Sample records for glenwood pu laski

  1. Glenwood Hot Springs Lodge Space Heating Low Temperature Geothermal...

    Open Energy Info (EERE)

    Lodge Space Heating Low Temperature Geothermal Facility Jump to: navigation, search Name Glenwood Hot Springs Lodge Space Heating Low Temperature Geothermal Facility Facility...

  2. Municipal geothermal heat utilization plan for Glenwood Springs, Colorado

    SciTech Connect (OSTI)

    Not Available

    1980-12-31T23:59:59.000Z

    A study has been made of the engineering and economic feasibility of utilizing the geothermal resource underlying Glenwood Springs Colorado, to heat a group of public buildings. The results have shown that the use of geothermal heat is indeed feasible when compared to the cost of natural gas. The proposed system is composed of a wellhead plate heat exchanger which feeds a closed distribution loop of treated water circulated to the buildings which form the load. The base case system was designed to supply twice the demand created by the seven public buildings in order to take advantage of some economies of scale. To increase the utilization factor of the available geothermal energy, a peaking boiler which burns natural gas is recommended. Disposal of the cooled brine would be via underground injection. Considerable study was done to examine the impact of reduced operating temperature on the existing heating systems. Several options to minimize this problem were identified. Economic analyses were completed to determine the present values of heat from the geothermal system and from the present natural gas over a 30 year projected system life. For the base case savings of over $1 million were shown. Sensitivities of the economics to capital cost, operating cost, system size and other parameters were calculated. For all reasonable assumptions, the geothermal system was cheaper. Financing alternatives were also examined. An extensive survey of all existing data on the geology of the study has led to the prediction of resource parameters. The wellhead temperature of produced fluid is suspected to lie between 140 and 180/sup 0/F (60 and 82/sup 0/C). Flowrates may be as high as 1000 gpm (3800 liters per minute) from a reservoir formation that is 300 ft (90 m) thick beginning about 500 ft (150 m) below the suggested drill site in the proposed Two Rivers Park.

  3. Elastic properties of Pu metal and Pu-Ga alloys

    SciTech Connect (OSTI)

    Soderlind, P; Landa, A; Klepeis, J E; Suzuki, Y; Migliori, A

    2010-01-05T23:59:59.000Z

    We present elastic properties, theoretical and experimental, of Pu metal and Pu-Ga ({delta}) alloys together with ab initio equilibrium equation-of-state for these systems. For the theoretical treatment we employ density-functional theory in conjunction with spin-orbit coupling and orbital polarization for the metal and coherent-potential approximation for the alloys. Pu and Pu-Ga alloys are also investigated experimentally using resonant ultrasound spectroscopy. We show that orbital correlations become more important proceeding from {alpha} {yields} {beta} {yields} {gamma} plutonium, thus suggesting increasing f-electron correlation (localization). For the {delta}-Pu-Ga alloys we find a softening with larger Ga content, i.e., atomic volume, bulk modulus, and elastic constants, suggest a weakened chemical bonding with addition of Ga. Our measurements confirm qualitatively the theory but uncertainties remain when comparing the model with experiments.

  4. Economical Production of Pu-238

    SciTech Connect (OSTI)

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

    2013-02-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2012-11-06T23:59:59.000Z

    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.

  6. Delayed neutron measurements for Th-232, Np-237, Pu-239, Pu-241 and depleted uranium

    E-Print Network [OSTI]

    Stone, Joseph C.

    2001-01-01T23:59:59.000Z

    The neutron emission rates from five very pure actinide samples (Th-232, Np-237, Pu-239, Pu-241 and depleted uranium) were measured following equilibrium irradiation in fast and thermal neutron fluxes. The relative abundances (alphas) for the first...

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

    SciTech Connect (OSTI)

    Beasley, T. M.

    1982-01-01T23:59:59.000Z

    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.

  8. Sources for Pu in near surface air

    SciTech Connect (OSTI)

    Hartmann, G.; Thom, C.; Baechmann, K.

    1989-01-01T23:59:59.000Z

    This paper provides evidence that most of the Pu in the near surface air today is due to resuspension. Vertical and particle size distribution in near surface air over a period of three years were measured. The seasonal variations of Pu in air and the influence of meteorological parameters on these variations are shown. Samples were taken before the Chernobyl accident in an area where only Pu fallout from the atmospheric nuclear tests of the early sixties occurs. The comparison of the behavior of Pu with other trace elements, which were also measured, showed similar behavior of Pu and elements like Ca, Ti and Fe in near surface air. This confirms that most Pu is resuspended because the main source for these elements in air is the soil surface. Resuspension factors and resuspension rate are estimated for all measured elements. A resuspension factor of 0.8 X 10(-8) m-1 and a resuspension rate of 0.09 X 10(-9) s-1 is calculated for Pu.

  9. MCSNA: Experimental Benchmarking of Pu Electronic Structure

    SciTech Connect (OSTI)

    Tobin, J G

    2007-01-29T23:59:59.000Z

    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.

  10. Redefining design criteria for Pu-238 gloveboxes

    SciTech Connect (OSTI)

    Acosta, S.V.

    1998-12-31T23:59:59.000Z

    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.

  11. A=3He (2010PU04)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

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  12. Pu Qian | Photosynthetic Antenna Research Center

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

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

    SciTech Connect (OSTI)

    Turchi, P; Kaufman, L; Liu, Z

    2006-06-30T23:59:59.000Z

    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.

  14. Complementary Pu Resuspension Study at Palomares, Spain

    SciTech Connect (OSTI)

    Shinn, J

    2002-10-01T23:59:59.000Z

    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{sup -10} m{sup -1} and agrees very well with the values between 10{sup -10} m{sup -1} and 10{sup -9} m{sup -1} previously reported. We observed a mean Pu/Am ratio of 7.1 with a relative variation of 30%, which compares well with a mean value of 6.5 for nearby plot 2-2. The resuspension rate (R) was in the middle of the range, 10{sup -11} s{sup -1} to 10{sup -12} s{sup -1} as observed in other stable sites, and indicates low potential for Pu redistribution.

  15. Pu speciation in actual and simulated aged wastes

    SciTech Connect (OSTI)

    Lezama-pacheco, Juan S [Los Alamos National Laboratory; Conradson, Steven D [Los Alamos National Laboratory

    2008-01-01T23:59:59.000Z

    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.

  16. A=3H (2010PU04)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

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  17. A=3Li (2010PU04)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

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  18. A=3n (2010PU04)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

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  19. Nuclear magnetic resonance offers new insights into Pu 239

    E-Print Network [OSTI]

    - 1 - Nuclear magnetic resonance offers new insights into Pu 239 May 29, 2012 Nuclear magnetic signal of plutonium 239's unique nuclear magnetic resonance signature has been detected by scientists on the subject, "Observation of 239 Pu Nuclear Magnetic Resonance," was published in the May 18 issue of Science

  20. Pu-238 fuel form activities, January 1-31, 1983

    SciTech Connect (OSTI)

    Not Available

    1983-03-01T23:59:59.000Z

    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.

  1. Sputtering yield of Pu bombarded by fission Fragments from Cf

    SciTech Connect (OSTI)

    Danagoulian, Areg [Los Alamos National Laboratory; Klein, Andreas [Los Alamos National Laboratory; Mcneil, Wendy V [Los Alamos National Laboratory; Yuan, Vincent W [Los Alamos National Laboratory

    2008-01-01T23:59:59.000Z

    We present results on the yield of sputtering of Pu atoms from a Pu foil, bombarded by fission fragments from a {sup 252}Cf source in transmission geometry. We have found the number of Pu atoms/incoming fission fragments ejected to be 63 {+-} 1. In addition, we show measurements of the sputtering yield as a function of distance from the central axis, which can be understood as an angular distribution of the yield. The results are quite surprising in light of the fact that the Pu foil is several times the thickness of the range of fission fragment particles in Pu. This indicates that models like the binary collision model are not sufficient to explain this behavior.

  2. Surrogate studies of the Pu-Be reaction II

    SciTech Connect (OSTI)

    Hanrahan, R.J. Jr.; Thoma, D.J.; Jacobson, L.A.; Zocco, T.G.; Lowery, J.L.; Pereyra, R.

    1997-09-01T23:59:59.000Z

    Unusual circumstances could result in contact between molten Pu and solid Be components. Since intimate contact between Pu and Be results in an intense neutron source via the ({alpha},n) reaction it is very difficult to study the kinetics of the Pu-Be interaction. The published Pu-Be phase diagram is characterized by a single intermetallic compound, PuBe{sub 13} which exists in equilibrium with all of the Pu allotropes, and two eutectics with no measurable solubility range in any of the solid phases. This pattern is known to be followed by all of the rare earths, and some of the actinides. Although most of these phase diagrams are poorly defined in terms of solubility and the location of the eutectics, all have been characterized to the extent of having a single intermetallic compound with a melting point in excess of 1,500 C. Consequently it seems reasonable that by studying the reactions of these other metals with Be, it should be possible to predict kinetics and mechanisms of the Pu-Be interaction using known physical and thermodynamic properties. The most obvious difference between Pu and potential surrogate elements is in their melting points. This is not the most important consideration however due to the presence of the two eutectics in each of the M-Be systems (where M is used to refer to any of the surrogates). In the last meeting of this series the authors reported the results of studies using Yb as a surrogate for Pu. Here they will discuss the results obtained using Sm and Ce and contrast them with the earlier results as well as with Pu-Be experiments conducted using very similar experimental conditions.

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

    SciTech Connect (OSTI)

    McCoy, Kevin [Areva NP; Blanpain, Patrick [AREVA NP SAS; Morris, Robert Noel [ORNL

    2014-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Migliori, Albert [Los Alamos National Laboratory; Betts, J [Los Alamos National Laboratory; Trugman, A [Los Alamos National Laboratory; Mielke, C H [Los Alamos National Laboratory; Mitchell, J N [Los Alamos National Laboratory; Ramos, M [Los Alamos National Laboratory; Stroe, I [WORXESTER, MA

    2009-01-01T23:59:59.000Z

    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.

  5. Glenwood Springs Amendments | Open Energy Information

    Open Energy Info (EERE)

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  6. Glenwood, Illinois: Energy Resources | Open Energy Information

    Open Energy Info (EERE)

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  7. Glenwood, Iowa: Energy Resources | Open Energy Information

    Open Energy Info (EERE)

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

    DOE Patents [OSTI]

    Tallent, Othar K. (Oak Ridge, TN); Mailen, James C. (Oak Ridge, TN); Bell, Jimmy T. (Kingston, TN); Arwood, Phillip C. (Harriman, TN)

    1982-01-01T23:59:59.000Z

    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.

  9. Re-publication of the data from the BILL magnetic spectrometer: The cumulative $?$ spectra of the fission products of $^{235}$U, $^{239}$Pu, and $^{241}$Pu

    E-Print Network [OSTI]

    N. Haag; W. Gelletly; F. von Feilitzsch; L. Oberauer; W. Potzel; K. Schreckenbach; A. A. Sonzogni

    2014-05-30T23:59:59.000Z

    In the 1980s, measurements of the cumulative $\\beta$ spectra of the fission products following the thermal neutron induced fission of $^{235}$U, $^{239}$Pu, and $^{241}$Pu were performed at the magnetic spectrometer BILL at the ILL in Grenoble. This data was published in bins of 250 keV. In this paper, we re-publish the original data in a binning of 50 keV for $^{235}$U and 100 keV for $^{239}$Pu and $^{241}$Pu.

  10. Neutron Capture Cross Section Measurement on $^{238}$Pu at DANCE

    SciTech Connect (OSTI)

    Chyzh, A; Wu, C Y

    2011-02-14T23:59:59.000Z

    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.

  11. Transuranic concentrations in reef and pelagic fish from the Marshall Islands. [/sup 239/Pu, /sup 240/Pu

    SciTech Connect (OSTI)

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

    1980-09-01T23:59:59.000Z

    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.

  12. Selection and Evaluation of a new Pu Density Measurement Fluid

    SciTech Connect (OSTI)

    Dziewinska, Krystyna [Los Alamos National Laboratory; Peters, Michael A [Los Alamos National Laboratory; Martinez, Patrick P [Los Alamos National Laboratory; Dziewinski, Jacek J [Los Alamos National Laboratory; Pugmire, David L [Los Alamos National Laboratory; Trujillo, Stephen M [Los Alamos National Laboratory; La Verne, Jake A [UNIV OF NOTRE DAME; Rajesh, P [UNIV OF NOTRE DAME

    2009-01-01T23:59:59.000Z

    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.

  13. Atomic Structure and Phase Transformations in Pu Alloys

    SciTech Connect (OSTI)

    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

    2008-04-28T23:59:59.000Z

    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.

  14. Delayed neutron emission measurements for U-235 and Pu-239 

    E-Print Network [OSTI]

    Chen, Yong

    2009-05-15T23:59:59.000Z

    , and a graphite-moderated counting system were constructed to perform all these experiments. The calculated values for the five-group U-235 delayed neutron parameters and the six-group Pu-239 delayed neutron parameters were compared with the values...

  15. Pu#ng Groundwater into Agro-IBIS

    E-Print Network [OSTI]

    Pu#ng Groundwater into Agro-IBIS Evren Soylu Chris Kucharik Steve Loheide step ­ Coupling of Agro-IBIS and a 3-D groundwater model · Conclusions #12;Why the RepresentaCon of Groundwater is Important? · Groundwater runoff is a dominant

  16. Delayed neutron emission measurements for U-235 and Pu-239

    E-Print Network [OSTI]

    Chen, Yong

    2009-05-15T23:59:59.000Z

    in the counting station. A set of highly purified actinide samples (U-235 and Pu-239) was irradiated in these experiments by using the Texas A&M University Nuclear Science Center Reactor. A fast pneumatic transfer system, an integrated computer control system...

  17. Spent Nuclear Fuel Self-Induced XRF to Predict Pu to U Content

    E-Print Network [OSTI]

    Stafford, Alissa Sarah

    2010-10-12T23:59:59.000Z

    ) in July 2008 and January 2009. These measurements successfully showed that it is possible to measure the Pu x-ray peak at 103.7 keV in PWR spent fuel (~1 percent Pu) using a planar HPGe detector. Prior to these measurement campaigns, the Pu peak has only...

  18. Characterization of Pu-238 heat source granule containment

    SciTech Connect (OSTI)

    Richardson Ii, P D [Los Alamos National Laboratory; Thronas, D L [Los Alamos National Laboratory; Romero, J P [Los Alamos National Laboratory; Sandoval, F E [Los Alamos National Laboratory; Neuman, A D [Los Alamos National Laboratory; Duncan, W S [Los Alamos National Laboratory

    2008-01-01T23:59:59.000Z

    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.

  19. Corrosion of a Pu-doped zirconolite-rich ceramic

    SciTech Connect (OSTI)

    Bakel, A.J.; Buck, E.C.; Wolf, S.F.; Chamberlain, D.B.; Bates, J.K. [Argonne National Lab., IL (United States); Ebbinghaus, B.B. [Lawrence Livermore National Lab., CA (United States)

    1997-06-01T23:59:59.000Z

    As part of a large Pu disposition program, a zirconolite-rich titanate ceramic is being developed at Lawrence Livermore National Laboratory (LLNL) as a possible immobilization material. This same material is being tested at Argonne National Laboratory (ANL). The goal of this study is to describe the corrosion behavior of this ceramic, particularly the release of Pu and Gd, using results from several static corrosion tests (MCC-1, PCT-A, and PCT-B). The release of relatively large amounts of Al, Ba, and Ca in short-term tests (3 day MCC-1 and 7 day PCT-A) indicates that these elements are released from grain boundaries or from highly soluble phases. Results from long-term (28, 98, and 182 day) PCT-B show that the releases of Al, Ba, and Ca decrease with time, the releases of U and Zr increase with time, and that the releases of Cs, Gd, Mo, and Pu remain fairly constant. Formation of alteration phases may lead to the decrease of Ba and Ca in leachate solutions. Due to the heterogeneous nature of the material, the formation of alteration phases, and the inherently low solubility of several elements, no element(s) could be recommended as good markers for the overall corrosion of this ceramic. Data show that, due to the complex nature of this material, the release of each element should be considered separately.

  20. AFS-2 FLOWSHEET MODIFICATIONS TO ADDRESS THE INGROWTH OF PU(VI) DURING METAL DISSOLUTION

    SciTech Connect (OSTI)

    Crapse, K.; Rudisill, T.; O'Rourke, P.; Kyser, E.

    2014-07-02T23:59:59.000Z

    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 to 95 °C reduced the oxidation rate of Pu(IV) to Pu(VI). For 8.1 M HNO{sub 3} simulated dissolution solutions, at near boiling conditions >35% Pu(VI) was present in 50 h while at 95 °C <10% Pu(VI) was present at 50 h. At near boiling temperatures, eliminating the presence of Cr and varying the HNO{sub 3} concentration in the range of 7–8.5 M had little effect on the rate of conversion of Pu(IV) to Pu(VI). HNO{sub 3} oxidation of Pu(IV) to Pu(VI) in a pure solution has been reported previously. Based on simulated dissolution experiments, this study concluded that dissolving Pu metal at 95°C using a 6 to 10 M HNO{sub 3} solution 0.05–0.2 M KF and 0–2 g/L B could reduce the rate of oxidation of Pu(IV) to Pu(VI) as compared to near boiling conditions. To demonstrate this flowsheet, two small-scale experiments were performed dissolving Pu metal up to 6.75 g/L. No Pu-containing residues were observed in the solutions after cooling. Using Pu metal dissolution rates measured during the experiments and a correlation developed by Holcomb, the time required to completely dissolve a batch of Pu metal in an H-Canyon dissolver using this flowsheet was estimated to require nearly 5 days (120 h). This value is reasonably consistent with an estimate based on the Batch 2 and 3 dissolution times in the 6.1D dissolver and Pu metal dissolution rates measured in this study and by Rudisill et al. Data from the present and previous studies show that the Pu metal dissolution rate decreases by a factor of approximately two when the temperature decreased from boiling (112 to 116°C) to 95°C. Therefore, the time required to dissolve a batch of Pu metal in an H-Canyon dissolver at 95°C would likely double (from 36 to 54 h) and require 72 to 108 h depending on the surface area of the Pu metal. Based on the experimental studies, a Pu metal dissolution flowsheet utilizing 6–10 M HNO{sub 3} containing 0.05–0.2 M KF (with 0–2 g/L B) at 95°C is recommended to reduce the oxidation of Pu(IV) to Pu(VI) as compared to near boiling conditions. The time required to completely di

  1. Removal of Pu238 from Neptunium Solution by Anion Exchange

    SciTech Connect (OSTI)

    KYSER, EDWARD

    2003-12-01T23:59:59.000Z

    A new anion flowsheet for use in HB-Line was tested in the lab with Reillex{trademark} HPQ for removal of Pu{sup 238} contamination from Np. Significant rejection of Pu{sup 238} was observed by washing with 6 to 12 bed volumes (BV) of reductive wash containing reduced nitric acid concentration along with both ferrous sulfamate (FS) and hydrazine. A shortened-height column was utilized in these tests to match changes in the plant equipment. Lab experiments scaled to plant batch sizes of 1500 to 2200 g Np were observed with modest losses for up-flow washing. Down-flow washing was observed to have high losses. The following are recommended conditions for removing Pu{sup 238} from Np solutions by anion exchange in HB-Line: (1) Feed conditions: Up-flow 6.4-8 M HNO{sub 3}, 0.02 M hydrazine, 0.05 M excess FS, less than 5 days storage of solution after FS addition. (2) Reductive Wash conditions: Up-flow 6-12 BV of 6.4 M HNO{sub 3}, 0.05 M FS, 0.05 M hydrazine. 1.8 mL/min/cm{sup 2} flowrate. (3) Decontamination Wash conditions: Up-flow 1-2 BV of 6.4-8 M HNO{sub 3}, no FS, no hydrazine. (4) Elution conditions: Down-flow 0.17 M HNO{sub 3}, 0.05 M hydrazine, no FS.

  2. Neutron Resonance Parameters and Covariance Matrix of 239Pu

    SciTech Connect (OSTI)

    Derrien, Herve [ORNL; Leal, Luiz C [ORNL; Larson, Nancy M [ORNL

    2008-08-01T23:59:59.000Z

    In order to obtain the resonance parameters in a single energy range and the corresponding covariance matrix, a reevaluation of 239Pu was performed with the code SAMMY. The most recent experimental data were analyzed or reanalyzed in the energy range thermal to 2.5 keV. The normalization of the fission cross section data was reconsidered by taking into account the most recent measurements of Weston et al. and Wagemans et al. A full resonance parameter covariance matrix was generated. The method used to obtain realistic uncertainties on the average cross section calculated by SAMMY or other processing codes was examined.

  3. First-principles elastic properties of (alpha)-Pu

    SciTech Connect (OSTI)

    Soderlind, P; Klepeis, J

    2009-02-18T23:59:59.000Z

    Density-functional electronic-structure calculations have been used to investigate the ambient pressure and low temperature elastic properties of the ground-state {alpha} phase of plutonium metal. The electronic structure and correlation effects are modeled within a fully relativistic antiferromagnetic treatment with a generalized gradient approximation for the electron exchange and correlation functional. The 13 independent elastic constants, for the monoclinic {alpha}-Pu system, are calculated for the observed geometry. A comparison of the results with measured data from recent resonant ultrasound spectroscopy for a cast sample is made.

