zr2O7 and Gd2Pu0.5Zr1.5O7 are energetically less stable than their respective separated states. The calculations show that both the Gd and Zr lattice sites can be substituted by the Pu, which is consistent with the immobilization behavior of uranium in Gd2Zr2O7 observed experimentally. The site preference of Pu in Gd2Zr2O7 is found to be dependent on the chemical environment, i.e., Pu prefers to substitute for Gd-site under Gd-rich and O2-rich conditions and for Zr-site under Zr-rich and O2-rich conditions.

  3. Vertical distribution and migration of global fallout Pu in forest soils in southwestern China.

    PubMed

    Bu, Wenting; Zheng, Jian; Guo, Qiuju; Uchida, Shigeo

    2014-10-01

    Soil samples collected in southwestern China were analyzed for Pu isotopes. The (240)Pu/(239)Pu atom ratios were around 0.18, which indicated the dominant source of global fallout. Consistent sub-surface maximums followed by exponential decline of (239+240)Pu activities in the soil cores were observed. Most of the Pu has still remained in the 0-10 cm layers since its deposition. Convection velocities and dispersion coefficients for Pu migration in the soils were estimated by the convection-dispersion equation (CDE) model. The effective convection velocities and effective dispersion coefficients ranged from 0.05 to 0.11 cm/y and from 0.06 to 0.29 cm(2)/y, respectively. Other factors that control the vertical migration of Pu in soil besides precipitation, soil particle size distribution and organic matter were suggested. Long-term migration behaviors of Pu in the soils were simulated. The results provide the Pu background baseline for further environmental monitoring and source identification of non-global fallout Pu inputs in the future.

  4. Ab-Initio Study on Plutonium Compounds Pu3M (M=Al, Ga, In), PuNp and Elemental Neptunium

    SciTech Connect

    Kutepov, A L

    2005-09-07

    Using spin-polarized relativistic density functional theory the electronic and magnetic structures for the plutonium compounds Pu{sub 3}M(M = Al; Ga; In) and PuNp have been investigated. For the first group of compounds the enhanced hybridization between Pu 5f and p-states of alloying element, as it has been found in spin-polarized calculations, is believed to be the main reason for the higher formation energies obtained in such kind of studies in comparison with the non-spin-polarized case. Also, comparative analysis of the actinides U, Np, Pu, Am, and Cm has been performed based on their electronic and magnetic structure. Some noticeable difference in the calculated magnetic structure was discovered between the actinide with local magnetic moments (Cm) and the actinides (Pu, Am) in which magnetic moments were found only in the calculations.

  5. Safety Report - Experiments 999 and 891 Muon Spin Relaxation in Pu and Pu-based Heavy Fermion Materials

    SciTech Connect

    Fluss, M; Heffner, R; Morris, G

    2004-04-23

    Experiment E999 proposes to carry out conventional muon spin relaxation ({mu}SR) measurements on solid samples of plutonium and plutonium alloys. Experiment 891 will be involved with {mu}SR experiments on PuCoGa{sub 5} and related Pu-based superconductors. Other than a dedicated cryostat to be provided by Los Alamos and a pumping station provided by Livermore, the experiments will use existing {mu}SR User Facility spectrometers and associated equipment such as detectors and electronics. The main topics of this report are therefore (1) the passivation of the samples with a polymer coating, (2) the design, fabrication and testing of a sealed titanium sample secondary encapsulation cell, (3) the transport of samples to and from TRIUMF and (4) the related on-site procedures for the safe handling of the encapsulated samples. Because both E999 and E891 share the same equipment and Pu-safety related issues, we are submitting a single safety report for both experiments.

  6. Development of Metallic Fuels for Actinide Transmutation

    SciTech Connect

    Hayes, Steven Lowe; Fielding, Randall Sidney; Benson, Michael Timothy; Chichester, Heather Jean MacLean; Carmack, William Jonathan

    2015-09-01

    Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimized for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10

  7. A method of measurement of (239)Pu, (240)Pu, (241)Pu in high U content marine sediments by sector field ICP-MS and its application to Fukushima sediment samples.

    PubMed

    Bu, Wenting; Zheng, Jian; Guo, Qiuju; Aono, Tatsuo; Tazoe, Hirofumi; Tagami, Keiko; Uchida, Shigeo; Yamada, Masatoshi

    2014-01-01

    An accurate and precise analytical method is highly needed for the determination of Pu isotopes in marine sediments for the long-term marine environment monitoring that is being done since the Fukushima Dai-ichi Nuclear Power Plant accident. The elimination of uranium from the sediment samples needs to be carefully checked. We established an analytical method based on anion-exchange chromatography and SF-ICP-MS in this work. A uranium decontamination factor of 2 × 10(6) was achieved, and the U concentrations in the final sample solutions were typically below 4 pg mL(-1), thus no extra correction of (238)U interferences from the Pu spectra was needed. The method was suitable for the analysis of (241)Pu in marine sediments using large sample amounts (>10 g). We validated the method by measuring marine sediment reference materials and our results agreed well with the certified and the literature values. Surface sediments and one sediment core sample collected after the nuclear accident were analyzed. The characterization of (241)Pu/(239)Pu atom ratios in the surface sediments and the vertical distribution of Pu isotopes showed that there was no detectable Pu contamination from the nuclear accident in the marine sediments collected 30 km off the plant site.

  8. In vitro dissolution of respirable aerosols of industrial uranium and plutonium mixed-oxide nuclear fuels.

    PubMed

    Eidson, A F; Mewhinney, J A

    1983-12-01

    Dissolution characteristics of mixed-oxide nuclear fuels are important considerations for prediction of biological behavior of inhaled particles. Four representative industrial mixed-oxide powders were obtained from fuel fabrication enclosures. Studies of the dissolution of Pu, Am and U from aerosol particles of these materials in a serum simulant solution and in 0.1M HCl showed: (1) dissolution occurred at a rapid rate initially and slowed at longer times, (2) greater percentages of U dissolved than Pu or Am: with the dissolution rates of U and Pu generally reflecting the physical nature of the UO2-PuO2 matrix, (3) the temperature history of industrial mixed-oxides could not be reliably related to Pu dissolution except for a 3-5% increase when incorporated into a solid solution by sintering at 1750 degrees C, and (4) dissolution in the serum simulant agreed with the in vivo UO2 dissolution rate and suggested the dominant role of mechanical processes in PuO2 clearance from the lung. The rapid initial dissolution rate was shown to be related, in part, to an altered surface layer. The advantages and uses of in vitro solubility data for estimation of biological behavior of inhaled industrial mixed oxides, such as assessing the use of chelation therapy and interpretation of urinary excretion data, are discussed. It was concluded that in vitro solubility tests were useful, simple and easily applied to individual materials potentially inhaled by humans.

  9. Fuel pump

    SciTech Connect

    Bellis, P.D.; Nesselrode, F.

    1991-04-16

    This patent describes a fuel pump. It includes: a fuel reservoir member, the fuel reservoir member being formed with fuel chambers, the chambers comprising an inlet chamber and an outlet chamber, means to supply fuel to the inlet chamber, means to deliver fuel from the outlet chamber to a point of use, the fuel reservoir member chambers also including a bypass chamber, means interconnecting the bypass chamber with the outlet chamber; the fuel pump also comprising pump means interconnecting the inlet chamber and the outlet chamber and adapted to suck fuel from the fuel supply means into the inlet chamber, through the pump means, out the outlet chamber, and to the fuel delivery means; the bypass chamber and the pump means providing two substantially separate paths of fuel flow in the fuel reservoir member, bypass plunger means normally closing off the flow of fuel through the bypass chamber one of the substantially separate paths including the fuel supply means and the fuel delivery means when the bypass plunger means is closed, the second of the substantially separate paths including the bypass chamber when the bypass plunger means is open, and all of the chambers and the interconnecting means therebetween being configured so as to create turbulence in the flow of any fuel supplied to the outlet chamber by the pump means and bypassed through the bypass chamber and the interconnecting means.

  10. Molecular Dynamics study of the mixed oxide fuel thermal conductivity

    NASA Astrophysics Data System (ADS)

    Nichenko, S.; Staicu, D.