  4. First-principles elastic properties of (alpha)-Pu

    SciTech Connect (OSTI)

    Soderlind, P; Klepeis, J E

    2008-11-04T23:59:59.000Z

    Density-functional electronic structure calculations have been used to investigate the ambient pressure and low temperature elastic properties of the ground-state {alpha} phase of plutonium metal. The electronic structure and correlation effects are modeled within a fully relativistic anti-ferromagnetic treatment with a generalized gradient approximation for the electron exchange and correlation functionals. The 13 independent elastic constants, for the monoclinic {alpha}-Pu system, are calculated for the observed geometry. A comparison of the results with measured data from resonant ultrasound spectroscopy for a cast sample is made.

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

    SciTech Connect (OSTI)

    Pavlovichev, A.M.

    2001-01-11T23:59:59.000Z

    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.

  6. Optimization Online - PuLP: A Linear Programming Toolkit for Python

    E-Print Network [OSTI]

    Stuart Mitchell

    2011-09-22T23:59:59.000Z

    Sep 22, 2011 ... PuLP is a high-level modelling library that leverages the power of the ... Keywords: mathematical programming; Python; modelling language.

  7. Development of U and Pu Co-Recovery Process (Co-Processing) for Future Reprocessing

    SciTech Connect (OSTI)

    Yamamoto, K.; Yanagibashi, F.; Fujimoto, I.; Sato, T.; Ohbu, T.; Taki, K.; Hayashi, S. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1194 (Japan)

    2013-07-01T23:59:59.000Z

    Co-processing process, which is the modified Purex process focused on co-recovery of Pu and U, has been studied at Operation Testing Laboratory, Tokai Reprocessing Plant in JAEA. The set up of the process was performed with flow-sheets study by process calculation to avoid Pu isolation in the whole process and to co-recover Pu/U product solution with a suitable Pu/U ratio (0.5< Pu/U <2). The initial Pu/U ratios of the feed solutions were taken as 1%, 3% and 20% considering the composition of the future spent fuels. The verification of the flow-sheets for each feed solutions were carried out with mixer-setters and active Pu/U feed solutions, focusing on the partitioning unit, and favorable back extraction performances of Pu accompanied by U were observed at all cases of the given feed solutions. According to these results, the co-processing process showed a good prospect to treat all kinds of future fuels from LWR, LWR-MOX and FBR, and a good prospect to be simplified by omitting the Pu/U purification cycle.

  8. Thermal Analysis of ZPPR High Pu Content Stored Fuel

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

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

    2014-01-01T23:59:59.000Z

    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. To: Department of Energy From: Andrew deLaski, Appliance Standards Awareness Project

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently AskedEnergyIssuesEnergy SolarRadioactiveI Disposal Sites AnnualSTATEMENT

  10. Microsoft Word - ex parte memo deLaski Harris.doc

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently Asked Questions forCheneyNovemberi CONTENTSSTATEMENT OF DAVID14,4.doc14.docA p p20 On

  11. 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 (OSTI)

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

    2009-02-23T23:59:59.000Z

    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 component to construct a central core zone with a composition closer to that in an LMFBR core with high burnup. The high {sup 240}Pu configuration was constructed for two reasons. First, the composition of the high {sup 240}Pu zone more closely matched the composition of LMFBR cores anticipated in design work in 1970. Second, comparison of measurements in the ZPR-6/7 uniform core with corresponding measurements in the high {sup 240}Pu zone provided an assessment of some of the effects of long-term {sup 240}Pu buildup in LMFBR cores. The uniform core version of ZPR-6/7 is evaluated in ZPR-LMFR-EXP-001. This document only addresses measurements in the high {sup 240}Pu core version of ZPR-6/7. Many types of measurements were performed as part of the ZPR-6/7 program. Measurements of criticality, sodium void worth, control rod worth and reaction rate distributions in the high {sup 240}Pu core configuration are evaluated here. For each category of measurements, the uncertainties are evaluated, and benchmark model data are provided.

  12. Spent Nuclear Fuel Self-Induced XRF to Predict Pu to U Content 

    E-Print Network [OSTI]

    Stafford, Alissa Sarah

    2010-10-12T23:59:59.000Z

    The quantification of plutonium (Pu) in spent nuclear fuel is an increasingly important safeguards issue. There exists an estimated worldwide 980 metric tons of Pu in the nuclear fuel cycle and the majority is in spent nuclear fuel waiting for long...

  13. Preparation and characterization of {sup 238}Pu-ceramics for radiation damage experiments

    SciTech Connect (OSTI)

    DM Strachan; RD Scheele; WC Buchmiller; JD Vienna; RL Sell; RJ Elovich

    2000-06-15T23:59:59.000Z

    As a result of treaty agreements between Russia and the US, portions of their respective plutonium and nuclear weapons stockpiles have been declared excess. In support of the US Department of Energy's 1998 decision to pursue immobilization of a portion of the remaining Pu in a titanate-based ceramic, the authors prepared nearly 200 radiation-damage test specimens of five Pu- and {sup 238}Pu-ceramics containing 10 mass% Pu to determine the effects of irradiation from the contained Pu and U on the ceramic. The five Pu-ceramics were (1) phase-pure pyrochlore [ideally, Ca(U, Pu)Ti{sub 2}O{sub 7}], (2) pyrochlore-rich baseline, (3) pyrochlore-rich baseline with impurities, (4) phase-pure zirconolite [ideally Ca(U, Pu)Ti{sub 2}O{sub 7}], and (5) a zirconolite-rich baseline. These ceramics were prepared with either normal weapons-grade Pu, which is predominantly {sup 239}Pu, or {sup 238}Pu. The {sup 238}Pu accelerates the radiation damage relative to the {sup 239}Pu because of its much higher specific activity. The authors were unsuccessful in preparing phase-pure (Pu, U) brannerite, which is the third crystalline phase present in the baseline immobilization form. Since these materials will contain {approximately}10 mass% Pu and about 20 mass% U, radiation damage to the crystalline structure of these materials will occur overtime. As the material becomes damaged from the decay of the Pu and U, it is possible for the material to swell as both the alpha particles and recoiling atoms rupture chemical bonds within the solid. As the material changes density, cracking, perhaps in the form of microcracks, may occur. If cracking occurs in ceramic that has been placed in a repository, the calculated rate of radionuclide release if the can has corroded would increase proportionately to the increase in surface area. To investigate the effects of radiation damage on the five ceramics prepared, the authors are storing the specimens at 20, 125, and 250 C until the {sup 238}Pu specimens become metamict and the damage saturates. They will characterize and test these specimens every 6 months by (1) monitoring the dimensions, (2) monitoring the geometric and pycnometric densities, (3) monitoring the appearance, (4) determining the normalized amount leached during a 3-day, static, 90 C leach test in high purity water, and (5) monitoring the crystal structure with x-ray diffraction crystallography (XRD). In this paper, the authors document the preparation and initial characterization of the materials that were made in this study. The initial XRD characterizations indicate that the phase assemblages appear to be correct with the exception of the {sup 238}Pu-zirconolite baseline material. They made this latter material using too much Pu, so this material contains unreacted PuO{sub 2}. The characterization of the physical properties of these materials found that the densities for all but three materials appear to be > 94% of theoretical, and only a few of the specimens have significant cracking. Those with cracking were the {sup 239}Pu-zirconolite specimens, which were sintered with a heat-up rate of 5 C/min. They sintered the {sup 238}Pu-zirconolite specimens with a heat-up rate of 2.5 C/min and obtained specimens with only minor surface cracking. Elemental releases during the 3-day MCC leach tests show that the normalized elemental releases depend on (1) whether the Pu is {sup 239}Pu or {sup 238}Pu, (2) the material type, and (3) the identity of the constituent. The effect of the Pu isotope in the ceramic is most dramatic for Pu release, with nominally 50 to 100 times more Pu activity released from the {sup 238}Pu specimens. This is unlikely to be an early indicator of radiation damage, because of the short time between specimen preparation and testing. In contrast greater amounts of Mo are released from the {sup 239}Pu specimens. Of the contained constituents, Ca Al, Pu, and U are the species found at relatively higher levels in the leachates.

  14. U.S.-Russian experts NATO collaborative research grant exchange visit meeting on excess Pu ceramics formulations and characterizations

    SciTech Connect (OSTI)

    Jardine, L.J., LLNL

    1998-11-24T23:59:59.000Z

    This document contains the agenda and meeting notes. Topics of discussion included US Pu disposition ceramics activities, Russian experience and proposals in Pu ceramics, and development of possible Russian ceramic proposals or collaborations.

  15. Consistent Data Assimilation of Isotopes: 242Pu and 105Pd

    SciTech Connect (OSTI)

    G. Palmiotti; H. Hiruta; M. Salvatores

    2012-09-01T23:59:59.000Z

    In this annual report we illustrate the methodology of the consistent data assimilation that allows to use the information coming from integral experiments for improving the basic nuclear parameters used in cross section evaluation. A series of integral experiments are analyzed using the EMPIRE evaluated files for 242Pu and 105Pd. In particular irradiation experiments (PROFIL-1 and -2, TRAPU-1, -2 and -3) provide information about capture cross sections, and a critical configuration, COSMO, where fission spectral indexes were measured, provides information about fission cross section. The observed discrepancies between calculated and experimental results are used in conjunction with the computed sensitivity coefficients and covariance matrix for nuclear parameters in a consistent data assimilation. The results obtained by the consistent data assimilation indicate that not so large modifications on some key identified nuclear parameters allow to obtain reasonable C/E. However, for some parameters such variations are outside the range of 1 s of their initial standard deviation. This can indicate a possible conflict between differential measurements (used to calculate the initial standard deviations) and the integral measurements used in the statistical data adjustment. Moreover, an inconsistency between the C/E of two sets of irradiation experiments (PROFIL and TRAPU) is observed for 242Pu. This is the end of this project funded by the Nuclear Physics Program of the DOE Office of Science. We can indicate that a proof of principle has been demonstrated for a few isotopes for this innovative methodology. However, we are still far from having explored all the possibilities and made this methodology to be considered proved and robust. In particular many issues are worth further investigation: • Non-linear effects • Flexibility of nuclear parameters in describing cross sections • Multi-isotope consistent assimilation • Consistency between differential and integral experiments

  16. Hydrothermal synthesis, structure, and magnetic properties of Pu(SeO{sub 3}){sub 2}

    SciTech Connect (OSTI)

    Bray, Travis H. [Department of Chemistry and Biochemistry and Center for Actinide Science, Auburn University, Auburn, AL 36849 (United States); Skanthakumar, S.; Soderholm, L. [Chemistry Division, Argonne National Laboratory, Argonne, IL 60439 (United States); Sykora, Richard E. [Department of Chemistry, University of South Alabama, Mobile, AL 36688 (United States); Haire, Richard G. [Chemical Sciences Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Albrecht-Schmitt, Thomas E. [Department of Chemistry and Biochemistry and Center for Actinide Science, Auburn University, Auburn, AL 36849 (United States)], E-mail: albreth@auburn.edu

    2008-03-15T23:59:59.000Z

    The reaction between PuO{sub 2} and SeO{sub 2} under mild hydrothermal conditions results in the formation of Pu(SeO{sub 3}){sub 2} as brick-red prisms. This compound adopts the Ce(SeO{sub 3}){sub 2} structure type, and consists of one-dimensional chains of edge-sharing [PuO{sub 8}] distorted bicapped trigonal prisms linked by [SeO{sub 3}] units into a three-dimensional network. Crystallographic data: Pu(SeO{sub 3}){sub 2}, monoclinic, space group P2{sub 1}/n, a=6.960(1) A, b=10.547(2) A, c=7.245(1) A, {beta}=106.880(9){sup o}, V=508.98(17) A{sup 3}, Z=4 (T=193 K), R(F)=2.92% for 83 parameters with 1140 reflections with I>2{sigma}(I). Magnetic susceptibility data for Pu(SeO{sub 3}){sub 2} are linear from 35 to 320 K and yield an effective moment of 2.71(5) {mu}{sub B} and a Weiss constant of -500(5) K. - Graphical abstract: A depiction of the three-dimensional structure of Pu(SeO{sub 3}){sub 2} formed from the interconnection of one-dimensional chains of edge-sharing PuO{sub 8} dodecahedra by selenite anions.

  17. 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 (OSTI)

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

    2007-10-01T23:59:59.000Z

    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-size LMFBRs. As a benchmark, ZPR-6/7 was devoid of many 'real' reactor features, such as simulated control rods and multiple enrichment zones, in its reference form. Those kinds of features were investigated experimentally in variants of the reference ZPR-6/7 or in other critical assemblies in the Demonstration Reactor Benchmark Program.

  18. RAPID DETERMINATION OF 237 NP AND PU ISOTOPES IN WATER BY INDUCTIVELY COUPLED PLASMA MASS SPECTROMETRY AND ALPHA SPECTROMETRY

    SciTech Connect (OSTI)

    Maxwell, S.; Jones, V.; Culligan, B.; Nichols, S.; Noyes, G.

    2010-06-23T23:59:59.000Z

    A new method that allows rapid preconcentration and separation of plutonium and neptunium in water samples was developed for the measurement of {sup 237}Np and Pu isotopes by inductively-coupled plasma mass spectrometry (ICP-MS) and alpha spectrometry; a hybrid approach. {sup 238}U can interfere with {sup 239}Pu measurement by ICP-MS as {sup 238}UH{sup +} mass overlap and {sup 237}Np via peak tailing. The method provide enhanced removal of uranium by separating Pu and Np initially on TEVA Resin, then moving Pu to DGA resin for additional removal of uranium. The decontamination factor for uranium from Pu is almost 100,000 and the decontamination factor for U from Np is greater than 10,000. This method uses stacked extraction chromatography cartridges and vacuum box technology to facilitate rapid separations. Preconcentration is performed using a streamlined calcium phosphate precipitation method. Purified solutions are split between ICP-MS and alpha spectrometry so that long and short-lived Pu isotopes can be measured successfully. The method allows for simultaneous extraction of 20 samples (including QC samples) in 4 to 6 hours, and can also be used for emergency response. {sup 239}Pu, {sup 242}Pu and {sup 237}Np were measured by ICP-MS, while {sup 236}Pu, {sup 238}Pu, and {sup 239}Pu were measured by alpha spectrometry.

  19. Osteosarcomas among beagles exposed to /sup 239/Pu

    SciTech Connect (OSTI)

    Whittemore, A.S.; McMillan, A.

    1982-04-01T23:59:59.000Z

    A Weibull distribution was fit to the osteosarcoma death times of beagles given single intravenous injections of /sup 239/Pu. For injected doses in the range 0-1..mu..Ci/kg the osteosarcoma incidence rate h(t) at t days after injection can be fit by a quadratic function of injected dose d: h(t) = 2.61 X 10/sup -18/ d/sup 2/t/sup 4.91/. The best-fitting linear function was rejected by the data (P < 0.001). A different formula for h(t), derived from a multistage theory for osteosarcoma induction, was also fit to these data. For this purpose microdosimetry calculations were used to estimate the dose to the cells at risk in the endosteal layer (endosteal dose). According to the best fit, h(t) is a quadratic function of endosteal dose at low doses. A linear dose-response relationship was again rejected. The absence of a linear component at low doses might be explained by the fact that 108 of the 185 animals injected at the lowest doses (<0.02 ..mu..Ci/kg) were still alive at the time these data were collected.

  20. Clearwater: Extensible, Flexible, Modular Code Generation Galen S. Swint, Calton Pu,

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Clearwater: Extensible, Flexible, Modular Code Generation Galen S. Swint, Calton Pu, Gueyoung Jung address the challenge of implementing code generators for two such DSLs that are flexible (resilient [4]. However, significant research challenges remain for generating flexible, reusable, and modular

  1. DETERMINATION OF 237NP AND PU ISOTOPES IN LARGE SOIL SAMPLES BY INDUCTIVELY COUPLED PLASMA MASS SPECTROMETRY

    SciTech Connect (OSTI)

    Maxwell, S.

    2010-07-26T23:59:59.000Z

    A new method for the determination of {sup 237}Np and Pu isotopes in large soil samples has been developed that provides enhanced uranium removal to facilitate assay by inductively coupled plasma mass spectrometry (ICP-MS). This method allows rapid preconcentration and separation of plutonium and neptunium in large soil samples for the measurement of {sup 237}Np and Pu isotopes by ICP-MS. {sup 238}U can interfere with {sup 239}Pu measurement by ICP-MS as {sup 238}UH{sup +} mass overlap and {sup 237}Np via {sup 238}U peak tailing. The method provides enhanced removal of uranium by separating Pu and Np initially on TEVA Resin, then transferring Pu to DGA resin for additional purification. The decontamination factor for removal of uranium from plutonium for this method is greater than 1 x 10{sup 6}. Alpha spectrometry can also be applied so that the shorter-lived {sup 238}Pu isotope can be measured successfully. {sup 239}Pu, {sup 242}Pu and {sup 237}Np were measured by ICP-MS, while {sup 236}Pu and {sup 238}Pu were measured by alpha spectrometry.

  2. Thermal Analysis of ZPPR High Pu Content Stored Fuel

    SciTech Connect (OSTI)

    Charles W. Solbrig; Chad Pope; Jason Andrus

    2014-09-01T23:59:59.000Z

    This paper estimates the temperature of high Pu content ZPPR fuel while in storage to determine the probablilty of fuel damage during storage. The Zero Power Physics Reactor (ZPPR) is an experimental reactor which has been decomissioned. It ran only at extremely low power, for testing nuclear reactor designs and was operated as a criticality facility from April 18, 1969 until decommissioned in 1990. Its fuel was manufactured in 1967 and has been in storage since the reactor was decomissioned. Heat is generated in the fuel due to Pu and Am decay and is a concern for possible fuel damage. Any damage to the cladding would be expected to lead to the fuel hydriding and oxidizing over a long period of storage as was described in the analysis of the damage to the ZPPR uranium fuel resulting in the fuel becoming unuseable and a large potential source of contamination. (Ref. Solbrig, 1994). A series of computer runs were made to scope out the range of temperatures that can occur in the ZPPR fuel in storage. The maximum calculated conservative fuel temperature is high (292 degrees C [558 degrees F]) in spite of the fact that the fuel element heat generation rates seem quite low, between 35 and 10 W for containers (called clamshells) full of fuel. However, the ZPPR storage bins, built for safeguards, are very effective insulators. The calculated clamshells and the cavity doors temperatures are also high. No record exists of people receiving skin burns by touching the cavity doors or clamshells, which indicates the computed temperatures may be higher than actual. (Note, gloves are worn when handling hotter clamshells.) Given the high calculated temperatures, a cursory measurement program was conducted to calibrate the calculated results. The measurement of bin doors, cavity doors, and clamshell temperatures would be easy to make if it were not for regulations resulting from security and potential contamination. Due to conservative assumptions in the model like high heat transfer contact resistance between contact surfaces (such as between the fuel and the clamshell), the calculated temperatures are intended to be overestimated. The temperatures of the stored fuel in a particular clamshell are dependent, among other parameters, on the distribution of fuel in the surrounding storage compartments, the heat generation rate of the fuel, and the orientation of fuel in the clamshell (parallel or perpendicular to the door). The distribution of fuel in this analysis was selected to give higher temperatures than actual distributions might give. Due to possible contamination and security concerns, fuel temperatures could not be measured but the bin doors, storage sleeve doors, and clamshell temperatures could be and were measured. The comparison of these three temperatures to the calculations indicates that the temperatures calculated with conservative assumptions are higher than the actual temperatures. This implies that the calculated fuel temperatures are higher than actual also. The maximum calculated fuel temperature with the most conservative assumptions (292 degrees C, (558 degrees F)) is significantly below the no fuel failure criterion of 600 degrees C (1,112 degrees F). Some fuel failures have occurred but these results indicate that the failures are not due to high temperatures encountered in fuel storage.

  3. Heterogeneous Reduction of PuO2 with Fe(II): Importance of the Fe(III) Reaction Product

    SciTech Connect (OSTI)

    Felmy, Andrew R.; Moore, Dean A.; Rosso, Kevin M.; Qafoku, Odeta; Rai, Dhanpat; Buck, Edgar C.; Ilton, Eugene S.

    2011-05-01T23:59:59.000Z

    Abstract Heterogeneous reduction of actinides in higher and more soluble oxidation states to lower more insoluble oxidation states by reductants such as Fe(II) has been the subject of intensive study for more than two decades. However, Fe(II)-induced reduction of sparingly soluble Pu(IV) to the more soluble lower oxidation state Pu(III) has been much less studied even though such reactions can potentially increase the mobility of Pu in the subsurface. Thermodynamic calculations are presented that show how differences in the free energy of various possible solid-phase Fe(III) reaction products can greatly influence aqueous Pu(III) concentrations resulting from reduction of PuO2(am) by Fe(II). We present the first experimental evidence that reduction of PuO2(am) to Pu(III) by Fe(II) was enhanced when the Fe(III) mineral goethite was spiked into the reaction. The effect of goethite on reduction of Pu(IV) was demonstrated by measuring the time-dependence of total aqueous Pu concentration, its oxidation state, and system pe/pH. We also re-evaluated established protocols for determining Pu(III) [(Pu(III) + Pu(IV)) - Pu(IV)] by using thenoyltrifluoroacetone (TTA) in toluene extractions; the study showed that it is important to eliminate dissolved oxygen from the TTA solutions for accurate determinations. More broadly, this study highlights the importance of the Fe(III) reaction product in actinide reduction rate and extent by Fe(II).

  4. Am phases in the matrix of a U-Pu-Zr alloy with Np, Am, and rare-earth elements

    SciTech Connect (OSTI)

    Dawn E Janney; J. Rory Kennedy; James W. Madden; Thomas P. O'Holleran

    2015-01-01T23:59:59.000Z

    Phases and microstructures in the matrix of an as-cast U-Pu-Zr alloy with 3 wt% Am, 2% Np, and 8% rare-earth elements were characterized by scanning and transmission electron microscopy. The matrix consists primarily of two phases, both of which contain Am: ?-(U, Np, Pu, Am) (~70 at% U, 5% Np, 14% Pu, 1% Am, and 10% Zr) and ?-(U, Np, Pu, Am)Zr2 (~25% U, 2% Np, 10-15% Pu, 1-2% Am, and 55-60 at% Zr). These phases are similar to those in U-Pu-Zr alloys, although the Zr content in ?-(U, Np, Pu, Am) is higher than that in ?-(U, Pu) and the Zr content in ?-(U, Np, Pu, Am)Zr2 is lower than that in ?-UZr2. Nanocrystalline actinide oxides with structures similar to UO2 occurred in some areas, but may have formed by reactions with the atmosphere during sample handling. Planar features consisting of a central zone of ?-(U, Np, Pu, Am) bracketed by zones of ?-(U, Np, Pu, Am)Zr2 bound irregular polygons ranging in size from a few micrometers to a few tens of micrometers across. The rest of the matrix consists of elongated domains of ?-(U, Np, Pu, Am) and ?-(U, Np, Pu, Am)Zr2. Each of these domains is a few tens of nanometers across and a few hundred nanometers long. The domains display strong preferred orientations involving areas a few hundred nanometers to a few micrometers across.

  5. Glenwood Springs Vapor Caves Pool & Spa Low Temperature Geothermal...

    Open Energy Info (EERE)

    syntax: * Display map Temperature No Data Listed Flow No Data Listed Capacity 1.00x106 Btuhr 0.300 MWt Annual Generation 7.00x109 Btuyr 2.10 GWhyr Load Factor 0.80 Contact...

  6. Glenwood Hot Springs Lodge Space Heating Low Temperature Geothermal

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdfGetec AG Contracting Jump to:Echo, Maryland:Glenwillow, Ohio: EnergyFacility |

  7. Glenwood Springs Resource Management Plan (1984) | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdfGetec AG Contracting Jump to:Echo, Maryland:Glenwillow, Ohio:

  8. City of Glenwood Springs, Colorado (Utility Company) | Open Energy

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand JumpConceptual Model, click here.TelluricPowerCity ofInformation CityIowaCityGlenCity of

  9. Glenwood Springs, Colorado: Energy Resources | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnual SiteofEvaluating AGeothermal/Exploration <Glacial Energy HoldingsGlacial LakesGlassSprings,

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

    SciTech Connect (OSTI)

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

    1994-11-01T23:59:59.000Z

    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.

  11. EIS-0299: Proposed Production of Plutonium-238 (Pu-238) for Use in Advanced Radioisotope Power Systems (RPS) for Space Missions

    Broader source: Energy.gov [DOE]

    This EIS is for the proposed production of plutonium-238 (Pu-238) using one or more DOE research reactors and facilities.

  12. PuReMD-GPU: A reactive molecular dynamics simulation package for GPUs

    SciTech Connect (OSTI)

    Kylasa, S.B., E-mail: skylasa@purdue.edu [Department of Elec. and Comp. Eng., Purdue University, West Lafayette, IN 47907 (United States); Aktulga, H.M., E-mail: hmaktulga@lbl.gov [Lawrence Berkeley National Laboratory, 1 Cyclotron Rd, MS 50F-1650, Berkeley, CA 94720 (United States); Grama, A.Y., E-mail: ayg@cs.purdue.edu [Department of Computer Science, Purdue University, West Lafayette, IN 47907 (United States)

    2014-09-01T23:59:59.000Z

    We present an efficient and highly accurate GP-GPU implementation of our community code, PuReMD, for reactive molecular dynamics simulations using the ReaxFF force field. PuReMD and its incorporation into LAMMPS (Reax/C) is used by a large number of research groups worldwide for simulating diverse systems ranging from biomembranes to explosives (RDX) at atomistic level of detail. The sub-femtosecond time-steps associated with ReaxFF strongly motivate significant improvements to per-timestep simulation time through effective use of GPUs. This paper presents, in detail, the design and implementation of PuReMD-GPU, which enables ReaxFF simulations on GPUs, as well as various performance optimization techniques we developed to obtain high performance on state-of-the-art hardware. Comprehensive experiments on model systems (bulk water and amorphous silica) are presented to quantify the performance improvements achieved by PuReMD-GPU and to verify its accuracy. In particular, our experiments show up to 16× improvement in runtime compared to our highly optimized CPU-only single-core ReaxFF implementation. PuReMD-GPU is a unique production code, and is currently available on request from the authors.

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

    SciTech Connect (OSTI)

    Duncan, A.; Kane, M.

    2009-11-24T23:59:59.000Z

    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 experience at PuFF indicates that the Pu-238 contamination was observed to move along surfaces and through High Efficiency Particulate Air (HEPA) filters over time. Recent research into the phenomenon known as alpha recoil offers a potential explanation for this observed behavior. Momentum is conserved when an alpha particle is ejected from a Pu-238 atom due to radioactive decay. Consequently, the entire particle of which that Pu-238 atom is a constituent experiences a movement similar to the recoil of a gun when a bullet is ejected. Furthermore, the particle often fractures in response to Pu-238 atom disintegration (yielding an alpha particle), with a small particle fragment also being ejected in order to conserve momentum. This process results in the continuous size reduction and transport of particles containing Pu-238 atoms, thus explaining movement of contamination along surfaces and through HEPA filters. A better understanding of the thermal behavior of {sup 238}PuO{sub 2} particles is needed to inform the planning process for the PuFF D&D project at the 235-F facility. There has been a concern that the surface temperature of individual particles may be high enough to cause problems with decontamination equipment and materials as a result of heat generation due to radioactive decay. A calculation under conservative assumptions shows that the surface temperature of particles less than about 100 {micro}m diameter is not appreciably above ambient. Since most particles in PuFF are on order of 1 {micro}m in diameter, the effect of particle surface temperature on decontamination equipment and materials is expected to be minimal. The result of this calculation also indicates that thermal imaging, which has been under consideration as a method to monitor the progress of system decontamination efforts would not likely be effective. The use of strippable coating was suggested as a possible alternative to other decontamination techniques. One particular system (i.e., Decon Gel 1101) may offer significant advantages over conventional liquid decontamination solut

  14. Verification of the content, isotopic composition and age of plutonium in Pu-Be neutron sources by gamma-spectrometry

    E-Print Network [OSTI]

    Cong Tam Nguyen

    2005-08-29T23:59:59.000Z

    A non-destructive, gamma-spectrometric method for verifying the plutonium content of Pu-Be neutron sources has been developed. It is also shown that the isotopic composition and the age of plutonium (Pu) can be determined in the intensive neutron field of these sources by the ``Multi-Group Analysis'' method. Gamma spectra were taken in the far-field of the sample, which was assumed to be cylindrical. The isotopic composition and the age of Pu were determined using a commercial implementation of the Multi-Group Analysis algorithm. The Pu content of the sources was evaluated from the count rates of the gamma-peaks of 239Pu, relying on the assumption that the gamma-rays are coming to the detector parallel to each other. The determination of the specific neutron yields and the problem of neutron damage to the detector are also discussed.

  15. Verification of the content, isotopic composition and age of plutonium in Pu-Be neutron sources by gamma-spectrometry

    E-Print Network [OSTI]

    Nguyen, C T

    2006-01-01T23:59:59.000Z

    A non-destructive, gamma-spectrometric method for verifying the plutonium content of Pu-Be neutron sources has been developed. It is also shown that the isotopic composition and the age of plutonium (Pu) can be determined in the intensive neutron field of these sources by the ``Multi-Group Analysis'' method. Gamma spectra were taken in the far-field of the sample, which was assumed to be cylindrical. The isotopic composition and the age of Pu were determined using a commercial implementation of the Multi-Group Analysis algorithm. The Pu content of the sources was evaluated from the count rates of the gamma-peaks of 239Pu, relying on the assumption that the gamma-rays are coming to the detector parallel to each other. The determination of the specific neutron yields and the problem of neutron damage to the detector are also discussed.

  16. Conceptual designs for a long term {sup 238}PuO{sub 2} storage vessel

    SciTech Connect (OSTI)

    Kwon, D.M.; Replogle, W.C.

    1996-08-01T23:59:59.000Z

    This is a report on conceptual designs for a long term, 250 years, storage container for plutonium oxide ([sup 238]PuO[sub 2]). These conceptual designs are based on the use of a quartz filter to release the helium generated during the plutonium decay. In this report a review of filter material selection, design concepts, thermal modeling, and filter performance are discussed.

  17. Hidden disorder in the ?'?? transformation of Pu-1.9 at.% Ga

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Jeffries, J. R.; Manley, M. E.; Wall, M. A.; Blobaum, K. J. M.; Schwartz, A. J.

    2012-06-01T23:59:59.000Z

    Enthalpy and entropy are thermodynamic quantities critical to determining how and at what temperature a phase transition occurs. At a phase transition, the enthalpy and temperature-weighted entropy differences between two phases are equal (?H=T?S), but there are materials where this balance has not been experimentally or theoretically realized, leading to the idea of hidden order and disorder. In a Pu-1.9 at. % Ga alloy, the ? phase is retained as a metastable state at room temperature, but at low temperatures, the ? phase yields to a mixed-phase microstructure of ?- and ?'-Pu. The previously measured sources of entropy associated withmore »the ?'?? transformation fail to sum to the entropy predicted theoretically. We report an experimental measurement of the entropy of the ?'?? transformation that corroborates the theoretical prediction, and implies that only about 65% of the entropy stabilizing the ? phase is accounted for, leaving a missing entropy of about 0.5 kB/atom. Some previously proposed mechanisms for generating entropy are discussed, but none seem capable of providing the necessary disorder to stabilize the ? phase. This hidden disorder represents multiple accessible states per atom within the ? phase of Pu that may not be included in our current understanding of the properties and phase stability of ?-Pu.« less

  18. IN FORMATION PU BLIC ATION SC H EME TITLE Agency plan for The Australian National University

    E-Print Network [OSTI]

    1 | IN FORMATION PU BLIC ATION SC H EME TITLE Agency plan for The Australian National University on its website. It will be directly accessible from the webpage foi.anu.edu.au and be identified possible, provide online content that can be searched by web browsers Provide a search function

  19. Production of 239 Pu from a natural Uranium disk and "hot" rock using a neutron howitzer

    E-Print Network [OSTI]

    Joseph Steiner; Aaron Anderson; Michael De Marco

    2008-05-23T23:59:59.000Z

    A neutron howitzer was used to produce 239Np from the targets of natural U and a hot rock. An intrinsic Germanium detector enabled the observations of the gamma rays in the decay of 239Np and a determination of its half life of 2.3 days. This shows that 239Pu had been produced in both targets

  20. Determination of the 242Pu Branching Ratio via Alpha-Gamma Coincidence

    SciTech Connect (OSTI)

    Wang, T F

    2012-05-24T23:59:59.000Z

    When the burn-up is high, the {sup 242}Pu isotopic content becomes more important. The traditional correlation method will fail. The {sup 242}Pu isotopic content in the sample plays an essential role if the neutron coincidence method is used to quantify the total amount of plutonium. In one of the earlier measurements we had a chance to measure an isotopic pure (> 99.95 %) {sup 242}Pu thick sample and realized that the difference in the branching ratio (BR) value among current nuclear data3) for the two important gamma-rays at 103.5-keV and 158.8-keV. In this study, the thick sample was counted on a 15% ORTEC safeguards type HPGe to further improve BR determination of the 159-keV gamma-ray. Furthermore, we have made a thin {sup 242}Pu sample from the thick sample and performed alpha-gamma coincidence measurements. Our preliminary gamma-ray BR results are 4.37(6) E-4, 2.79(8) E-5, and 2.25(8) E-6 for 44.9-keV, 103.5-keV, and 158.9-keV, respectively.

  1. Recovery of UO{sub 2}/PuO{sub 2} in IFR electrorefining process

    DOE Patents [OSTI]

    Tomczuk, Z.; Miller, W.E.

    1992-01-01T23:59:59.000Z

    This invention is comprised of a process for converting PuO{sub 2} and U0{sub 2} present in an electrorefiner to the chlorides, by contacting the PuO{sub 2} and U0{sub 2} with Li metal in the presence of an alkali metal chloride salt substantially free of rare earth and actinide chlorides for a time and at a temperature sufficient to convert the U0{sub 2} and PuO{sub 2} to metals while converting Li metal to Li{sub 2}O. Li{sub 2}O is removed either by reducing with rare earth metals or by providing an oxygen electrode for transporting 0{sub 2} out of the electrorefiner and a cathode, and thereafter applying an emf to the electrorefiner electrodes sufficient to cause the Li{sub 2}O to disassociate to 0{sub 2} and Li metal but insufficient to decompose the alkali metal chloride salt. The U and Pu and excess lithium are then converted to chlorides by reaction with CdCl{sub 2}.

  2. Recovery of UO[sub 2]/PuO[sub 2] in IFR electrorefining process

    DOE Patents [OSTI]

    Tomczuk, Z.; Miller, W.E.

    1994-10-18T23:59:59.000Z

    A process is described for converting PuO[sub 2] and UO[sub 2] present in an electrorefiner to the chlorides, by contacting the PuO[sub 2] and UO[sub 2] with Li metal in the presence of an alkali metal chloride salt substantially free of rare earth and actinide chlorides for a time and at a temperature sufficient to convert the UO[sub 2] and PuO[sub 2] to metals while converting Li metal to Li[sub 2]O. Li[sub 2]O is removed either by reducing with rare earth metals or by providing an oxygen electrode for transporting O[sub 2] out of the electrorefiner and a cathode, and thereafter applying an emf to the electrorefiner electrodes sufficient to cause the Li[sub 2]O to disassociate to O[sub 2] and Li metal but insufficient to decompose the alkali metal chloride salt. The U and Pu and excess lithium are then converted to chlorides by reaction with CdCl[sub 2].

  3. Recovery of UO.sub.2 /Pu O.sub.2 in IFR electrorefining process

    DOE Patents [OSTI]

    Tomczuk, Zygmunt (Lockport, IL); Miller, William E. (Naperville, IL)

    1994-01-01T23:59:59.000Z

    A process for converting PuO.sub.2 and UO.sub.2 present in an electrorefiner to the chlorides, by contacting the PuO.sub.2 and UO.sub.2 with Li metal in the presence of an alkali metal chloride salt substantially free of rare earth and actinide chlorides for a time and at a temperature sufficient to convert the UO.sub.2 and PuO.sub.2 to metals while converting Li metal to Li.sub.2 O. Li.sub.2 O is removed either by reducing with rare earth metals or by providing an oxygen electrode for transporting O.sub.2 out of the electrorefiner and a cathode, and thereafter applying an emf to the electrorefiner electrodes sufficient to cause the Li.sub.2 O to disassociate to O.sub.2 and Li metal but insufficient to decompose the alkali metal chloride salt. The U and Pu and excess lithium are then converted to chlorides by reaction with CdCl.sub.2.

  4. Hidden disorder in the ?'?? transformation of Pu-1.9 at.% Ga

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Jeffries, J. R.; Manley, M. E.; Wall, M. A.; Blobaum, K. J. M.; Schwartz, A. J.

    2012-06-01T23:59:59.000Z

    Enthalpy and entropy are thermodynamic quantities critical to determining how and at what temperature a phase transition occurs. At a phase transition, the enthalpy and temperature-weighted entropy differences between two phases are equal (?H=T?S), but there are materials where this balance has not been experimentally or theoretically realized, leading to the idea of hidden order and disorder. In a Pu-1.9 at. % Ga alloy, the ? phase is retained as a metastable state at room temperature, but at low temperatures, the ? phase yields to a mixed-phase microstructure of ?- and ?'-Pu. The previously measured sources of entropy associated with the ?'?? transformation fail to sum to the entropy predicted theoretically. We report an experimental measurement of the entropy of the ?'?? transformation that corroborates the theoretical prediction, and implies that only about 65% of the entropy stabilizing the ? phase is accounted for, leaving a missing entropy of about 0.5 kB/atom. Some previously proposed mechanisms for generating entropy are discussed, but none seem capable of providing the necessary disorder to stabilize the ? phase. This hidden disorder represents multiple accessible states per atom within the ? phase of Pu that may not be included in our current understanding of the properties and phase stability of ?-Pu.

  5. Estimates for Pu-239 loadings in burial ground culverts based on fast/slow neutron measurements

    SciTech Connect (OSTI)

    Winn, W.G.; Hochel, R.C.; Hofstetter, K.J.; Sigg, R.A.

    1989-08-15T23:59:59.000Z

    This report provides guideline estimates for Pu-239 mass loadings in selected burial ground culverts. The relatively high recorded Pu-239 contents of these culverts have been appraised as suspect relative to criticality concerns, because they were assayed only with the solid waste monitor (SWM) per gamma-ray counting. After 1985, subsequent waste was also assayed with the neutron coincidence counter (NCC), and a comparison of the assay methods showed that the NCC generally yielded higher assays than the SWM. These higher NCC readings signaled a need to conduct non-destructive/non-intrusive nuclear interrogations of these culverts, and a technical team conducted scoping measurements to illustrate potential assay methods based on neutron and/or gamma counting. A fast/slow neutron method has been developed to estimate the Pu-239 in the culverts. In addition, loading records include the SWM assays of all Pu-239 cuts of some of the culvert drums and these data are useful in estimating the corresponding NCC drum assays from NCC vs SWM data. Together, these methods yield predictions based on direct measurements and statistical inference.

  6. /sup 238/Pu fuel-form processes. Quarterly report, October-December 1981

    SciTech Connect (OSTI)

    Not Available

    1982-05-01T23:59:59.000Z

    Progress in the Savannah River /sup 238/Pu Fuel Form Program is summarized. Work during this period concentrated on the extensive cracking of the /sup 238/PuO/sub 2/ fuel form prior to encapsulation in the iridium containment shell for heat sources. This cracking results in increased recycle cost and decreased production efficiency. To better understand this cracking, Savannah River Laboratory (SRL) has made an extensive review of the development of /sup 238/PuO/sub 2/ fuel forms from small-scale Multi-hundred Watt (MHW) pellets through the current GPHS full-scale pellet production. Historically, /sup 238/PuO/sub 2/ fuel has almost always been uncracked after hot pressing in a graphite die, but has emerged cracked and fragile from the final heat-treatment furnace. The cracking tendency depends on the microstructure of the fuel form and on the hot pressing conditions used to fabricate it. In general, a microstructure of large intershard porosity is more desirable because it allows internal gas to escape more readily and it can absorb more reoxidation strain. Studies of the GPHS microstructure showed that the internal structures of typical GPHS Pellets fabricated at LANL and in the PEF differed significantly. The LANL pellets had severe density gradients and were extensively cracked.

  7. The Solubility of 242PuO2 in the Presence of Aqueous Fe(II): The Impact of Precipitate Preparation

    SciTech Connect (OSTI)

    Felmy, Andrew R.; Moore, Dean A.; Buck, Edgar C.; Conradson, Steven D.; Kukkadapu, Ravi K.; Sweet, Lucas E.; Abrecht, David G.; Ilton, Eugene S.

    2014-10-01T23:59:59.000Z

    The solubility of different forms of precipitated 242PuO2(am) were examined in solutions containing aqueous Fe(II) over a range of pH values. The first series of 242PuO2(am) suspensions were prepared from a 242Pu(IV) stock that had been treated with thenoyltrifluoroacetone (TTA) to remove the 241Am originating from the decay of 241Pu. These 242PuO2(am) suspensions showed much higher solubilities at the same pH value and Fe(II) concentration than previous studies using 239PuO2(am). X ray absorption fine structure (XAFS) spectroscopy of the precipitates showed a substantially reduced Pu-Pu backscatter over that previously observed in 242PuO2(am) precipitates, indicating that the 242PuO2(am) precipitates purified using TTA lacked the long range order previously found in 239PuO2(am) precipitates. The Pu(IV) stock solution was subsequently repurified using an ion exchange resin and an additional series of 242PuO2(am) precipitates prepared. These suspensions showed higher redox potentials and total aqueous Pu concentrations than the TTA purified stock solution. The higher redox potential and aqueous Pu concentrations were in general agreement with previous studies on 242PuO2(am) precipitates, presumably due to the removal of possible organic compounds originally present in the TTA purified stock. 242PuO2(am) suspensions prepared with both stock solutions showed almost identical solubilities in Fe(II) containing solutions even though the initial aqueous Pu concentrations before the addition of Fe(II) were orders of magnitude different. By examining the solubility of 242PuO2(am) prepared from both stocks in this way we have essentially approached equilibrium from both the undersaturated and oversaturated conditions. The final aqueous Pu concentrations are predictable using a chemical equilibrium model which includes the formation of a nanometer sized Fe(III) reaction product, identified in the 242PuO2(am) suspension both by use of 57Fe Mössbauer spectroscopy and transmission electron microscopy (TEM) analysis.