    2013-08-01

    There is still no clear understanding of the plutonium content influence on the thermal conductivity behaviour of the (U,Pu) O2 MOX fuels. In this work Classical Molecular Dynamics (MD) was used to investigate the (U,Pu) O2 thermal conductivity in the whole concentration range and in the temperature range from 400 K to 1600 K. The Green-Kubo approach was used for the thermal conductivity calculation and an algorithm was proposed to improve the accuracy of the calculation. The obtained results are in good agreement with the literature experimental data and results of modelling of other authors. On the basis of the obtained results we give recommendations for the MOX thermal conductivity evaluation in the concentration range from pure UO2 up to pure PuO2.

  11. Enhancing BWR proliferation resistance fuel with minor actinides

    NASA Astrophysics Data System (ADS)

    Chang, Gray S.

    2009-03-01

    To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm 3) to the top (0.35 g/cm 3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides ( 237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in

  12. Enhancing BWR Proliferation Resistance Fuel with Minor Actinides

    SciTech Connect

    Gray S. Chang

    2009-03-01

    To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides (237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm3) to the top (0.35 g/cm3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides (237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms

  13. Characterization of ^{239,240}Pu Radionuclide Adsorption to Soil Particles and Mineral Dust Aerosols

    NASA Astrophysics Data System (ADS)

    Tatro, D. P.; Arimoto, R.; McMillan, N. J.; Barnes, M.

    2006-12-01

    The release of ^{239,240}Pu into the environment by nuclear weapons testing 50 years ago initiated the cyclic mobilization of Pu-contaminated soil particles via the resuspension of dust resulting in a widespread distribution of Pu and other radionuclides. It is unclear what enables the aeolian transport of Pu in the environment; plausible hypotheses of Pu binding to dust and soil particles include Pu adsorption to iron oxides/hydroxides, organic acids, or silicate minerals such as clays. To investigate the connections between surface soils, dust and radionuclides, samples of soil and/or dust were collected from the Project Gnome Site in Eddy County, NM, the Jemez Mountains near Los Alamos, NM, and two 50-year old attics and wind-blown dust in Big Spring, TX. This study tests the hypothesis that Pu is adsorbed onto Fe oxides and hydroxides that coat dust/soil particles. The samples are generally low in organic carbon (0.2 - 4.8%, except for the unburned Los Alamos sample at 9.4%), as measured by LOI (Loss On Ignition) at 360 °C. The citrate-bicarbonate-dithionite method (CDB) of Fe oxide removal, first proposed by Mehra and Jackson in 1960, was used to selectively extract Fe oxides from the samples while leaving silicate Fe intact. Chemical digestion of each sample creates two fractions, the extracted supernatant and a solid pellet residue. If the Pu were associated with Fe oxides, then Fe and Pu should both be selectively removed from the bulk sample during the CBD process, leaving the pellet depleted in Fe and Pu and the supernatant enriched. For Fe, this was confirmed by scanning electron microscope and petrographic analyses. Preliminary radiochemical analyses of Pu activity also verify this hypothesis. Pu activity is significantly lower in pellets than bulk samples (Pu activitypellet/Pu activitybulk average = 0.07, range 0.02-0.12); Pu activity in supernatants is significantly higher than in bulk samples (Pu activitysupernatant/Pu activitybulk average = 4

  14. Analysis of spent fuel assay with a lead slowing down spectrometer

    SciTech Connect

    Gavron, Victor I; Smith, L. Eric; Ressler, Jennifer J

    2010-10-29

    Assay of fissile materials in spent fuel that are produced or depleted during the operation of a reactor, is of paramount importance to nuclear materials accounting, verification of the reactor operation history, as well as for criticality considerations for storage. In order to prevent future proliferation following the spread of nuclear energy, we must develop accurate methods to assay large quantities of nuclear fuels. We analyze the potential of using a Lead Slowing Down Spectrometer for assaying spent fuel. We conclude that it is possible to design a system that will provide around 1% statistical precision in the determination of the {sup 239}Pu, {sup 241}Pu and {sup 235}U concentrations in a PWR spent-fuel assembly, for intermediate-to-high burnup levels, using commercial neutron sources, and a system of {sup 238}U threshold fission detectors. Pending further analysis of systematic errors, it is possible that missing pins can be detected, as can asymmetry in the fuel bundle.