  8. Moisture corrections in neutron coincidence counting of PuO/sub 2/

    SciTech Connect (OSTI)

    Stewart, J.E.; Menlove, H.O.

    1987-01-01T23:59:59.000Z

    Passive neutron coincidence counting is capable of 1% assay accuracy for pure, well-characterized PuO/sub 2/ samples that contain plutonium masses from a few tens of grams to several kilograms. Moisture in the sample can significantly bias the assay high by changing the (..cap alpha..,n) neutron production, the sample multiplication, and the detection efficiency. Monte Carlo calculations and an analytical model of coincidence counting have been used to quantify the individual and cumulative effects of moisture biases for two PuO/sub 2/ sample sizes and a range of moisture levels from 0 to 9 wt %. Results of the calculations suggest a simple correction procedure for moisture bias that is effective from 0 to 3 wt % H/sub 2/O. The procedure requires that the moisture level in the sample be known before the coincidence measurement.

  9. Phonon dispersion curves determination in (delta)-phase Pu-Ga alloys

    SciTech Connect (OSTI)

    Wong, J; Clatterbuck, D; Occelli, F; Farber, D; Schwartz, A; Wall, M; Boro, C; Krisch, M; Beraud, A; Chiang, T; Xu, R; Hong, H; Zschack, P; Tamura, N

    2006-02-07T23:59:59.000Z

    We have designed and successfully employed a novel microbeam on large grain sample concept to conduct high resolution inelastic x-ray scattering (HRIXS) experiments to map the full phonon dispersion curves of an fcc {delta}-phase Pu-Ga alloy. This approach obviates experimental difficulties with conventional inelastic neutron scattering due to the high absorption cross section of the common {sup 239}Pu isotope and the non-availability of large (mm size) single crystal materials for Pu and its alloys. A classical Born von-Karman force constant model was used to model the experimental results, and no less than 4th nearest neighbor interactions had to be included to account for the observation. Several unusual features including, a large elastic anisotropy, a small shear elastic modulus, (C{sub 11}-C{sub 12})/2, a Kohn-like anomaly in the T{sub 1}[011] branch, and a pronounced softening of the T[111] branch towards the L point in the Brillouin are found. These features may be related to the phase transitions of plutonium and to strong coupling between the crystal structure and the 5f valence instabilities. Our results represent the first full phonon dispersions ever obtained for any Pu-bearing material, thus ending a 40-year quest for this fundamental data. The phonon data also provide a critical test for theoretical treatments of highly correlated 5f electron systems as exemplified by recent dynamical mean field theory (DMFT) calculations for {delta}-plutonium. We also conducted thermal diffuse scattering experiments to study the T(111) dispersion at low temperatures with an attempt to gain insight into bending of the T(111) branch in relationship to the {delta} {yields} {alpha}{prime} transformation.

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

    SciTech Connect (OSTI)

    Delegard, Calvin H.

    2011-09-29T23:59:59.000Z

    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.

  11. Genome-scale definition of the transcriptional programme associated with compromised PU.1 activity in Acute Myeloid Leukaemia

    E-Print Network [OSTI]

    Sive, Jonathan I.; Basilico, Silvia; Hannah, Rebecca; Kinston, Sarah J.; Calero-Nieto, Fernando J.; Göttgens, Berthold

    2015-01-01T23:59:59.000Z

    Transcriptional dysregulation is associated with haematological malignancy. Although mutations of the key haematopoietic transcription factor PU.1 are rare in human acute myeloid leukemia (AML), they are common in murine models of radiation...

  12. Code Generation Through Annotation of Macromolecular Structure Data John Biggs 1 , Calton Pu 1 , and Philip Bourne 2

    E-Print Network [OSTI]

    Bourne, Philip E.

    Code Generation Through Annotation of Macromolecular Structure Data John Biggs 1 , Calton Pu 1 & Technology P.O. Box 91000 Portland, OR 97291­1000 {biggs,calton}@cse.ogi.edu 2 San Diego Supercomputer Center

  13. NMIS With Gamma Spectrometry for Attributes of Pu and HEU, Explosives and Chemical Agents

    SciTech Connect (OSTI)

    Mihalczo, J. T.; Mattingly, J. K.; Mullens, J. A.; Neal, J. S.

    2002-05-10T23:59:59.000Z

    The concept for the system described herein is an active/passive Nuclear Materials Identification System{sup 2} (NMIS) that incorporates gamma ray spectrometry{sup 3}. This incorporation of gamma ray spectrometry would add existing capability into this system. This Multiple Attribute System can determine a wide variety of attributes for Pu and highly enriched uranium (HEU) of which a selected subset could be chosen. This system can be built using commercial off the shelf (COTS) components. NMIS systems are at All-Russian Scientific Research Institute of Experimental Physics (VNIIEF) and Russian Federal Nuclear Center Institute of Technical Physics, (VNIITF) and measurements with Pu have been performed at VNIIEF and analyzed successfully for mass and thickness of Pu. NMIS systems are being used successfully for HEU at the Y-12 National Security Complex. The use of active gamma ray spectrometry for high explosive HE and chemical agent detection is a well known activation analysis technique, and it is incorporated here. This report describes the system, explains the attribute determination methods for fissile materials, discusses technical issues to be resolved, discusses additional development needs, presents a schedule for building from COTS components, and assembly with existing components, and discusses implementation issues such as lack of need for facility modification and low radiation exposure.

  14. Effect of improved target designs on the sup 238 Pu production at the Fast Flux Test Reactor

    SciTech Connect (OSTI)

    Karnesky, R.A.; Wootan, D.W.; Jordheim, D.P. (Westinghouse Hanford Co., Richland, WA (United States))

    1991-11-01T23:59:59.000Z

    This paper present the results of a series of calculations made to determine the {sup 238}Pu production potential of several advanced target assembly designs in the Fast Flux Test Facility (FFTF). These calculations show that by using advanced target designs the intimately mix the {sup 237}Np target material with an yttrium hydride moderator, the FFTF has the potential of producing up to 30 kg of high-quality {sup 238}Pu per year.

  15. Technical Basis for Safe Operations with Pu-239 in NMS and S Facilities (F and H Areas)

    SciTech Connect (OSTI)

    Bronikowski, M.G.

    1999-03-18T23:59:59.000Z

    Plutonium-239 is now being processed in HB-Line and H-Canyon as well as FB-Line and F-Canyon. As part of the effort to upgrade the Authorization Basis for H Area facilities relative to nuclear criticality, a literature review of Pu polymer characteristics was conducted to establish a more quantitative vs. qualitative technical basis for safe operations. The results are also applicable to processing in F Area facilities.The chemistry of Pu polymer formation, precipitation, and depolymerization is complex. Establishing limits on acid concentrations of solutions or changing the valence to Pu(III) or Pu(VI) can prevent plutonium polymer formation in tanks in the B lines and canyons. For Pu(IV) solutions of 7 g/L or less, 0.22 M HNO3 prevents polymer formation at ambient temperature. This concentration should remain the minimum acid limit for the canyons and B lines when processing Pu-239 solutions. If the minimum acid concentration is compromised, the solution may need to be sampled and tested for the presence of polymer. If polymer is not detected, processing may proceed. If polymer is detected, adding HNO3 to a final concentration above 4 M is the safest method for handling the solution. The solution could also be heated to speed up the depolymerization process. Heating with > 4 M HNO3 will depolymerize the solution for further processing.Adsorption of Pu(IV) polymer onto the steel walls of canyon and B line tanks is likely to be 11 mg/cm2, a literature value for unpolished steel. This value will be confirmed by experimental work. Tank-to-tank transfers via steam jets are not expected to produce Pu(IV) polymer unless a larger than normal dilution occurs (e.g., >3 percent) at acidities below 0.4 M.

  16. Possible experimental evidence for the presence of double octupole states in {sup 240}Pu

    SciTech Connect (OSTI)

    Pascu, S.; Spieker, M.; Bucurescu, D.; Faestermann, T.; Hertenberger, R.; Skalacki, S.; Weber, S.; Wirth, H. F.; Zamfir, N. V.; Zilges, A. [Institut fuer Kernphysik, Universitaet zu Koeln, D-50937 Koeln, Germany and National Institute for Physics and Nuclear Engineering, R-77125, Bucharest-Magurele (Romania); Institut fuer Kernphysik, Universitaet zu Koeln, D-50937 Koeln (Germany); National Institute for Physics and Nuclear Engineering, R-77125, Bucharest-Magurele (Romania); Physik Department E12, Technische Universitaet Muenchen, D-85748 Garching (Germany); Sektion Physik, Ludwig-Maximilians-Universitaet Muenchen, D-85748 Garching (Germany); Institut fuer Kernphysik, Universitaet zu Koeln, D-50937 Koeln (Germany); Sektion Physik, Ludwig-Maximilians-Universitaet Muenchen, D-85748 Garching (Germany); National Institute for Physics and Nuclear Engineering, R-77125, Bucharest-Magurele (Romania); Institut fuer Kernphysik, Universitaet zu Koeln, D-50937 Koeln (Germany)

    2012-10-20T23:59:59.000Z

    Excited states in the {sup 240}Pu nucleus have been studied by means of the (p,t) reaction using the Q3D spectrometer and the focal plane detector from Munich. The comparison between experimental angular distributions and the DWBA calculations allowed the extraction of relative two-neutron transfer strengths. These observables may reveal important information about the structure of different states. The experimental two neutron strength for the 0{sup +}{sub 2} and 0{sup +}{sub 3} states is found in good agreement with the predictions of the IBA model, confirming the double octupole nature for the 0{sup +}{sub 2} state proposed in the previous studies.

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

    SciTech Connect (OSTI)

    Not Available

    1993-05-15T23:59:59.000Z

    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.

  18. Possible Fine Structure in the Delayed Neutron Yields in the Resonance Region for Pu-239

    SciTech Connect (OSTI)

    Ohsawa, Takaaki; Torii, Takayuki [Department of Electrical and Electronic Engineering, Kinki University, Higashi-osaka 577-8502 (Japan); Hambsch, Franz-Josef [EC-JRC-Institute for Reference Materials and Measurements, Retieseweg 111, B-2440 Geel (Belgium)

    2005-05-24T23:59:59.000Z

    A method of analysis of fluctuation in the delayed neutron yield on the basis of the multimodal fission model was applied to the low-energy resolved resonances for Pu-239. The present calculation using recent data of the fluctuation of the mode branching ratios for the resolved resonances showed both positive and negative resonance structure in the delayed neutron yield relative to the thermal neutron value. This is in contrast to the U-235 case, for which mainly negative dips of about -3.5% were predicted.

  19. RAPID SEPARATION METHOD FOR 237NP AND PU ISOTOPES IN LARGE SOIL SAMPLES

    SciTech Connect (OSTI)

    Maxwell, S.; Culligan, B.; Noyes, G.

    2010-07-26T23:59:59.000Z

    A new rapid method for the determination of {sup 237}Np and Pu isotopes in soil and sediment samples has been developed at the Savannah River Site Environmental Lab (Aiken, SC, USA) that can be used for large soil samples. The new soil method utilizes an acid leaching method, iron/titanium hydroxide precipitation, a lanthanum fluoride soil matrix removal step, and a rapid column separation process with TEVA Resin. The large soil matrix is removed easily and rapidly using this two simple precipitations with high chemical recoveries and effective removal of interferences. Vacuum box technology and rapid flow rates are used to reduce analytical time.

  20. Attempt to produce element 120 in the 244Pu + 58Fe reaction

    SciTech Connect (OSTI)

    Oganessian, Y T; Utyonkov, V K; Lobanov, Y V; Abdullin, F S; Polyakov, A N; Sagaidak, R N; Shorokovsky, I V; Tsyganov, Y S; Voinov, A A; Mezentsev, A N; Subbotin, V G; Sukhov, A M; Subotic, K; Zagrebaev, V I; Dmitriev, S N; Henderson, R A; Moody, K J; Kenneally, J M; Landrum, J H; Shaughnessy, D A; Stoyer, M A; Stoyer, N J; Wilk, P A

    2008-10-24T23:59:59.000Z

    An experiment aimed at the synthesis of isotopes of element 120 has been performed using the {sup 244}Pu({sup 58}Fe,xn){sup 302-x} 120 reaction. No decay chains consistent with fusion-evaporation reaction products were observed during an irradiation with a beam dose of 7.1 x 10{sup 18} 330-MeV {sup 58}Fe projectiles. The sensitivity of the experiment corresponds to a cross section of 0.4 pb for the detection of one decay.

  1. Microsoft Word - Template_SLAC Proprietary Use Agreement_PU 11_14_13

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville PowerCherries 82981-1cnHighandSWPA / SPRA / USACE625 FINALOptimization ofIDM UID: D_R22L3M verPU

  2. Summary of Puʻu ʻOʻo - Kupaianaha Eruption, Kilauea Volcano, Hawaii |

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere IRaghuraji Agro Industries PvtStratosolar Jump to: navigation, searchNewOpen Energy Information Puʻu

  3. Criticality Safety Evaluations on the Use of 200-gram Pu Mass Limit for RHWM Waste Storage Operations

    SciTech Connect (OSTI)

    Chou, P

    2011-12-14T23:59:59.000Z

    This work establishes the criticality safety technical basis to increase the fissile mass limit from 120 grams to 200 grams for Type A 55-gallon drums and their equivalents. Current RHWM fissile mass limit is 120 grams Pu for Type A 55-gallon containers and their equivalent. In order to increase the Type A 55-gallon drum limit to 200 grams, a few additional criticality safety control requirements are needed on moderators, reflectors, and array controls to ensure that the 200-gram Pu drums remain criticality safe with inadvertent criticality remains incredible. The purpose of this work is to analyze the use of 200-gram Pu drum mass limit for waste storage operations in Radioactive and Hazardous Waste Management (RHWM) Facilities. In this evaluation, the criticality safety controls associated with the 200-gram Pu drums are established for the RHWM waste storage operations. With the implementation of these criticality safety controls, the 200-gram Pu waste drum storage operations are demonstrated to be criticality safe and meet the double-contingency-principle requirement per DOE O 420.1.

  4. Nuclear Materials Identification System (NMIS) with Gamma Spectrometry for Attributes of Pu, HEU, and Detection of HE and Chemical Agents

    SciTech Connect (OSTI)

    Mihalczo, J. T.; Mattingly, J. K.; Mullens, J. A.; Neal, J. S.

    2002-05-01T23:59:59.000Z

    A combined Nuclear Materials Identification System (NMIS)-gamma ray spectrometry system can be used passively to obtain the following attributes of Pu: presence, fissile mass, 240/239 ratio, and metal vs. oxide. This system can also be used with a small, portable, DT neutron generator to measure the attributes of highly enriched uranium (HEU): presence, fissile mass, enrichment, metal vs. oxide; and detect the presence of high explosives (HE). For the passive system, time-dependent coincidence distributions can be used for the presence, fissile mass, metal vs. oxide for Pu, and gamma-ray spectrometry can be used for 239/240 ratio and presence. So presence can be confirmed by two methods. For the active system with a DT neutron generator, all four attributes for both Pu and HEU can be determined from various features of the time-dependent coincidence distribution measurements for both Pu and HEU. Active gamma ray spectrometry would also give presence and 240/239 ratio for Pu, enrichment for HEU, and metal vs. oxide for both. Active gamma ray spectrometry would determine the presence of HE. The various features of time-dependent coincidence distributions and gamma ray spectrometry that determine these attributes are discussed with some examples from previous determinations.

  5. GRPAUT: a program for Pu isotopic analysis (a user's guide). ISPO task A. 76

    SciTech Connect (OSTI)

    Fleissner, J G

    1981-01-30T23:59:59.000Z

    GRPAUT is a modular program for performing automated Pu isotopic analysis supplied to the International Atomic Energy Agency (IAEA) per ISPO Task A.76. Section I of this user's guide for GRPAUT presents an overview of the various programs and disk files that are used in performing a Pu isotopic analysis. Section II describes the program GRFEDT which is used in creating and editing the analysis parameter file that contains all the spectroscopic information needed at runtime by GRPAUT. An example of the dialog and output of GRFEDT is shown in Appendix B. Section III describes the operation of the various GRPAUT modules: GRPNL2, the peak stripping module; EFFCH2, the efficiency calculation module; and ISOAUT, the isotopic calculation module. (A description of the peak fitting methodology employed by GRPNL2 is presented in Appendix A.) Finally, Section IV outlines the procedure for determining the peak shape constants for a detector system and describes the operation of the program used to create and edit the peak shape parameter files. An output of GRPAUT, showing an example of a complete isotopic analysis, is presented in Appendix C. Source listings of all the Fortran programs supplied to the Agency under ISPO Task A.76 are contained in Appendix E.

  6. /sup 238/PuO/sub 2//Mo-50 wt% Re compatibility at 800 and 1000/sup 0/C

    SciTech Connect (OSTI)

    Schaeffer, D.R.; Teaney, P.E.

    1980-07-18T23:59:59.000Z

    The compatibility of Mo-50 wt % Re with /sup 238/PuO/sub 2/ was investigated after heat treatments of up to 720 days at 800/sup 0/C and 180 days at 1000/sup 0/C. At 800/sup 0/C, a 1-..mu..m thick, continuous layer of molybdenum oxide resulted. At 1000/sup 0/C, the oxide reaction product contained some plutonium and did not appear continuous. At 1000/sup 0/C, a layer of intermetallic formed at the Mo-Re edge, beneath the oxide layer, creating a barrier between the Mo-50 wt % Re and the /sup 238/PuO/sub 2/. The intermetallic layer was promoted by the iron impurity in the /sup 238/PuO/sub 2/.

  7. PLANTS AS BIO-MONITORS FOR 137CS, 238PU, 239, 240PU AND 40K AT THE SAVANNAH RIVER SITE

    SciTech Connect (OSTI)

    Caldwell, E.; Duff, M.; Ferguson, C.

    2010-12-16T23:59:59.000Z

    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.

  8. 137Cs(90Sr) and Pu isotopes in the Pacific Ocean sources & trends

    SciTech Connect (OSTI)

    Hamilton, T.F., Millies-Lacrox, J.C. [Service Mixte de Securite Radologique, Mondhery (France); Hong, G.H. [Korea Ocean Research and Development Institute, Ansan (Korea)

    1996-11-01T23:59:59.000Z

    The main source of artificial radioactivity in the world`s oceans can be attributed to worldwide fallout from atmospheric nuclear weapons testing. Measurements of selected artificial radionuclides in the Pacific Ocean were first conducted in the 1960`s where it was observed that fallout radioactivity had penetrated the deep ocean. Extensive studies carried out during the 1973-74 GEOSECS provided the first comprehensive data on the lateral and vertical distributions of {sup 9O}Sr, {sup 137}Cs and Pu isotopes in the Pacific on a basin wide scale. Estimates of radionuclide inventories in excess of amounts predicted to be delivered by global fallout alone were attributed to close-in fallout and tropospheric inputs from early U.S. tests conducted on Bikini and Enewetak Atolls in the Equatorial Pacific. In general, levels of fallout radionuclides (including {sup 9O}Sr, {sup 137}Cs and Pu isotopes) in the surface waters of the Pacific Ocean have decreased considerably over the past 4 decades and are now much more homogeneously distributed. Resuspension and the subsequent deposition of fallout radionuclides from previously deposited debris on land has become an important source term for the surface ocean. This can be clearly seen in measurements of fallout radionuclides in mineral aerosols over the Korean Peninsula (Yellow dust events). Radionuclides may also be transported from land to sea in river runoff-these transport mechanisms are more important in the Pacific Ocean where large quantities of river water and suspended sands/fluvial sediments reach the coastal zone. Another unique source of artificial radionuclides in the Pacific Ocean is derived from the slow resolubilization and transport of radionuclides deposited in contaminated lagoon and slope sediments near U.S. and French test sites. Although there is a small but significant flux of artificial radionuclides depositing on the sea floor, > 80% of the total 239, {sup 240}Pu inventory and > 95% of the total {sup 137}Cs inventory remains in the water column. Studies conducted through the 1980`s appear to be consistent with earlier findings and indicate that radionuclide inventories in mid-northern latitudes are at least a factor of two above those expected from global fallout alone. The long term persistence of close-in and/or stratospheric fallout from nuclear weapons testing in the Marshall Islands still appears to be the only plausible explanation for this anomaly.

  9. Isothermal Martensitic and Pressure-Induced ? to ?? Phase Transformations in a Pu-Ga Alloy

    SciTech Connect (OSTI)

    Schwartz, A J; Wall, M A; Farber, D L; Moore, K T; Blobaum, K M

    2007-09-10T23:59:59.000Z

    A Pu-2 at.% Ga alloy specimen is slowly compressed to {approx}1 GPa in a large volume moissanite anvil cell to induce the face-centered cubic {delta} to simple monoclinic {alpha}{prime} phase transformation. Optical microscopy, x-ray diffraction, and transmission electron microscopy of the specimen recovered to ambient pressure reveal that the vast majority of the microstructure consists of the {alpha}{prime} phase with grain sizes ranging from 10 nm to several hundred nm, with the remainder being {delta} phase dispersed between the {alpha}{prime} grains. This morphology is in contrast to the transformation product of the low-temperature isothermal martensite in which the lath-shaped {alpha}{prime} particles are {approx}20 {micro}m by 2 {micro}m.

  10. An evaluation of alternate production methods for Pu-238 general purpose heat source pellets

    SciTech Connect (OSTI)

    Mark Borland; Steve Frank

    2009-06-01T23:59:59.000Z

    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.