  15. Analysis of spent fuel assay with a lead slowing down spectrometer

    SciTech Connect

    Gavron, Victor I; Smith, L Eric; Ressler, Jennifer J

    2008-01-01

    Assay of fissile materials in spent fuel that are produced or depleted during the operation of a reactor, is of paramount importance to nuclear materials accounting, verification of the reactor operation history, as well as for criticality considerations for storage. In order to prevent future proliferation following the spread of nuclear energy, we must develop accurate methods to assay large quantities of nuclear fuels. We analyze the potential of using a Lead Slowing Down Spectrometer for assaying spent fuel. We conclude that it is possible to design a system that will provide around 1% statistical precision in the determination of the {sup 239}Pu, {sup 241}Pu and {sup 235}U concentrations in a PWR spent-fuel assembly, for intermediate-to-high burnup levels, using commercial neutron sources, and a system of {sup 238}U threshold fission detectors. Pending further analysis of systematic errors, it is possible that missing pins can be detected, as can asymmetry in the fuel bundle.

  16. Vibrocompacted fuel for the liquid-metal FBR BOR 60

    SciTech Connect

    Herbig, R.; Rudolph, K.; Lindau, B. ); Skiba, O.V. )

    1992-01-01

    In view of fuel element refabrication on a highly radioactive and toxic level, a technology based on vibrocompacted granular fuel with consequent automation and remote control was developed. The main advantage of the granular technology is to enable the insertion of getter materials, e.g., metallic uranium, to control the U/Pu-O ratio, influence the thermodynamic state of the fuel, control the chemical state of fission products, and absorb impurities. The corrosion of the inner surface in this way may be lowered to a negligible depth. Future developments will lead to accomplished vibrotechnology by a controlled gravimetric dosing method.

  17. INERT-MATRIX FUEL: ACTINIDE ''BURINGIN'' AND DIRECT DISPOSAL

    SciTech Connect

    Rodney C. Ewing; Lumin Wang

    2002-10-30

    Excess actinides result from the dismantlement of nuclear weapons (Pu) and the reprocessing of commercial spent nuclear fuel (mainly 241 Am, 244 Cm and 237 Np). In Europe, Canada and Japan studies have determined much improved efficiencies for burnup of actinides using inert-matrix fuels. This innovative approach also considers the properties of the inert-matrix fuel as a nuclear waste form for direct disposal after one-cycle of burn-up. Direct disposal can considerably reduce cost, processing requirements, and radiation exposure to workers.

  18. Structure of mixed U(IV)-An(III) precursors synthesized by co-conversion methods (where An = Pu, Am or Cm)

    NASA Astrophysics Data System (ADS)

    Grandjean, S.; Arab-Chapelet, B.; Robisson, A. C.; Abraham, F.; Martin, Ph.; Dancausse, J.-Ph.; Herlet, N.; Léorier, C.

    2009-03-01

    Current concepts for future nuclear systems aim at improving the fuel cycle with the main following criteria: economy of resources, minimized volume and lower long-term potential radiotoxicity of ultimate wastes and proliferation risk reduction. Co-management of two (or more) actinides has recently been proposed for recycling reusable energy-producing actinides (mainly U and Pu) together, or for transmuting radiotoxic minor actinides within UO 2-based materials. Co-conversion processes play an important role by closing the actinide separation-purification operations and at the same time producing mixed actinide solid compounds for the fabrication of fresh fuel. Handling of actinides mixtures, from the initial solution up to the solid product, requires innovative synthesis methods and structures, particularly for the minor actinides such as americium and curium. Considering the different designs of future nuclear fuels, various uranium-actinide co-conversion routes are currently investigated in the CEA-ATALANTE facility.