  11. Microscopic Calculation of 240Pu Scission with a Finite-Range Effective Force

    SciTech Connect (OSTI)

    Younes, W; Gogny, D

    2009-05-04T23:59:59.000Z

    Hartree-Fock-Bogoliubov calculations of hot fission in {sup 240}Pu have been performed with a newly-implemented code that uses the D1S finite-range effective interaction. The hot-scission line is identified in the quadrupole-octupole-moment coordinate space. Fission-fragment shapes are extracted from the calculations. A benchmark calculation for {sup 226}Th is obtained and compared to results in the literature. In addition, technical aspects of the use of HFB calculations for fission studies are examined in detail. In particular, the identification of scission configurations, the sensitivity of near-scission calculations to the choice of collective coordinates in the HFB iterations, and the formalism for the adjustment of collective-variable constraints are discussed. The power of the constraint-adjustment algorithm is illustrated with calculations near the critical scission configurations with up to seven simultaneous constraints.

  12. DOE Plutonium Disposition Study: Pu consumption in ALWRs. Volume 1, Final report

    SciTech Connect (OSTI)

    Not Available

    1993-05-15T23:59:59.000Z

    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 1, presents a technical description of the various elements of the System 80 + Standard Plant Design upon which the Plutonium Disposition Study was based. The System 80 + Standard Design is fully developed and directly suited to meeting the mission objectives for plutonium disposal. The bass U0{sub 2} plant design is discussed here.

  13. Exploring simultaneous single and coincident gamma-ray measurements for U/Pu assay in safeguards

    SciTech Connect (OSTI)

    Wang, T. F. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Horne, S. M. [Nuclear and Radiation Engineering Program, Mechanical Engineering Dept., Univ. of Texas at Austin, Austin, TX 78712 (United States); Henderson, R. A.; Roberts, K. E.; Vogt, D. K. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

    2011-07-01T23:59:59.000Z

    Using a broad range of gamma-ray uranium standards and two plutonium samples of known isotopic content, list mode gamma ray information from two Compton suppressed and one planar HPGe detectors were analyzed according to the time information of the signals. Interferences from Cs-137 were introduced. In this study, we extended singles measurements by exploring the potential of simultaneously using both singles and coincidence data for U/Pu assay. The main goals of this exploratory study are: 1) whether one will be able to use coincidence information in addition to the complicated 100-keV unfolding to obtain extra information of uranium and plutonium isotopic ratios, and 2) with higher energy interference gamma-rays from isotopes such as Cs-137, can the coincidence information help to provide the isotopic information. (authors)

  14. 8-group relative delayed neutron yields for monoenergetic neutron induced fission of sup 2 sup 3 sup 9 Pu

    E-Print Network [OSTI]

    Piksaikin, V M; Kazakov, L E; Korolev, G G; Roshchenko, V A; Tertychnyj, R G

    2001-01-01T23:59:59.000Z

    The energy dependence of the relative yield of delayed neutrons in an 8-group model representation was obtained for monoenergetic neutron induced fission of sup 2 sup 3 sup 9 Pu. A comparison of this data with the available experimental data by other was made in terms of the mean half-life of the delayed neutron precursors.

  15. Effect of fluoride in NTS groundwaters on the aqueous speciation of U, Np, Pu, Am and Eu

    SciTech Connect (OSTI)

    Bruton, C J; Nimz, G J

    2005-03-22T23:59:59.000Z

    To address SNJV concerns that fluoride in Nevada Test site (NTS) groundwaters may impact radionuclide speciation and transport, NTS water quality databases were obtained and scanned for analyses with high fluoride concentrations (> 10 mg/L). The aqueous speciation of nine representative samples of these groundwaters with added trace amounts of uranium (U), neptunium (Np), plutonium (Pu), americium (Am) and europium (Eu) was then calculated with the computer code EQ3NR assuming a temperature of 25 C, using currently available thermodynamic data for these species. Under conditions where U(VI), Np(V), Pu(IV), Am(III) and Eu(III) dominate, F complexes are insignificant (<1 mole %) for U, Np, Pu and Am. Eu-F complexes may be significant in groundwaters that lack bicarbonate, possess pH values less than about 7 at ambient temperatures, or contain F in extremely high concentrations (e.g. > 50 mg/L). The objective is to evaluate the extent to which fluoride in NTS groundwaters complex U(VI), Np(V), Pu(IV), Am(III) and Eu(III). The approach used is to screen existing databases of groundwater chemistry at NTS for waters with high fluoride concentrations and calculate the extent to which fluoride complexes with the nuclides of interest in these waters.

  16. Burnup estimation of fuel sourcing radioactive material based on monitored Cs and Pu isotopic activity ratios in Fukushima N. P. S. accident

    SciTech Connect (OSTI)

    Yamamoto, T.; Suzuki, M.; Ando, Y. [Japan Nuclear Energy Safety Organization, Toranomon Towers Office, 14F, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

    2012-07-01T23:59:59.000Z

    After the severe core damage of Fukushima Dai-Ichi Nuclear Power Station, radioactive material leaked from the reactor buildings. As part of monitoring of radioactivity in the site, measurements of radioactivity in soils at three fixed points have been performed for {sup 134}Cs and {sup 137}Cs with gamma-ray spectrometry and for Pu, Pu, and {sup 240}Pu with {alpha}-ray spectrometry. Correlations of radioactivity ratios of {sup 134}Cs to {sup 137}Cs, and {sup 238}Pu to the sum of {sup 239}Pu and {sup 240}Pu with fuel burnup were studied by using theoretical burnup calculations and measurements on isotopic inventories, and compared with the Cs and Pu radioactivity rations in the soils. The comparison indicated that the burnup of the fuel sourcing the radioactivity was from 18 to 38 GWd/t, which corresponded to that of the fuel in the highest power and, therefore, the highest decay heat in operating high-burnup fueled BWR cores. (authors)

  17. Rational Ligand Design for U(VI) and Pu(IV)

    SciTech Connect (OSTI)

    Szigethy, Geza

    2009-08-12T23:59:59.000Z

    Nuclear power is an attractive alternative to hydrocarbon-based energy production at a time when moving away from carbon-producing processes is widely accepted as a significant developmental need. Hence, the radioactive actinide power sources for this industry are necessarily becoming more widespread, which is accompanied by the increased risk of exposure to both biological and environmental systems. This, in turn, requires the development of technology designed to remove such radioactive threats efficiently and selectively from contaminated material, whether that be contained nuclear waste streams or the human body. Raymond and coworkers (University of California, Berkeley) have for decades investigated the interaction of biologically-inspired, hard Lewis-base ligands with high-valent, early-actinide cations. It has been established that such ligands bind strongly to the hard Lewis-acidic early actinides, and many poly-bidentate ligands have been developed and shown to be effective chelators of actinide contaminants in vivo. Work reported herein explores the effect of ligand geometry on the linear U(IV) dioxo dication (uranyl, UO{sub 2}{sup 2+}). The goal is to utilize rational ligand design to develop ligands that exhibit shape selectivity towards linear dioxo cations and provides thermodynamically favorable binding interactions. The uranyl complexes with a series of tetradentate 3-hydroxy-pyridin-2-one (3,2-HOPO) ligands were studied in both the crystalline state as well as in solution. Despite significant geometric differences, the uranyl affinities of these ligands vary only slightly but are better than DTPA, the only FDA-approved chelation therapy for actinide contamination. The terepthalamide (TAM) moiety was combined into tris-beidentate ligands with 1,2- and 3,2-HOPO moieties were combined into hexadentate ligands whose structural preferences and solution thermodynamics were measured with the uranyl cation. In addition to achieving coordinative saturation, these ligands exhibited increased uranyl affinity compared to bis-Me-3,2-HOPO ligands. This result is due in part to their increased denticity, but is primarily the result of the presence of the TAM moiety. In an effort to explore the relatively unexplored coordination chemistry of Pu(IV) with bidentate moieties, a series of Pu(IV) complexes were also crystallized using bidentate hydroxypyridinone and hydroxypyrone ligands. The geometries of these complexes are compared to that of the analogous Ce(IV) complexes. While in some cases these showed the expected structural similarities, some ligand systems led to significant coordination changes. A series of crystal structure analyses with Ce(IV) indicated that these differences are most likely the result of crystallization condition differences and solvent inclusion effects.

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

    SciTech Connect (OSTI)

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

    2014-07-01T23:59:59.000Z

    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 diffusivities estimated from the literature), then it is unlikely that fuel element bulging will occur.

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

    SciTech Connect (OSTI)

    Not Available

    2008-06-01T23:59:59.000Z

    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.

  20. Use of multivariate calibration for plutonium quantitation by the Pu(III) spectrophotometric method

    SciTech Connect (OSTI)

    Wangen, L.E.; Phillips, M.V.; Walker, L.F.

    1988-05-01T23:59:59.000Z

    Two new multivariate calibration methods for using all of the relevant spectral information are applied to the determination of plutonium. The analyte response signal originates from the absorbance spectrum of Pu(III)from 500 to 900 nm. Partial least squares (PLS) regression gives an average absolute error of 0.114 /+-/ 0.108 mg when predicting plutonium content of standards containing 65 to 90 mg total plutonium. PLS uses all of the signal in the spectrum and is a more robust calibration procedure than a method based on absorbances at five wavelengths. Another calibration procedure involving least squares curve fitting (LSCF) fits either the entire spectrum or individual spectral intervals derived from standards to spectra of unknowns. In addition, an arbitrary linear base line can be included. The best LSCF option for the same calibration and test set as used for PLS was the full spectrum (522 to 900 nm) with a linear base-line option. The average absolute error when predicting with LSCF was 0.130 /+-/ 0.092 mg plutonium. LSCF has an advantage over PLS in that the linear base line can account for certain types of interferences that have been observed for this plutonium assay procedure. An example is given. 6 refs., 3 figs., 5 tabs.

  1. Gadolinium-neutron-activation determination with a Pu-Be source

    SciTech Connect (OSTI)

    Konyaev, A.E.; Kositsyn, V.F.; Medvedev, A.B.; Rudenko, V.S.

    1988-05-01T23:59:59.000Z

    A nondestructive neutron activation method for determining gadolinium content for reactor construction materials was developed. The method uses a Pu-Be neutron source capable of giving 10/sup 8/ neutrons per second and the neutron reaction with a /sup 160/Gd target. To determine the flux attenuation, induced-activity distributions were measured along the radius with artificial compacts of Al/sub 2/O/sub 3/ + Gd/sub 2/O/sub 3/ specimens with varying gadolinium contents. The specimens were irradiated in unscreened and screened containers. The ratios of the unfiltered and filtered activities were not more than 1.06 +/- 0.04. The dependence of the gamma-ray absorption coefficient on gadolinium content and the effect of gadolinium content on the count rate due to /sup 161/Gd were determined. The nondestructive neutron-activation determination of gadolinium was possible for gadolinium concentrations where the radial induced-activity distribution was constant. The method for calculating the gamma-ray absorption coefficient was simple and reliable for measurement geometry close to 4pi. Neutron activation results agreed with chemical measurement within the error limits.

  2. Disposition of transuranic residues from plutonium isentropic compression experiment (Pu-ICE) constucted at Z machine

    SciTech Connect (OSTI)

    Goyal, Kapil K [Los Alamos National Laboratory; French, David M [Los Alamos National Laboratory; Humphrey, Betty J [WESTON SOLUTIONS INC.; Gluth, Jeffry [SNL

    2010-01-01T23:59:59.000Z

    In 1992, the U.S. Congress passed legislation to discontinue above- and below-ground testing of nuclear weapons. Because of this, the U.S. Department of Energy (DOE) must rely on laboratory experiments and computer-based calculations to verify the reliability of the nuclear stockpile. The Sandia National Laboratories/New Mexico (SNL/NM) Z machine was developed to support the science-based approach for mimicking nuclear explosions and stockpile stewardship. Plutonium (Pu) isotopes with greater than ninety-eight percent enrichment were used in the experiments. In May 2006, SNL/NM received authority that the Z Machine Isentropic Compression Experiments could commence. Los Alamos National Laboratory (LANL) provided the plutonium targets and loaded the target assemblies provided by SNL/NM. Three experiments were conducted from May through July 2006. The residues from each experiment, which weighed up to 913 pounds, were metallic and were packaged into a 55-gallon drum each. SNL/NM conducts the experiments and provides temporary storage for the drums until shipment to LANL for final waste certification for disposal at the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. This paper presents a comprehensive approach for documenting generator knowledge for characterization of waste in cooperation with scientists at the two laboratories and addresses a variety of essential topics.

  3. Assessment of a mechanistic model in U-Pu-Zr metallic alloy fuel fission-gas behavior simulations

    SciTech Connect (OSTI)

    Yun, D.; Rest, J.; Yacout, A. M. [Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States)

    2012-07-01T23:59:59.000Z

    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 look at 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 the simulation of fission gas release, fission gas and fission product induced swelling, and the evolution of the gas bubble size distribution in three different fuel zones: the outer {alpha}-U, the intermediate, and the inner {gamma}-U zones. Due to its special microstructural features, the {alpha}-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 consistent 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 {alpha}-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 {gamma}-U zone of U-Pu-Zr fuel are also compared to experimental measurements from available SEM micrographs. These comparisons show good agreements 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. In addition, the results of parametric studies for several key parameters are presented for both the intermediate zone and the {gamma}-U zone simulations. (authors)

  4. Intrinsic Nanoscience of ? Pu-Ga Alloys: Local Structure and Speciation, Collective Behavior, Nanoscale Heterogeneity, and Aging Mechanisms

    SciTech Connect (OSTI)

    Conradson, Steven D.; Bock, Nicolas; Castro, Julio M.; Conradson, Dylan R.; Cox, Lawrence E.; Dmowski, Wojtek; Dooley, David E.; Egami, Takeshi; Espinosa-Faller, Francisco J.; Freibert, Franz J.; Garcia-Adeva, Angel J.; Hess, Nancy J.; Holmstrom, Erik; Howell, Rafael C.; Katz, Barbara A.; Lashley, Jason C.; Martinez, Raymond J.; Moore, David P.; Morales, Luis A.; Olivas, J David; Pereyra, Ramiro A.; Ramos, Michael; Terry, Jeff H.; Villella, Phillip M.

    2014-04-24T23:59:59.000Z

    Because diffraction measurements are sensitive only to the long range average arrangement of the atoms in the coherent portion of a crystal, complementary local structure measurements are required for a complete understanding of the structure of a complex material. This is particularly an issue in solid solutions where even random distributions of a solute will result in nanometer-scale fluctuations in the local composition. The structure will be further complicated if collective and cooperative phenomena organize the solute distribution via longer range interactions between non-bonded solute sites. If the solute affects the phase stability then the question is raised of whether the atoms in domains with local compositions outside the limits of the bulk phase will rearrange into the structure stable for that composition and temperature or if the resulting stress would prevent such a local phase transition. If the former, then phase separated, heterogeneous structures at or below the diffraction limit will form. This nanometerscale competition between the phase transition and the epitaxial mismatch – exacerbated by the added strain if the transition involves a volume change – raises the potential for the formation of novel structures that do not occur in bulk material, e.g., fcc Fe. This coupling over multiple scales between inhomogeneity ordering, elastic forces, phase competition, and texture in the form of coexisting structures is a hallmark of martensites, a class of complex materials that includes ?-stabilized PuGa and that often exhibit correlated atomic and electronic properties. The enigmatic and extreme nature of Pu is consistent with its exhibiting unusual structural behavior of this type, including nanoscale heterogeneity in ?-stabilized PuGa and its enhanced homogeneity on aging that has been suggested based on earlier X-ray Absorption Fine Structure (XAFS) spectroscopy and x-ray pair distribution function (pdf) measurements. Measurements on a defined set of laboratory-prepared materials now corroborate and better describe this heterogeneity while additional aged samples demonstrate the role of heterogeneity in aging processes in Pu.

  5. Modeling of Selected Ceramic Processing Parameters Employed in the Fabrication of 238PuO2 Fuel Pellets

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Brockman, R. A. [Univ. of Dayton, OH (United States); Kramer, Daniel P. [Univ. of Dayton, OH (United States); Barklay, Chadwick D. [Univ. of Dayton, OH (United States); Cairns-Gallimore, Dirk [U.S. Department of Energy, Germantown, MD (United States); Brown, J. L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Huling, J. C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); van Pelt, C. E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2011-01-01T23:59:59.000Z

    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. Quantifying the importance of orbital over spin correlations in delta-Pu within density-functional theory

    SciTech Connect (OSTI)

    Soderlind, P; Wolfer, W

    2007-07-27T23:59:59.000Z

    Spin and orbital and electron correlations are known to be important when treating the high-temperature {delta} phase of plutonium within the framework of density-functional theory (DFT). One of the more successful attempts to model {delta}-Pu within this approach has included condensed-matter generalizations of Hund's three rules for atoms, i.e., spin polarization, orbital polarization, and spin-orbit coupling. Here they perform a quantitative analysis of these interactions relative rank for the bonding and electronic structure in {delta}-Pu within the DFT model. The result is somewhat surprising in that spin-orbit coupling and orbital polarization are far more important than spin polarization for a realistic description of {delta}-Pu. They show that these orbital correlations on their own, without any formation of magnetic spin moments, can account for the low atomic density of the {delta} phase with a reasonable equation-of-state. In addition, this unambiguously non-magnetic (NM) treatment produces a one-electron spectra with resonances close to the Fermi level consistent with experimental valence band photoemission spectra.

  7. Modeling of Selected Ceramic Processing Parameters Employed in the Fabrication of 238PuO2 Fuel Pellets

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

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

    2011-01-01T23:59:59.000Z

    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. Results of the modeling efforts will be discussed.« less

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

    SciTech Connect (OSTI)

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

    2010-02-18T23:59:59.000Z

    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.

  9. Disposition of transuranic residues from plutonium isentropic compression experiment (Pu-ice) conducted at Z machine

    SciTech Connect (OSTI)

    Goyal, Kapil K [Los Alamos National Laboratory; French, David M [Los Alamos National Laboratory; Humphrey, Betty J [WESTON SOLUTIONS INC.; Gluth, Jeffry [SNL

    2010-01-01T23:59:59.000Z

    In 1992, the U.S. Congress passed legislation to discontinue above- and below-ground testing of nuclear weapons. Because of this, the U.S. Department of Energy (DOE) must rely on laboratory experiments and computer-based calculations to verify the reliability of the nation's nuclear stockpile. The Sandia National Laboratories/New Mexico (SNL/NM) Z machine was developed by the DOE to support its science-based approach to stockpile stewardship. SNL/NM researchers also use the Z machine to test radiation effects on various materials in experiments designed to mimic nuclear explosions. Numerous components, parts, and materials have been tested. These experiments use a variety of radionuclides; however, plutonium (Pu) isotopes with greater than ninety-eight percent enrichment are the primary radionuclides used in the experiments designed for stockpile stewardship. In May 2006, SNL/NM received authority that the Z Machine Isentropic Compression Experiments could commence. Los Alamos National Laboratory (LANL) provided the plutonium targets and loaded the target assemblies, which were fabricated by SNL/NM. LANL shipped the loaded assemblies to SNL/NM for Z machine experiments. Three experiments were conducted from May through July 2006. The residues from each experiment, which weighed up to 913 pounds, were metallic and packaged into a respective 55-gallon drum each. Based on a memorandum of understanding between the two laboratories, LANL provides the plutonium samples and the respective radio-isotopic information. SNL/NM conducts the experiments and provides temporary storage for the drums until shipment to LANL for final waste certification for disposal at the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. This paper presents a comprehensive approach for documenting generator knowledge for characterization of waste in cooperation with scientists at the two laboratories and addresses a variety of topics such as material control and accountability, safeguards of material, termination of safeguards for eventual shipment from SNL/NM to LANL, associated approvals from DOE-Carlsbad Field Office, which governs WIPP and various notifications. It portrays a comprehensive approach needed for successful completion of a complex project between two national laboratories.

  10. Interaction of Pu(IV,VI) hydroxides/oxides with metal hydroxides/oxides in alkaline media

    SciTech Connect (OSTI)

    Fedoseev, A.M.; Krot, N.N.; Budantseva, N.A.; Bessonov, A.A.; Nikonov, M.V.; Grigoriev, M.S.; Garnov, A.Y.; Perminov, V.P.; Astafurova, L.N. [Russian Academy of Sciences, Moscow (Russian Federation). Inst. of Physical Chemistry

    1998-08-01T23:59:59.000Z

    The primary goal of this investigation was to obtain data on the possibility, extent, and characteristics of interaction of Pu(IV) and (VI) with hydroxides and oxides of d-elements and other metals [Al(III), LA(III), and U(VI)] in alkaline media. Such information is important in fundamental understanding of plutonium disposition and behavior in Hanford Site radioactive tank waste sludge. These results supply essential data for determining criticality safety and in understanding transuranic waste behavior in storage, retrieval, and treatment of Hanford Site tank waste.

  11. Effect of Fe2+ Oxidation on the Removal of 238Pu from Neptunium Solution by Anion Exchange

    SciTech Connect (OSTI)

    KYSER, EDWARD

    2004-06-01T23:59:59.000Z

    The effect of ferrous sulfamate (FS) oxidation and variation in nitric acid concentration on the removal of {sup 238}Pu contamination from Np by the HB-Line anion exchange flowsheet has been tested. Significant rejection of {sup 238}Pu was observed by washing with a reductive wash solution containing 6.0 to 6.8 M nitric acid (HNO{sub 3}) with as little as 30% of the Fe{sup 2+} from the FS remaining in its reduced form. To achieve the desired 30% removal of {sup 238}Pu from the process, conditions should be controlled to maintain the Fe{sup 2+}/Fe{sup 3+} ratio in the reductive wash to be greater than 60%/40% (or 1.5). Since Fe{sup 2+} oxidation is strongly affected by temperature and nitric acid concentration, these parameters (as well as time after FS addition) need to be controlled to ensure predictable results. A shortened-height column was utilized in these tests to match changes in the plant equipment. Lab experiments scaled to plant batch sizes of 2000 g Np were observed with modest losses for ''up-flow'' washing. The following are recommended conditions for removing {sup 238}Pu from Np solutions by anion exchange in HB-Line: (1) Feed conditions: ''Up-flow'' 6.4-8.0 M HNO{sub 3}, 0.02 M hydrazine (N{sub 2}H{sub 4}), 0.05 M excess FS. (2) Reductive Wash conditions: ''Up-flow'' 6 Bed volumes (BV) of 6.4 M HNO{sub 3}, 0.05 M FS (minimum 0.03M Fe{sup 2+} during wash cycle), 0.05 M hydrazine, less than 1.8 mL/min/cm{sup 2} flowrate. (3) Decontamination Wash conditions: ''Up-flow'' 1-2 BV of 6.4-8.0 M HNO{sub 3}, no FS, no hydrazine, less than 1.8 mL/min/cm{sup 2} flowrate. (4) Elution conditions: ''Down-flow'' 0.17 M HNO{sub 3}, 0.05 M hydrazine, no FS.

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

    SciTech Connect (OSTI)

    Pinder, J.E. III; McLeod, K.W.; Adriano, D.C.; Corey, J.C.; Boni, A.L. (Savannah River Ecology Laboratory, Aiken, SC (USA))

    1990-12-01T23:59:59.000Z

    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.

  13. Glenwood Springs technical conference proceedings. Volume II. Bibliography of publications, state coupled geothermal resource assessment program

    SciTech Connect (OSTI)

    Ruscetta, C.A.; Foley, D. (eds.)

    1981-05-01T23:59:59.000Z

    The bibliography of publications is divided by state as follows: Alaska, California, Colorado, Hawaii, Montana, Nebraska, New Mexico, New York, North Dakota, Oregon, Texas and Washington. (MHR)

  14. Glenwood Hot Springs Hotel Pool & Spa Low Temperature Geothermal Facility |

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdfGetec AG Contracting Jump to:Echo, Maryland:Glenwillow, Ohio: Energy

  15. Glenwood Springs Vapor Caves Pool & Spa Low Temperature Geothermal Facility

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdfGetec AG Contracting Jump to:Echo, Maryland:Glenwillow, Ohio:| Open Energy

  16. Photochemical removal of NpF sub 6 and PuF sub 6 from UF sub 6 gas streams

    SciTech Connect (OSTI)

    Beitz, J.V.; Williams, C.W.

    1990-01-01T23:59:59.000Z

    A novel photochemical method of removing reactive fluorides from UF{sub 6} gas has been discovered. This method reduces generated waste to little more than the volume of the removed impurities, minimizes loss of UF{sub 6}, and can produce a recyclable by-product, fluorine gas. In our new method, impure UF{sub 6}, is exposed to ultraviolet light which dissociates the UF{sub 6} to UF{sub 5} and fluorine atom. Impurities which chemically react with UF{sub 5} are reduced and form solid compounds easily removed from the gas while UF{sub 5} is converted back to UF{sub 6}. Proof-of-concept testing involved UF{sub 6} containing NpF{sub 6} and PuF{sub 6} with CO added as a fluorine atom scavenger. In a single photolysis step, greater than 5000-fold reduction of PuF{sub 6} was demonstrated while reducing NpF{sub 6} by more than 40-fold. This process is likely to remove corrosion and fission product fluorides that are more reactive than UF{sub 6} and has been demonstrated without an added fluorine atom scavenger by periodically removing photogenerated fluorine gas. 44 refs., 3 figs., 2 tabs.

  17. Isothermal Martensitic and Pressure-Induced Delta to Alpha-Prime Phase Transformations in a Pu-Ga Alloy

    SciTech Connect (OSTI)

    Schwartz, A J; Wall, M A; Farber, D L; Moore, K T; Blobaum, K M

    2008-01-18T23:59:59.000Z

    A well-homogenized Pu-2 at.% Ga alloy can be retained in the metastable face-centered cubic {delta} phase at room temperature. Ultimately, this metastable {delta} phase will decompose via a eutectoid transformation to the thermodynamically stable monoclinic {alpha} phase and the intermetallic compound Pu{sub 3}Ga over a period of approximately 10,000 years [1]. In addition, these low solute-containing {delta}-phase Pu alloys are metastable with respect to an isothermal martensitic phase transformation to the {alpha}{prime} phase during low temperature excursions [2, 3] and are also metastable with respect to a {delta} {yields} {alpha}{prime} phase transformation with increases in pressure [3-5]. The low temperature {delta} {yields} {alpha}{prime} isothermal martensitic phase transformation in the Pu-2 at.% Ga alloy only goes to {approx}25% completion with the resultant {approx}20 {micro}m long by 2 {micro}m wide lath-shaped {alpha}{prime} particles dispersed within the {delta} matrix. In recently reported studies, Faure et al. [4] have observed a {delta} {yields} {gamma} {yields} {alpha}{prime} pressure-induced phase transformation sequence during a diamond anvil cell investigation and, based on x-ray diffraction and density and compressibility experiments, Harbur [5] has concluded that both {alpha}{prime} and an amorphous phase are present in samples that were pressurized and recovered. In this work, a large volume moissanite anvil cell is constructed to permit the pressurization and recovery of specimens of a size suitable for TEM and electron diffraction studies. The cell, shown in Fig. 1, has an overall diameter of 101.6 mm, a moissanite anvil diameter of 9.00 mm, a culet size of 3 mm, and a spring steel gasket 0.5 mm thick with a hole diameter of 2.5 mm. A 2.3 mm diameter by 100 {micro}m thick sample of {delta}-phase Pu-2 at.% Ga is compressed at a rate of approximately 0.05 GPa/minute to {approx}1 GPa to induce the phase transformation to {alpha}{prime}. Optical microscopy of the recovered specimen reveals a very fine microstructure that appears to be single phase, although the resolution of this technique is insufficient to differentiate between single and multiple phases if the grain size is below approximately 1 {micro}m. X-ray diffraction, using a laboratory Cu K{sub {alpha}} source with wavelength of 1.542{angstrom}, shows the monoclinic reflections from the {alpha}{prime} phase, strong peaks from the aluminum specimen holder, and weak peaks from the face-centered cubic {delta} phase as shown in Fig. 2. The recovered specimen is prepared for TEM and electron diffraction studies as described in Moore et al. [6]. TEM reveals small regions of {delta} phase with a very high dislocation density interspersed between the 10-100's nm {alpha}{prime} grains as shown in Fig. 3. Electron diffraction, shown in the insert in Fig. 3, clearly reveals the presence of the {delta} phase. This microstructure is in contrast to the {alpha}{prime} particles that form as a result of the low-temperature isothermal martensite in which the {alpha}{prime} particles are lath-shaped and significantly larger as shown in the optical micrograph in Fig. 4 of a sample cooled to -120 C and held for 10 hours. In these preliminary results, there is no evidence of either an amorphous phase, as suggested by Harbur [5], or the presence of a {gamma} phase. We expected to observe an amorphous phase based on the similarity of this experiment to that of Harbur [5]. It is possible that the {gamma} phase, as reported by Faure et al. [4], does form as an intermediate, but it is not retained to ambient pressure.

  18. Analysis of Pu-Only Partitioning Strategies in LMFBR Fuel Cycles

    SciTech Connect (OSTI)

    Samuel Bays; Gilles Youinou

    2013-02-01T23:59:59.000Z

    Sodium cooled Fast Reactors (SFR) have been under consideration for production of electricity, fissile material production, and for destruction of transuranics for decades. The neutron economy of a SFR can be operated in one of two ways. One possibility is to operate the reactor in a transuranic burner mode which has been the focus of active R&D in the last 15 years. However, prior to that the focus was on breeding transuranics. This later mode of managing the neutron economy relies on ensuring the maximum fuel utilization possible in such a way as to maximize the amount of plutonium produced per unit of fission energy in the reactor core. The goal of maximizing plutonium production in this study is as fissile feed stock for the production of MOX fuel to be used in Light Water Reactors (LWR). Throughout the l970’s, this fuel cycle scenario was the focus of much research by the Atomic Energy Commission in the event that uranium supplies would be scarce. To date, there has been sufficient uranium to supply the once through nuclear fuel cycle. However, interest in a synergistic relationship Liquid Metal Fast Breeder Reactors (LMFBR) and a consumer LWR fleet persists, prompting this study. This study considered LMFBR concepts with varying additions of axial and radial reflectors. Three scenarios were considered in collaboration with a companion study on the LWR-MOX designs based on plutonium nuclide vectors produced by this study. The first scenario is a LMFBR providing fissile material to make MOX fuel where the MOX part of the fuel cycle is operated in a once-through-then-out mode. The second scenario is the same as the first but with the MOX part of the fuel cycle multi-recycling its own plutonium with LMFBR being used for the make-up feed. In these first two scenarios, plutonium partitioning from the minor actinides (MA) was assumed. Also, the plutonium management strategy of the LMFBR ensured that only the high fissile purity plutonium bred from blankets was sold to the MOX LWRs. The third scenario considered a LMFBR fuel cycle in an expansionary mode where excess bred transuranic material is accumulated for spinning off additional LMFBR cores. In this latter scenario, no plutonium partitioning was considered. After every cycle, transuranic from both driver and blankets is sold to the MOX LWRs. The MA production from LMFBR operated in a Pu-only fuel cycle is roughly only 1% that of the transuranic production rate. This is in contrast to LWR fuel cycles where the MA content in TRU is closer to 10% or more. If such a LMFBR were operated to provide fissile material to a fleet of MOX reactors, then 1 GWe of LMFBR could support between approximately 0.11 and 0.43 GWe of LWR-MOX reactors for a LMFBR conversion ratio between 1.1 and 1.5, if the MOX reactors were operated in a once-through-then out mode. If the plutonium is continuously recycled in the MOX reactors then the support ratio is approximately 1 GWe of LMFBR for between 0.13 and 0.65 GWe of LWR-MOX reactors depending on the LMFBR conversion ratio. Also, it was found that if the LMFBR fleet were operated in a purely expansionary mode, the smallest doubling time achievable would be seven years.

  19. Near-infrared photoluminescence and ligand K-edge x-ray absorption spectroscopies of AnO2Cl42-(An:u, NP, Pu)

    SciTech Connect (OSTI)

    Wilkerson, Marianne P [Los Alamos National Laboratory; Berg, John M [Los Alamos National Laboratory; Clark, David L [Los Alamos National Laboratory; Conradson, Steven D [Los Alamos National Laboratory; Hobart, David E [Los Alamos National Laboratory; Kozimor, Stosh A [Los Alamos National Laboratory; Scott, Brian L [Los Alamos National Laboratory

    2008-01-01T23:59:59.000Z

    We have used photoluminescence and X-ray absorption spectroscopies to investigate electronic structures and metal-ligand bonding of a series of An02CI/ ' (An = U, Np, Pu) compounds. Specifically, we will discuss time-resolved near-infrared emission spectra of crystalline Cs2U(An)02C14 (An = Np and Pu) both at 23 K and 75 K, as well as chlorine Kedge X-ray absorption spectra ofCs2An02CI4 (An = U, Np).

  20. Hybridization and superconducting gaps in heavy-fermion superconductor PuCoGa5 probed via the dynamics of photoinduced quasiparticles

    SciTech Connect (OSTI)

    Talbayev, Diyar [Los Alamos National Laboratory; Trugman, Stuart A [Los Alamos National Laboratory; Zhu, Jian - Xin [Los Alamos National Laboratory; Bauer, Eric D [Los Alamos National Laboratory; Kennison, John A [Los Alamos National Laboratory; Mitchell, Jeremy N [Los Alamos National Laboratory; Thompson, Joe D [Los Alamos National Laboratory; Sarrao, John L [Los Alamos National Laboratory; Taylor, Antoinette J [Los Alamos National Laboratory; Burch, Kenneth S [CANADA; Chia, Elbert E. M. [CANADA

    2009-01-01T23:59:59.000Z

    We have examined the relaxation of photoinduced quasiparticles in the heavy-fermion superconductor PuCoGa{sub 5}. The deduced electron-phonon coupling constant is incompatible with the measured superconducting transition temperature T{sub c}, which speaks against phonon-mediated superconducting pairing. Upon lowering the temperature, we observe an order-of-magnitude increase of the quasiparticle relaxation time in agreement with the phonon bottleneck scenario - evidence for the presence of a hybridization gap in the electronic density of states. The modification of photoinduced reflectance in the superconducting state is consistent with the heavy character of the quasiparticles that participate in Cooper pairing. The discovery of relatively high-temperature superconductivity in the Pu-based compounds PuCoGa{sub 5} (T{sub c} = 18.5 K) and PuRhGa{sub 5} (T{sub c} = 8.7 K) has renewed the interest in actinide materials research. The Pu-based superconductors share the HoCoGa{sub 5}-type tetragonal lattice stucture with the Ce-based series of compounds (CeRhIn{sub 5}, CeCoIn{sub 5}, and CeIrIn{sub 5}) commonly referred to as '115' materials. In the Ce-based 115 compounds, CeIrIn{sub 5} (T{sub c} = 0.4 K) and CeCoIn{sub 5} (T{sub c} = 2.3 K), display superconductivity at ambient pressure. Both Ce- and Pu-based 115 compounds display the heavy fermion behavior resulting from the influence of 4f (Ce) and 5f (Pu) electrons. The most intriguing question concerns the origin of superconductivity (SC) in the 115 materials. In the Ce series, the d-wave symmetry of the SC order parameter and the proximity of SC order to magnetism have lead to a widespread belief that the unconventional SC is induced by antiferromagnetic spin fluctuations. In the Pu compounds, two possible scenarios regarding the SC mechanism have been considered: one approach favors a magnetically mediated unconventional SC similar to that in CeCoIn{sub 5}. In the other scenario, the conventional SC is mediated by phonons, where the strength of the electron-phonon (e-ph) coupling {lambda} is the crucial parameter that sets the superconducting transition temperature T{sub c}. In this Letter, we present a measurement of the e-ph coupling constant {lambda} via the pump-probe optical study of the room-temperature relaxation time of photoinduced reflectance. We find that e-ph coupling ({lambda} = 0.2-0.26) is too weak to explain the high T{sub c} of PuCoGa{sub 5} and that phonon-mediated superconductivity is unlikely in this material. Upon lowering the temperature in the normal state (T > T{sub c}), we find an order-of-magnitude increase in the relaxation time consistent with a phonon bottleneck, similar to other heavy-fermion materials, which provides the first optical evidence of the presence of a hybridization gap in the electronic density of states (DOS). Below T{sub c}, the photoinduced response exhibits dramatic changes that we ascribe to the opening of the superconducting (SC) gap at the Fermi level. The observed dynamics confirms that the same quasiparticles detected in the normal state, i.e., the heavy quasiparticles, also participate in the SC pairing. Our study is the first to directly probe the electronic structure of PuCoGa{sub 5} in the SC state and corroborate that fact. Our results are consistent with the theoretical investigations, which find that the electronic structure is dominated by cylindrical sheets of Fermi surfaces with large 5f electron character, suggesting that the delocalized 5f electrons of Pu playa key role in the superconducting pairing.

  1. Relativistic energy density functionals: Low-energy collective states of {sup 240}Pu and {sup 166}Er

    SciTech Connect (OSTI)

    Li, Z. P. [Physics Department, Faculty of Science, University of Zagreb, 10000 Zagreb (Croatia); State Key Laboratory of Nuclear Physics and Technology, School of Physics, Peking University, Beijing 100871 (China); School of Physical Science and Technology, Southwest University, Chongqing 400715 (China); Niksic, T.; Vretenar, D. [Physics Department, Faculty of Science, University of Zagreb, 10000 Zagreb (Croatia); Ring, P. [State Key Laboratory of Nuclear Physics and Technology, School of Physics, Peking University, Beijing 100871 (China); Physik-Department der Technischen Universitaet Muenchen, D-85748 Garching (Germany); Meng, J. [State Key Laboratory of Nuclear Physics and Technology, School of Physics, Peking University, Beijing 100871 (China); School of Physics and Nuclear Energy Engineering, Beihang University, Beijing 100191 (China)

    2010-06-15T23:59:59.000Z

    The empirical relativistic density-dependent, point-coupling energy density functional, adjusted exclusively to experimental binding energies of a large set of deformed nuclei with Aapprox =150-180 and Aapprox =230-250, is tested with spectroscopic data for {sup 166}Er and {sup 240}Pu. Starting from constrained self-consistent triaxial relativistic Hartree-Bogoliubov calculations of binding energy maps as functions of the quadrupole deformation in the beta-gamma plane, excitation spectra and E2 transition probabilities are calculated as solutions of the corresponding microscopic collective Hamiltonian in five dimensions for quadrupole vibrational and rotational degrees of freedom and compared with available data on low-energy collective states.

  2. Effects of self-irradiation on local crystal structure and 5flocalization in PuCoGa5

    SciTech Connect (OSTI)

    Booth, C.H.; Daniel, M.; Wilson, R.E.; Bauer, E.D.; Mitchell,J.N.; Moreno, N.O.; Morales, L.A.; Sarrao, J.L.; Allen, P.G.

    2006-10-20T23:59:59.000Z

    The 18.5 K superconductor PuCoGa{sub 5} has many unusual properties, including those due to damage induced by self-irradiation. The superconducting transition temperature decreases sharply with time, suggesting a radiation-induced Frenkel defect concentration much larger than predicted by current radiation damage theories. Extended x-ray absorption fine-structure measurements demonstrate that while the local crystal structure in fresh material is well ordered, aged material is disordered much more strongly than expected from simple defects, consistent with strong disorder throughout the damage cascade region. These data highlight the potential impact of local lattice distortions relative to defects on the properties of irradiated materials and underscore the need for more atomic-resolution structural comparisons between radiation damage experiments and theory.

  3. Production of new superheavy Z=108-114 nuclei with $^{238}$U, $^{244}$Pu and $^{248,250}$Cm targets

    E-Print Network [OSTI]

    Feng, Zhao-Qing; Li, Jun-Qing

    2009-01-01T23:59:59.000Z

    Within the framework of the dinuclear system (DNS) model, production cross sections of new superheavy nuclei with charged numbers Z=108-114 are analyzed systematically. Possible combinations based on the actinide nuclides $^{238}$U, $^{244}$Pu and $^{248,250}$Cm with the optimal excitation energies and evaporation channels are pointed out to synthesize new isotopes which lie between the nuclides produced in the cold fusion and the $^{48}$Ca induced fusion reactions experimentally, which are feasible to be constructed experimentally. It is found that the production cross sections of superheavy nuclei decrease drastically with the charged numbers of compound nuclei. Larger mass asymmetries of the entrance channels enhance the cross sections in 2n-5n channels.

  4. Production of new superheavy Z=108-114 nuclei with $^{238}$U, $^{244}$Pu and $^{248,250}$Cm targets

    E-Print Network [OSTI]

    Zhao-Qing Feng; Gen-Ming Jin; Jun-Qing Li

    2009-12-21T23:59:59.000Z

    Within the framework of the dinuclear system (DNS) model, production cross sections of new superheavy nuclei with charged numbers Z=108-114 are analyzed systematically. Possible combinations based on the actinide nuclides $^{238}$U, $^{244}$Pu and $^{248,250}$Cm with the optimal excitation energies and evaporation channels are pointed out to synthesize new isotopes which lie between the nuclides produced in the cold fusion and the $^{48}$Ca induced fusion reactions experimentally, which are feasible to be constructed experimentally. It is found that the production cross sections of superheavy nuclei decrease drastically with the charged numbers of compound nuclei. Larger mass asymmetries of the entrance channels enhance the cross sections in 2n-5n channels.

  5. CONTAINMENT VESSEL TEMPERATURE FOR PU-238 HEAT SOURCE CONTAINER UNDER AMBIENT, FREE CONVECTION AND LOW EMISSIVITY COOLING CONDITIONS

    SciTech Connect (OSTI)

    Gupta, N.; Smith, A.

    2011-02-14T23:59:59.000Z

    The EP-61 primary containment vessel of the 5320 shipping package has been used for storage and transportation of Pu-238 plutonium oxide heat source material. For storage, the material in its convenience canister called EP-60 is placed in the EP-61 and sealed by two threaded caps with elastomer O-ring seals. When the package is shipped, the outer cap is seal welded to the body. While stored, the EP-61s are placed in a cooling water bath. In preparation for welding, several containers are removed from storage and staged to the welding booth. The significant heat generation of the contents, and resulting rapid rise in component temperature necessitates special handling practices. The test described here was performed to determine the temperature rise with time and peak temperature attained for an EP-61 with 203 watts of internal heat generation, upon its removal from the cooling water bath.