  19. Bacterial Pu(V) reduction in the absence and presence of Fe(III)-NTA: modeling and experimental approach.

    PubMed

    Deo, Randhir P; Rittmann, Bruce E; Reed, Donald T

    2011-09-01

    Plutonium (Pu), a key contaminant at sites associated with the manufacture of nuclear weapons and with nuclear-energy wastes, can be precipitated to "immobilized" plutonium phases in systems that promote bioreduction. Ferric iron (Fe(3+)) is often present in contaminated sites, and its bioreduction to ferrous iron (Fe(2+)) may be involved in the reduction of Pu to forms that precipitate. Alternately, Pu can be reduced directly by the bacteria. Besides Fe, contaminated sites often contain strong complexing ligands, such as nitrilotriacetic acid (NTA). We used biogeochemical modeling to interpret the experimental fate of Pu in the absence and presence of ferric iron (Fe(3+)) and NTA under anaerobic conditions. In all cases, Shewanella alga BrY (S. alga) reduced Pu(V)(PuO(2) (+)) to Pu(III), and experimental evidence indicates that Pu(III) precipitated as PuPO(4(am).) In the absence of Fe(3+) and NTA, reduction of PuO(2) (+) was directly biotic, but modeling simulations support that PuO(2) (+) reduction in the presence of Fe(3+) and NTA was due to an abiotic stepwise reduction of PuO(2) (+) to Pu(4+), followed by reduction of Pu(4+) to Pu(3+), both through biogenically produced Fe(2+). This means that PuO(2) (+) reduction was slowed by first having Fe(3+) reduced to Fe(2+). Modeling results also show that the degree of PuPO(4(am)) precipitation depends on the NTA concentration. While precipitation out-competes complexation when NTA is present at the same or lower concentration than Pu, excess NTA can prevent precipitation of PuPO(4(am)).

  20. Roles of PU.1 in monocyte- and mast cell-specific gene regulation: PU.1 transactivates CIITA pIV in cooperation with IFN-gamma.

    PubMed

    Ito, Tomonobu; Nishiyama, Chiharu; Nakano, Nobuhiro; Nishiyama, Makoto; Usui, Yoshihiko; Takeda, Kazuyoshi; Kanada, Shunsuke; Fukuyama, Kanako; Akiba, Hisaya; Tokura, Tomoko; Hara, Mutsuko; Tsuboi, Ryoji; Ogawa, Hideoki; Okumura, Ko

    2009-07-01

    Over-expression of PU.1, a myeloid- and lymphoid-specific transcription factor belonging to the Ets family, induces monocyte-specific gene expression in mast cells. However, the effects of PU.1 on each target gene and the involvement of cytokine signaling in PU.1-mediated gene expression are largely unknown. In the present study, PU.1 was over-expressed in two different types of bone marrow-derived cultured mast cells (BMMCs): BMMCs cultured with IL-3 plus stem cell factor (SCF) and BMMCs cultured with pokeweed mitogen-stimulated spleen-conditioned medium (PWM-SCM). PU.1 over-expression induced expression of MHC class II, CD11b, CD11c and F4/80 on PWM-SCM-cultured BMMCs, whereas IL-3/SCF-cultured BMMCs expressed CD11b and F4/80, but not MHC class II or CD11c. When IFN-gamma was added to the IL-3/SCF-based medium, PU.1 transfectant acquired MHC class II expression, which was abolished by antibody neutralization or in Ifngr(-/-) BMMCs, through the induction of expression of the MHC class II transactivator, CIITA. Real-time PCR detected CIITA mRNA driven by the fourth promoter, pIV, and chromatin immunoprecipitation indicated direct binding of PU.1 to pIV in PU.1-over-expressing BMMCs. PU.1-over-expressing cells showed a marked increase in IL-6 production in response to LPS stimulation in both IL-3/SCF and PWM-SCM cultures. These results suggest that PU.1 overproduction alone is sufficient for both expression of CD11b and F4/80 and for amplification of LPS-induced IL-6 production. However, IFN-gamma stimulation is essential for PU.1-mediated transactivation of CIITA pIV. Reduced expression of mast cell-related molecules and transcription factors GATA-1/2 and up-regulation of C/EBPalpha in PU.1 transfectants indicate that enforced PU.1 suppresses mast cell-specific gene expression through these transcription factors.