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

    SciTech Connect (OSTI)

    Kato, M.; Murakami, T.; Sunaoshi, T. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency, Muramatsu Tokai-mura Ibaraki, 319-1194 (Japan); Nelson, A.T.; McClellan, K.J. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

    2013-07-01T23:59:59.000Z

    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)

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

    SciTech Connect (OSTI)

    Fukasawa, T.; Hoshino, K. [Hitachi-GE Nuclear Energy, Ltd, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Takano, M. [Japan Atomic Energy Agency, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Sato, S. [Hokkaido University, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Shimazu, Y. [Fukui University, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan)

    2013-07-01T23:59:59.000Z

    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)

  8. Test plan for non-radioactive testing of vertical calciner for development of direct denitration conversion of Pu-bearing liquors to stable, storage solids

    SciTech Connect (OSTI)

    Fisher, F.D.

    1995-03-30T23:59:59.000Z

    Plutonium-bearing liquors, including ANL scrap liquors, will be used for development and demonstration of a vertical calciner direct denitration process for conversion of those liquors to stable, storable PuO{sub 2}-rich solids. This test plan is to test with non-radioactive stand-in materials to demonstrate adequate performance of the vertical calciner and ancillary equipment.

  9. CALORIMETER-BASED ADJUSTMENT OF MULTIPLICITY DETERMINED 240PU EFF KNOWN-A ANALYSIS FOR THE ASSAY OF PLUTONIUM

    SciTech Connect (OSTI)

    Dubose, F.

    2012-02-21T23:59:59.000Z

    In nuclear material processing facilities, it is often necessary to balance the competing demands of accuracy and throughput. While passive neutron multiplicity counting is the preferred method for relatively fast assays of plutonium, the presence of low-Z impurities (fluorine, beryllium, etc.) rapidly erodes the assay precision of passive neutron counting techniques, frequently resulting in unacceptably large total measurement uncertainties. Conversely, while calorimeters are immune to these impurity effects, the long count times required for high accuracy can be a hindrance to efficiency. The higher uncertainties in passive neutron measurements of impure material are driven by the resulting large (>>2) {alpha}-values, defined as the ({alpha},n):spontaneous fission neutron emission ratio. To counter impurity impacts for high-{alpha} materials, a known-{alpha} approach may be adopted. In this method, {alpha} is determined for a single item using a combination of gamma-ray and calorimetric measurements. Because calorimetry is based on heat output, rather than a statistical distribution of emitted neutrons, an {alpha}-value determined in this way is far more accurate than one determined from passive neutron counts. This fixed {alpha} value can be used in conventional multiplicity analysis for any plutonium-bearing item having the same chemical composition and isotopic distribution as the original. With the results of single calorimeter/passive neutron/gamma-ray measurement, these subsequent items can then be assayed with high precision and accuracy in a relatively short time, despite the presence of impurities. A calorimeter-based known-{alpha} multiplicity analysis technique is especially useful when requiring rapid, high accuracy, high precision measurements of multiple plutonium bearing items having a common source. The technique has therefore found numerous applications at the Savannah River Site. In each case, a plutonium (or mixed U/Pu) bearing item is divided into multiple containers. A single item from that batch is then selected for both neutron and calorimetric measurements; all remaining items undergo a neutron measurement only. Using the technique mentioned above, the 'true' {alpha} value determined from the first (calorimeter and passive neutron measured) item is used in multiplicity analysis for all other items in the batch. The justification for using this {alpha} value in subsequent calculations is the assumption that the chemical composition and isotopic distribution of all batch items are the same, giving a constant ({alpha},n):spontaneous fission ratio. This analysis method has been successfully applied to the KIS Facility, significantly improving measurement uncertainties and reducing processing times for numerous items. Comprehensive plans were later developed to extend the use of this method to other applications, including the K-Area Shuffler and the H-Area Pu-Blending Project. While only the feasibility study for the Shuffler has been completed, implementation of the method in the H-Area Pu-Blending Project is currently in progress and has been successfully applied to multiple items. This report serves to document the details of this method in order to serve as a reference for future applications. Also contained herein are specific examples of the application of known-{alpha} multiplicity analysis.

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

    SciTech Connect (OSTI)

    Not Available

    1994-01-15T23:59:59.000Z

    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.

  11. An Interpretation of Energy Dependence of Delayed Neutron Yields in the Resonance Region for {sup 235}U and {sup 239}Pu

    SciTech Connect (OSTI)

    Ohsawa, Takaaki [Kinki University (Japan); Hambsch, Franz-Josef [EC-JRC0, Institute for Reference Materials and Measurements (Belgium)

    2004-09-15T23:59:59.000Z

    Possible fluctuation in the delayed neutron yields (DNYs) in the resonance region was predicted on the basis of experimental data of mass distribution of fission fragments at resonances. Analyzed according to the multimodal random neck rupture model of fission, the small variations in the experimental mass distribution were attributed to fluctuations in the branching ratios to different modes of fission. Using the results of analysis of measured data for {sup 235}U and {sup 239}Pu, the DNYs were calculated for each resonance by the summation method, considering 271 precursors and evaluated data of delayed neutron emission probability. It was found that the DNYs should have local dips for {sup 235}U and spikes for {sup 239}Pu at resonances.

  12. The CIELO Collaboration:Neutron Reactions on 1H, 16O, 56Fe, 235,238U, and 239Pu

    SciTech Connect (OSTI)

    Giuseppe Palmiotti; M. B. Chadwick

    2014-04-01T23:59:59.000Z

    CIELO (Collaborative International Evaluated Library Organization) provides a new working paradigm to facilitate evaluated nuclear reaction data advances. It brings together experts from across the international nuclear reaction data community to identify and document discrepancies among existing evaluated data libraries, measured data, and model calculation interpretations, and aims to make progress in reconciling these discrepancies to create more accurate ENDF-formatted files. The focus will initially be on a small number of the highest-priority isotopes, namely 1H, 16O, 56Fe, 235,238U, and 239Pu. This paper identifies discrepancies between various evaluations of the highest priority isotopes, and was commissioned by the OECD's Nuclear Energy Agency WPEC (Working Party on International Nuclear Data Evaluation Co-operation) during a meeting held in May 2012. The evaluated data for these materials in the existing nuclear data libraries — ENDF/B-VII.1, JEFF-3.1, JENDL-4.0, CENDL-3.1, ROSFOND, IRDFF 1.0 — are reviewed, discrepancies are identified, and some integral properties are given. The paper summarizes a program of nuclear science and computational work needed to create the new CIELO nuclear data evaluations.

  13. International Workshop on Gamma Spectrometry Analysis Codes for U and Pu Isotopics: Workshop Results and Next Steps

    SciTech Connect (OSTI)

    McGinnis, Brent R [ORNL; Solodov, Alexander A [ORNL; Shipwash, Jacqueline L [ORNL; Zhernosek, Alena V [ORNL; McKinney, Teressa L [ORNL; Pickett, Chris A [ORNL; Peerani, Paolo [ORNL

    2009-01-01T23:59:59.000Z

    In November 2008, the Institute of Nuclear Materials Management (INMM) and the European Safeguards Research and Development Association (ESARDA) co-hosted the International Workshop on Gamma Spectrometry Analysis Codes for U and Pu Isotopics at the Oak Ridge National Laboratory (ORNL). This workshop was conducted in response to needs expressed by the international safeguards community to understand better the capabilities and limitations of the codes; to ensure these codes are sustained; and to ensure updates or revisions are performed in a controlled manner. The workshop was attended by approximately 100 participants. The participants included code developers, code suppliers, safeguards specialists, domestic and international inspectors, process operators, regulators, and programme sponsors from various government agencies. The workshop provided a unique opportunity for code developers, commercial distributors and end users to interact in a hands-on laboratory environment to develop solutions for programmatic and technical issues associated with the various codes. The workshop also provided an international forum for discussing development of an internationally accepted standard test method. This paper discusses the organization of the workshop, its goals and objectives and feedback received from the participants. The paper also describes the significance of the working group's contribution to improving codes that are commonly used during inspections to verify that nuclear facilities are compliant with treaty obligations that ensure nuclear fuel cycle facilities are used for peaceful purposes.

  14. 8-group relative delayed neutron yields for epithermal neutron induced fission of sup 2 sup 3 sup 5 U and sup 2 sup 3 sup 9 Pu

    E-Print Network [OSTI]

    Piksaikin, V M; Kazakov, L E; Korolev, G G; Roshchenko, V A; Tertychnyj, R G

    2001-01-01T23:59:59.000Z

    An 8-group representation of relative delayed neutron yields was obtained for epithermal neutron induced fission of sup 2 sup 3 sup 5 U and sup 2 sup 3 sup 9 Pu. These data were compared with ENDF/B-VI data in terms of the average half- life of the delayed neutron precursors and on the basis of the dependence of reactivity on the asymptotic period.

  15. Single cell analyses of regulatory network perturbations using enhancer targeting TAL Effectors suggest novel roles for PU.1 during haematopoietic specification

    E-Print Network [OSTI]

    Wilkinson, Adam C.; Kawata, Viviane K. S.; Schütte, Judith; Gao, Xuefei; Antoniou, Stella; Baumann, Claudia; Woodhouse, Steven; Hannah, Rebecca; Tanaka, Yosuke; Swiers, Gemma; Moignard, Victoria; Fisher, Jasmin; Hidetoshi, Shimauchi; Tijssen, Marloes R.; de Bruijn, Marella F. T. R.; Liu, Pentao; Göttgens, Berthold

    2014-01-01T23:59:59.000Z

    and Scl+40kb are believed to mediate positive autoregulation through binding of 26 PU.1 and Scl proteins, respectively (Okuno et al., 2005; Ogilvy et al., 2007). 27 28 5 Recent technological advances in microfluidic technology have led... microfluidics-based single cell gene expression highlighted the potential for heterogeneity of 7 knockdown efficiency within single cells following siRNA-mediated gene silencing (Toriello 8 et al., 2008). However, the ability to accurately assess gene...

  16. Biotic and Abiotic Reduction and Solubilization of Pu(IV)O2•xH2O(am) as Affected by Anthraquinone-2,6-disulfonate (AQDS) and Ethylenediaminetetraacetate (EDTA)

    SciTech Connect (OSTI)

    Plymale, Andrew E.; Bailey, Vanessa L.; Fredrickson, Jim K.; Heald, Steve M.; Buck, Edgar C.; Shi, Liang; Wang, Zheming; Resch, Charles T.; Moore, Dean A.; Bolton, Harvey

    2012-01-24T23:59:59.000Z

    In the presence of hydrogen (H{sub 2}), the synthetic chelating agent ethylenediaminetetraacetate (EDTA), and the electron shuttle anthraquinone-2,6-disulfonate (AQDS), the dissimilatory metal-reducing bacteria (DMRB) Shewanella oneidensis and Geobacter sulfurreducens both reductively solubilized 100% of added 0.5 mM plutonium (IV) hydrous oxide (Pu(IV)O{sub 2} {lg_bullet} xH{sub 2}O{sub (am)}) in {approx}24 h at pH 7 in a non-complexing buffer. In the absence of AQDS, bioreduction was much slower ({approx}22 days) and less extensive ({approx}83-94%). In the absence of DMRB but under comparable conditions, 89% (without AQDS) to 98% (with AQDS) of added 0.5 mM PuO{sub 2} {lg_bullet} xH{sub 2}O{sub (am)} was reductively solubilized over 418 days. Under comparable conditions but in the absence of EDTA, <0.001% of the 0.5 mM PuO{sub 2} {lg_bullet} xH{sub 2}O{sub (am)} was solubilized, with or without bacteria. However, Pu(aq) increased by as much as an order of magnitude in some EDTA-free treatments, both biotic and abiotic, and increases in solubility were associated with the production of both Pu(OH)3(am) and Pu(III)(aq). Incubation with DMRB in the absence of EDTA increased the polymeric and crystalline content of the PuO{sub 2} {lg_bullet} xH{sub 2}O{sub (am)} and also decreased Pu solubility in 6-N HCl. Results from an in vitro assay demonstrated electron transfer to PuO{sub 2} {lg_bullet} xH{sub 2}O{sub (am)} from the S. oneidensis outer-membrane c-type cytochrome MtrC, and EDTA increased the oxidation of MtrC by PuO{sub 2} {lg_bullet} xH{sub 2}O{sub (am)}. Our results suggest that PuO{sub 2} {lg_bullet} xH{sub 2}O{sub (am)} biotic and abiotic reduction and solubilization may be important in anoxic, reducing environments, especially where complexing ligands and electron shuttling compounds are present.

  17. Heavy element radionuclides (Pu, Np, U) and {sup 137}Cs in soils collected from the Idaho National Engineering and Environmental Laboratory and other sites in Idaho, Montana, and Wyoming

    SciTech Connect (OSTI)

    Beasley, T.M.; Rivera, W. Jr. [Dept. of Energy, New York, NY (United States). Environmental Measurements Lab.; Kelley, J.M.; Bond, L.A. [Pacific Northwest National Lab., Richland, WA (United States); Liszewski, M.J. [Bureau of Reclamation (United States); Orlandini, K.A. [Argonne National Lab., IL (United States)

    1998-10-01T23:59:59.000Z

    The isotopic composition of Pu in soils on and near the Idaho National Engineering and Environmental Laboratory (INEEL) has been determined in order to apportion the sources of the Pu into those derived from stratospheric fallout, regional fallout from the Nevada Test Site (NTS), and facilities on the INEEL site. Soils collected offsite in Idaho, Montana, and Wyoming were collected to further characterize NTS fallout in the region. In addition, measurements of {sup 237}Np and {sup 137}Cs were used to further identify the source of the Pu from airborne emissions at the Idaho Chemical Processing Plant (ICPP) or fugitive releases from the Subsurface Disposal Area (SDA) in the Radioactive Waste Management Complex (RWMC). There is convincing evidence from this study that {sup 241}Am, in excess of that expected from weapons-grade Pu, constituted a part of the buried waste at the SDA that has subsequently been released to the environment. Measurements of {sup 236}U in waters from the Snake River Plain aquifer and a soil core near the ICPP suggest that this radionuclide may be a unique interrogator of airborne releases from the ICPP. Neptunium-237 and {sup 238}Pu activities in INEEL soils suggest that airborne releases of Pu from the ICPP, over its operating history, may have recently been overestimated.

  18. Evaluation of the ²³?Pu prompt fission neutron spectrum induced by neutrons of 500 keV and associated covariances

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Neudecker, D.; Talou, P.; Kawano, T.; Smith, D. L.; Capote, R.; Rising, M. E.; Kahler, A. C.

    2015-08-01T23:59:59.000Z

    We present evaluations of the prompt fission neutron spectrum (PFNS) of ²³?Pu induced by 500 keV neutrons, and associated covariances. In a previous evaluation by Talou et al. 2010, surprisingly low evaluated uncertainties were obtained, partly due to simplifying assumptions in the quantification of uncertainties from experiment and model. Therefore, special emphasis is placed here on a thorough uncertainty quantification of experimental data and of the Los Alamos model predicted values entering the evaluation. In addition, the Los Alamos model was extended and an evaluation technique was employed that takes into account the qualitative differences between normalized model predicted valuesmore »and experimental shape data. These improvements lead to changes in the evaluated PFNS and overall larger evaluated uncertainties than in the previous work. However, these evaluated uncertainties are still smaller than those obtained in a statistical analysis using experimental information only, due to strong model correlations. Hence, suggestions to estimate model defect uncertainties are presented, which lead to more reasonable evaluated uncertainties. The calculated keff of selected criticality benchmarks obtained with these new evaluations agree with each other within their uncertainties despite the different approaches to estimate model defect uncertainties. The keff one standard deviations overlap with some of those obtained using ENDF/B-VII.1, albeit their mean values are further away from unity. Spectral indexes for the Jezebel critical assembly calculated with the newly evaluated PFNS agree with the experimental data for selected (n,?) and (n,f) reactions, and show improvements for high-energy threshold (n,2n) reactions compared to ENDF/B-VII.1.« less

  19. Radiolysis of Salts and Long-Term Storage Issues for Both Pure and Impure PuO{sub 2} Materials in Plutonium Storage Containers

    SciTech Connect (OSTI)

    Lav Tandon

    2000-05-01T23:59:59.000Z

    The Material Identification and Surveillance (MIS) project sponsored a literature search on the effects of radiation on salts, with focus on alkali chlorides. The goal of the survey was to provide a basis for estimating the magnitude of {alpha} radiation effects on alkali chlorides that can accompany plutonium oxide (PuO{sub 2}) into storage. Chloride radiolysis can yield potentially corrosive gases in plutonium storage containers that can adversely affect long-term stability. This literature search was primarily done to provide a tutorial on this topic, especially for personnel with nonradiation chemistry backgrounds.

  20. (239)Pu neutron resonance parameters revisited and covariance matrix in the neutron energy range from thermal to 2.5 keV

    SciTech Connect (OSTI)

    Derrien, Herve [ORNL; Leal, Luiz C [ORNL; Larson, Nancy M [ORNL

    2008-01-01T23:59:59.000Z

    To obtain the resonance parameters in a single energy range up to 2.5 keV neutron energy and the corresponding covariance matrix, a reevaluation of 239Pu was performed with the analysis code SAMMY. The most recent experimental data were analyzed in the energy range thermal to 2.5 keV. The experimental data were renormalized, aligned on a common energy scale, and corrected for residual background. Average neutron transmission and cross sections calculated with the new resonance parameters were compared to the corresponding experimental data and to ENDF/B-VI.

  1. 2006 Long Range Development Plan Final Environmental Impact Report

    E-Print Network [OSTI]

    Philliber, Jeff

    2007-01-01T23:59:59.000Z

    Glenwood, New Mexico, [WTI 02-1], 1995. Zeiner, D.C. , W.F.Manual, Glenwood, New Mexico, [WTI 02-1], 1995. Yee, Henry,

  2. Colorado geothermal commercialization program. Geothermal energy opportunities at four Colorado towns: Durango, Glenwood Springs, Idaho Springs, Ouray

    SciTech Connect (OSTI)

    Coe, B.A.; Zimmerman, J.

    1981-01-01T23:59:59.000Z

    The potential of four prospective geothermal development sites in Colorado was analyzed and hypothetical plans prepared for their development. Several broad areas were investigated for each site. The first area of investigation was the site itself: its geographic, population, economic, energy demand characteristics and the attitudes of its residents relative to geothermal development potential. Secondly, the resource potential was described, to the extent it was known, along with information concerning any exploration or development that has been conducted. The third item investigated was the process required for development. There are financial, institutional, environmental, technological and economic criteria for development that must be known in order to realistically gauge the possible development. Using that information, the next concern, the geothermal energy potential, was then addressed. Planned, proposed and potential development are all described, along with a possible schedule for that development. An assessment of the development opportunities and constraints are included. Technical methodologies are described in the Appendix. (MHR)

  3. Supporting evidence for double-C curve kinetics in the isothermal (delta) --> (alpha)' phase transformation in a Pu-Ga alloy

    SciTech Connect (OSTI)

    Oudot, B; Blobaum, K M; Wall, M A; Schwartz, A J

    2006-07-21T23:59:59.000Z

    Time-temperature-transformation (TTT) diagrams for the {delta} {yields} {alpha}{prime} transformation in a number of Pu-Ga alloys were first reported in 1975 by Orme et al. Unlike typical single-C curve kinetics observed in most isothermal martensitic transformations, the Pu-1.9 at.% Ga alloy exhibits two noses, and thus double-C curve kinetics. The authors attributed the occurrence of the double C to a difference in mechanism: a massive transformation for the upper C and a martensitic transformation for the lower C. Since that time, the nature, and the existence of the double C have received only limited attention. The results of Deloffre et al. suggest a confirmation of this behavior, but the fundamental origin of the double C remains unknown. Here, we apply differential scanning calorimetry (DSC) as an alternative approach to acquiring the TTT data and our experimental evidence suggests a confirmation of the double-C behavior after 18 hours of isothermal hold time. In addition, we report three exothermic peaks corresponding to transformations during cooling at 20 C/min prior to the isothermal holds. These three peaks are reproducible and suggest a number of possibilities for the origin of the unique kinetics: {alpha}{prime} forms with different morphologies, or from different embryos in the upper and lower C curves; {alpha}{prime} forms directly in one C curve and forms via an intermediate phase in the other C curve; the two C curves result from {alpha}{prime} forming by two or more distinct mechanisms (e.g., massive and martensitic transformations).

  4. Daily Reporting Rainfall Station DAWSON RIVER Manual River Station

    E-Print Network [OSTI]

    Greenslade, Diana

    Bridge Barakula Bawnduggie TM Auburn R Chinchilla Glenwood Beruna Wombalano Ballon TM Beckers TM Dawson

  5. Thermal conductivity of Na/sub 3/(U/sub 1-y/Pu/sub y/)O/sub 4/: A preliminary in-pile determination

    SciTech Connect (OSTI)

    Lee, M.J.; Lambert, J.D.B.; Ukai, S.; Odo, T.

    1987-01-01T23:59:59.000Z

    During Run-Beyond-Cladding-Breach (RBCB) operation in an oxide LMR, the performance of a breached fuel element is intimately associated with the formation of fuel-sodium reaction product (FSRP), Na/sub 3/(U/sub 1-y/Pu/sub y/)O/sub 4/. In-pile experiments coupled with destructive examinations of breached fuel have consistently revealed noticeable changes in fuel structure accompanying FSRP formation at the fuel surface. Previous analyses have also indicated a significant impact of FSRP on fuel centerline temperature. Successful modeling of breached fuel thermal behavior therefore requires a reasonably accurate knowledge of the thermal properties of the FSRP, especially its thermal conductivity. But laboratory investigations have been scarce and limited to the Na/UO/sub 2/ system because of the toxicity of plutonium and hygroscopicity of the FSRP. Hence, post-irradiation observations of fuel samples remain the most amenable way of deriving the thermal conductivity of the FSRP. Such work is a spin-off of the RBCB program in the Experimental Breeder Reactor-II (EBR-II), a program jointly sponsored by the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan.

  6. Thermal conductivity of Na/sub 3/(U/sub 1-y/Pu/sub y/O/sub 4/: a preliminary in-pile determination

    SciTech Connect (OSTI)

    Lee, M.J.; Lambert, J.D.B.; Ukai, S.; Odo, T.

    1987-01-01T23:59:59.000Z

    During run-beyond-cladding-breach (RBCB) operation in an oxide liquid metal breeder, the performance of a breached fuel element is intimately associated with the formation of fuel/sodium reaction product (FSRP), Na/sub 3/(U/sub 1-y/Pu/sub y/)O/sub 4/. In-pile experiments coupled with destructive examinations of breached fuel have consistently revealed noticeable changes in fuel structure accompanying FSRP formation at the fuel surface. Previous analyses have also indicated a significant impact of FSRP on fuel centerline temperature. Successful modeling of breached fuel thermal behavior therefore requires a reasonably accurate knowledge of the thermal properties of the FSRP, especially its thermal conductivity. But laboratory investigations have been scarce and limited to the Na/UO/sub 2/ system because of the toxicity of plutonium and hygroscopicity of the FSRP. Hence, postirradiation observations of fuel samples remain the most amenable way of deriving the thermal conductivity of the FSRP. Such work is a spin-off of the RBCB program in the Experimental Breeder Reactor-II (EBR-II), a program jointly sponsored by the US Dept. of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan.

  7. Microdistribution and Long-Term Retention of 239Pu (NO3)4 in the Respiratory Tracts of an Acutely Exposed Plutonium Worker and Experimental Beagle Dogs

    SciTech Connect (OSTI)

    Nielsen, Christopher E.; Wilson, Dulaney A.; Brooks, Antone L.; McCord, Stacey; Dagle, Gerald E.; James, Anthony C.; Tolmachev, Sergei Y.; Thrall, Brian D.; Morgan, William F.

    2012-11-01T23:59:59.000Z

    The long-term retention of inhaled soluble forms of plutonium raises concerns as to the potential health effects in persons working in nuclear energy or the nuclear weapons program. The distributions of long-term retained inhaled plutonium-nitrate [239Pu (NO3)4] deposited in the lungs of an accidentally exposed nuclear worker (Human Case 0269) and in the lungs of experimentally exposed beagle dogs with varying initial lung depositions were determined via autoradiographs of selected histological lung, lymph node, trachea, and nasal turbinate tissue sections. These studies showed that both the human and dogs had a non-uniform distribution of plutonium throughout the lung tissue. Fibrotic scar tissue effectively encapsulated a portion of the plutonium and prevented its clearance from the body or translocation to other tissues and diminished dose to organ parenchyma. Alpha radiation activity from deposited plutonium in Human Case 0269 was observed primarily along the sub-pleural regions while no alpha activity was seen in the tracheobronchial lymph nodes of this individual. However, relatively high activity levels in the tracheobronchial lymph nodes of the beagles indicated the lymphatic system was effective in clearing deposited plutonium from the lung tissues. In both the human case and beagle dogs, the appearance of retained plutonium within the respiratory tract was inconsistent with current biokinetic models of clearance for soluble forms of plutonium. Bound plutonium can have a marked effect on the dose to the lungs and subsequent radiation exposure has the potential increase in cancer risk.

  8. Procedures for the use of Lexan and Makrofol SSNTDs in the detection of environmental concentrations of {sup 235}U and {sup 239}Pu

    SciTech Connect (OSTI)

    Henderson, C.L.

    1993-03-01T23:59:59.000Z

    Solid State Nuclear Track Detectors are used to study a variety of atomic particles. Polycarbonate SSNTD is used to study environmental concentrations of {sup 235}U and {sup 239}Pu in human urine and feces through fission track analysis. The samples of interest are deposited upon a Lexan slide, covered with a piece of Makrofol and exposed to a neutron fluence of 1.1 X 10{sup 17}. The fissile isotopes in the sample fission and the resulting fission fragments pass through either the surface of the Lexan or the surface of the Makrofol. The positive Coulombic attraction of the ionized fission fragments causes the electrons of the polycarbonate lattice to move towards the path of these particles, resulting in the breakage of chemical bonds in the lattice. The detector is then chemically etched in 6.5 N KOH that preferentially dissolves the damaged polycarbonate left in the path of the fission fragment. The chemically etched fission tracks are permanent records of the path of the fission fragment. The etched fission tracks in Lexan are optically counted using a microscope and the fission tracks in Makrofol are counted using a Spark Chamber. The amount of fissile material in the original sample can be calculated from the number of fission tracks. This paper presents further details of procedures for etching fission tracks in Lexan and Makrofol and for operating a Spark Chamber to count etched fission tracks in Makrofol. The physics of fission track formation in dielectric detectors is also discussed, as well as the physics of the Spark Chamber.

  9. Theoretical analyses of (n,xn) reactions on sup 235 U, sup 238 U, sup 237 Np, and sup 239 Pu for ENDF/B-VI

    SciTech Connect (OSTI)

    Young, P.G.; Arthur, E.D.

    1991-01-01T23:59:59.000Z

    Theoretical analyses were performed of neutron-induced reactions on {sup 235}U, {sup 238}U, {sup 237}Np, and {sup 239}Pu between 0.01 and 20 MeV in order to calculate neutron emission cross sections and spectra for ENDF/B-VI evaluations. Coupled-channel optical model potentials were obtained for each target nucleus by fitting total, elastic, and inelastic scattering cross section data, as well as low-energy average resonance data. The resulting deformed optical model potentials were used to calculate direct (n,n{prime}) cross sections and transmission coefficients for use in Hauser-Feshbach statistical theory analyses. A fission model with multiple barrier representation, width fluctuation corrections, and preequilibrium corrections were included in the analyses. Direct cross sections for higher-lying vibrational states were calculated using DWBA theory, normalized using B(E{ell}) values determined from (d,d{prime}) and Coulomb excitation data, where available, and from systematics otherwise. Initial fission barrier parameters and transition state density enhancements appropriate to the compound systems involved were obtained from previous analyses, especially fits to charged-particle fission probability data. The parameters for the fission model were adjusted for each target system to obtain optimum agreement with direct (n,f) cross section measurements, taking account of the various multichance fission channels, that is, the different compound systems involved. The results from these analyses were used to calculate most of the neutron (n,n), (n,n{prime}), and (n,xn) cross section data in the ENDF/B/VI evaluations for the above nuclei, and all of the energy-angle correlated spectra. The deformed optical model and fission model parameterizations are described. Comparisons are given between the results of these analyses and the previous ENDF/B-V evaluations as well as with the available experimental data. 14 refs., 3 figs., 1 tab.

  10. Anaerobic Biotransformation and Mobility of Pu and PuEDTA

    SciTech Connect (OSTI)

    Xun, Luying

    2005-06-01T23:59:59.000Z

    Although our goal is to isolate anaerobic EDTA degraders, we initiated the experiments to include nitrilotriacetate (NTA), which is a structure homologue of EDTA. All the aerobic EDTA degraders can degrade NTA, but the isolated NTA degraders cannot degrade EDTA. Since NTA is a simpler structure homologue, it is likely that EDTA-degrading ability is evolved from NTA degradation. This hypothesis is further supported from our characterization of EDTA and NTA-degrading enzymes and genes (J. Bact. 179:1112-1116; and Appl. Environ. Microbiol. 67:688-695). The EDTA monooxygenase and NTA monooxygenase are highly homologous. EDTA monooxygenase can use both EDTA and NTA as substrates, but NTA monooxygenase can only use NTA as a substrate. Thus, we put our effort to isolate both NTA and EDTA degraders. In case, an anaerobic EDTA degrader is not immediately enriched, we will try to evolve the NTA degraders to use EDTA. Both aerobic and anaerobic enrichment cultures were set.

  11. Anaerobic Biotransformation and Mobility of Pu and PuEDTA

    SciTech Connect (OSTI)

    Xun, Luying

    2005-06-01T23:59:59.000Z

    The objective of this report is to isolate anaerobic EDTA-degrading bacteria. Although our goal is to isolate anaerobic EDTA degraders, we initiated the experiments to include nitrilotriacetate (NTA), which is a structure homologue of EDTA. All the aerobic EDTA degraders can degrade NTA, but the isolated NTA degraders cannot degrade EDTA. Since NTA is a simpler structure homologue, it is likely that EDTA-degrading ability is evolved from NTA degradation. This hypothesis is further supported from our characterization of EDTA and NTA-degrading enzymes and genes (J. Bact. 179:1112-1116; and Appl. Environ. Microbiol. 67:688-695). The EDTA monooxygenase and NTA monooxygenase are highly homologous. EDTA monooxygenase can use both EDTA and NTA as substrates, but NTA monooxygenase can only use NTA as a substrate. Thus, we put our effort to isolate both NTA and EDTA degraders. In case, an anaerobic EDTA degrader is not immediately enriched, we will try to evolve the NTA degraders to use EDTA. Both aerobic and anaerobic enrichment cultures were set.

  12. Desired PU Loading During Vitrification

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic Plan Department of EnergyDeputy Secretary visits Oak

  13. The Political Rebellion of Carey McWilliams

    E-Print Network [OSTI]

    Critser, Greg

    1983-01-01T23:59:59.000Z

    May in Bruno Lasky, "Feudalism in California," The Survey 42to one reminiscent of feudalism in the middle ages." ^^ j^ a

  14. Antioxidant behavior in flexible PU foam

    SciTech Connect (OSTI)

    Skorpenske, R.G.; Schrock, A.K.

    1991-09-01T23:59:59.000Z

    In this article, data are given regarding the AO concentration profiles across foam buns as a function of the corresponding temperature profile of the foams studied, three were produced on a Varimax pilot line located at Dow Chemical Company Freeport, Texas. The fourth foam used for this study was a production scale foam made at Texas Fibers, a Division of Leggett and Platt, Brenham, Texas. The foams produced on the Varimas include formulations using 4.1, 5.1 and 6.1 parts per hundred parts (pphp) water based on polyol and can be compared to the 4.1 pphp water foam made at Texas Fibers. Temperature data, collected from a foam-in-place grid of thermocouples, gives the time-temperature profile within the foam bun as a function of location. Foam samples which have been removed from locations corresponding to the thermocouples are examined, via methylene chloride extraction and liquid chromatographic analysis, for antioxidant content. The objective is to determine the significance of the foam environment, as a function of formulation, on the behavior of antioxidants.

  15. Cs and 239,240 Pu concentration

    E-Print Network [OSTI]

    Buesseler, Ken

    weapons test sites, radioactive waste dumping sites and from possible nuclear accidents can be identified. Noshkinh , Shigeki Shimai , Orihiko Togawaa,1 a International Atomic Energy Agency, Marine Environment: p.povinec@iaea.org (P.P. Povinec). 1 Present address: Japan Atomic Energy Research Institute

  16. Plutonium stabilization and handling (PuSH)

    SciTech Connect (OSTI)

    Weiss, E.V.

    1997-01-23T23:59:59.000Z

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

  17. NR Pu SEIS Advisory 07272012_Clean

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Saleshttp://www.fnal.gov/directorate/nalcal/nalcal02_07_05_files/nalcal.gif Directorate1,Stewardship

  18. Examining 239Pu and 240Pu Nuclear Resonance Fluorescence Measurements on Spent Fuel for Nuclear Safeguards

    E-Print Network [OSTI]

    Quiter, Brian

    2013-01-01T23:59:59.000Z

    Spent Fuel Assay Using Nuclear Resonance Fluo- rescence,” Annual Meeting of the Institute of Nuclear Material Management,

  19. Examining 239Pu and 240Pu Nuclear Resonance Fluorescence Measurements on Spent Fuel for Nuclear Safeguards

    E-Print Network [OSTI]

    Quiter, Brian

    2013-01-01T23:59:59.000Z

    10- 01096) Journal of Nuclear Technology, p. 150, Vol. 175,linac and laser technologies for nuclear photonics gamma-rayNuclear resonance fluorescence (NRF) has been identified as a technology

  20. PowerPoint Presentation

    Broader source: Energy.gov (indexed) [DOE]

    - 400sec 464 MMscfd - Minimum Pressure - 830 psi CAES Aquifer Storage System Geology of Iowa Mt. Simon Sandstone Jordan Sandstone St. Peter Sandstone Glenwood Fm....

  1. Microsoft Word - Ex Parte Memo re October 28, 2014 Meeting on...

    Energy Savers [EERE]

    or on the telephone: John Cymbalsky, DOE Dan Cohen, DOE Andrew deLaski, ASAP Wade Smith, AMCA International Marc Bublitz, New York Blower Company Tom Catania, University of...

  2. Photochemical oxidation of oxalate in Pu-238 process streams

    SciTech Connect (OSTI)

    Long, K. M. (Kristy M.); Ford, D. K. (Doris K.); Trujillo, L. A. (Leonardo A.)

    2003-01-01T23:59:59.000Z

    For over forty years, NASA has relied on plutonium-238 in Radioisotope Thermoelectric Generator (RTG) units and Radioisotope Heater Units ( W s ) to provide power and heat for many space missions including Transit, Pioneer, Viking, Voyager, Galileo, Ulysses and Cassini. RHUs provide heat to keep key components warm in extremely cold environments found on planets, moons, or in deep space. RTGs convert heat generated from the radioactive decay of plutonium-238 into electricity using a themocouple, Plutonium-238 has proven to be an excellent heat source far deep space missions because of its high thermal power density, useful lifetime, minimal shielding requirements, and oxide stability.

  3. Ceramicrete stabilization of U-and Pu-bearing materials

    DOE Patents [OSTI]

    Wagh, Arun S. (Naperville, IL); Maloney, M. David (Evergreen, CO); Thompson, Gary H. (Thornton, CO)

    2007-11-13T23:59:59.000Z

    A method of stabilizing nuclear material is disclosed. Oxides or halides of actinides and/or transuranics (TRUs) and/or hydrocarbons and/or acids contaminated with actinides and/or TRUs are treated by adjusting the pH of the nuclear material to not less than about 5 and adding sufficient MgO to convert fluorides present to MgF.sub.2; alumina is added in an amount sufficient to absorb substantially all hydrocarbon liquid present, after which a binder including MgO and KH.sub.2PO.sub.4 is added to the treated nuclear material to form a slurry. Additional MgO may be added. A crystalline radioactive material is also disclosed having a binder of the reaction product of calcined MgO and KH.sub.2PO.sub.4 and a radioactive material of the oxides and/or halides of actinides and/or transuranics (TRUs). Acids contaminated with actinides and/or TRUs, and/or actinides and/or TRUs with or without oils and/or greases may be encapsulated and stabilized by the binder.

  4. actinides np pu: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    that hard sets S for NP must have exponential density, i.e. |Sn| 2n for some > 0 for Turing reductions that make n1- queries. In addition we study the instance complexity of...

  5. PuLP: A Linear Programming Toolkit for Python

    E-Print Network [OSTI]

    2011-09-05T23:59:59.000Z

    Sep 5, 2011 ... Many mixed-integer linear programming (MILP) solvers are .... tions, to place facilities and assigns the production of n products to these ...

  6. aktivnostej izotopov 239pu: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    were found between mean carcass radionuclide concentrations and mean pelt radionuclide concentrations, indicating that the two primary modes of contamination may be...

  7. NR Pu SEIS Advisory 07152010 _final_.doc

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparency Visit | National Nuclear Security Administration

  8. Nuclear magnetic resonance offers new insights into Pu 239

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for Renewable Energy:Nanowire3627 Federal Register /76SafeguardsSystemsTesting

  9. Separating expansion from contraction: generalized TOV condition, LTB models with pressure and CDM

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    , we adopt the Generalised Painlevé-Gullstrand (hereafter GPG) formalism used in Lasky & Lun [6], which TO LTB MODELS IN GPG SYSTEM We consider a spherically symmetric Generalised Lemaître-Tolman-Bondi metric to include pressure. Performing an ADM 3+1 splitting in the GPG coordinates [6] , the metric reads ds2 = -(t

  10. Synthesis and characterization of visible emission from rare-earth doped aluminum nitride, gallium nitride and gallium aluminum nitride powders and thin films

    E-Print Network [OSTI]

    Tao, Jonathan Huai-Tse

    2010-01-01T23:59:59.000Z

    from GaN:Tb 3+ Powders and Thin Films Deposited by MOVPE andHirata, "Eu 3+ Activated GaN Thin Films Grown on Sapphire byTb 3+ in GaN Powders and Thin Films," ECS Trans. , J. Laski,

  11. Code Generation for WSLAs using AXpect Galen S. Swint and Calton Pu, Senior Member, IEEE

    E-Print Network [OSTI]

    Pu, Calton

    of companies such as IBM and HP trend towards specifying web service contracts that capture expecta- tions, becomes one of mapping the service contract's particular domain specific lan- guage (DSL investigate this question and probe the power of AOP in implementing these service contracts by using IBM

  12. Quantitative NDA Measurements of Advanced Reprocessing Product Materials Containing U, NP, PU, and AM

    E-Print Network [OSTI]

    Goddard, Braden

    2013-04-05T23:59:59.000Z

    of this first principle technique have been identified: (1) quantitative measurement of uranium, neptunium, plutonium, and americium materials; (2) quantitative measurement of mixed oxide (MOX) materials; (3) quantitative measurement of uranium materials; and (4...

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

    DOE Patents [OSTI]

    Lloyd, Milton H. (Oak Ridge, TN)

    1983-01-01T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Lloyd, M.H.

    1981-01-09T23:59:59.000Z

    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.

  15. Rational Ligand Design for U(VI) and Pu(IV)

    E-Print Network [OSTI]

    Szigethy, Geza

    2010-01-01T23:59:59.000Z

    solutions into appropriate fractions and physical forms can be illustrated by the nuclear waste forms found at the Savannah River

  16. Type B Accident Investigation on the August 5, 2003, Pu-238 Multiple...

    Energy Savers [EERE]

    Board concluded that the direct cause of the accident was the release of airborne contamination from a degraded package that contained cellulose material and plutonium-238...

  17. Rational Ligand Design for U(VI) and Pu(IV)

    E-Print Network [OSTI]

    Szigethy, Geza

    2010-01-01T23:59:59.000Z

    Murali, M. S. ; Nash, K. L. Solv. Extr. Ion Exch. 2001, 19,D. C. ; Raymond, K. N. Solv. Extr. Ion Exch. 2004, 22, (22)DMF) and UO 2 (bis-Me-3,2-HOPO)(solv) tabulated in Table 2-

  18. {sup 239}Pu Holdup Measurements at Savannah River Site's FB-Line

    SciTech Connect (OSTI)

    Hodge, C.A.

    2001-06-20T23:59:59.000Z

    Plutonium holdup measurements were conducted in the dry cabinets of FB-Line at the Savannah River Site. This report will discuss the methodology, measurements, assumptions, calculations, and corrections.

  19. Materials Data on PuCo3 (SG:166) by Materials Project

    DOE Data Explorer [Office of Scientific and Technical Information (OSTI)]

    Kristin Persson

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

  20. Materials Data on PuGe2 (SG:141) by Materials Project

    DOE Data Explorer [Office of Scientific and Technical Information (OSTI)]

    Kristin Persson

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

  1. Materials Data on Pu2Co (SG:189) by Materials Project

    DOE Data Explorer [Office of Scientific and Technical Information (OSTI)]

    Kristin Persson

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

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

    SciTech Connect (OSTI)

    Kelly, Julian F. [Thor Energy AS, Sommerrogaten 13-15, Oslo 0255 (Norway)] [Thor Energy AS, Sommerrogaten 13-15, Oslo 0255 (Norway); Franceschini, Fausto [Westinghouse Electric Company LLC, 1000 Cranberry Woods Drive, Cranberry Township, PA 16066 (United States)] [Westinghouse Electric Company LLC, 1000 Cranberry Woods Drive, Cranberry Township, PA 16066 (United States)

    2013-07-01T23:59:59.000Z

    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)

  3. Microsoft Word - Template_SLAC Proprietary Use Agreement_PU 11...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    PROPERTY AND MATERIALS** USER may be permitted by the CONTRACTOR to furnish equipment, tooling, test apparatus, or materials necessary to assist in the performance of its...

  4. Oakwood crown closure estimation by unmixing Landsat TM data R. PU*, B. XU and P. GONG

    E-Print Network [OSTI]

    Silver, Whendee

    classi- fication and tree crown closure estimation. However, photo interpretation is dependent on the experience of photo interpretators (Biging et al. 1991, Gong and Chen 1992, Davis et al. 1995) and needs.g. Landsat sensor imagery) classification as an information extraction tool has been used for more than three

  5. El TRANSITION PROBABILITIES FROM Kn= O-AND Kn=1- STATES OF 238Pu

    E-Print Network [OSTI]

    Lederer, C. Michael

    2012-01-01T23:59:59.000Z

    By contrast, the El transitions in gadolinium and dysprosiumL lines of the 44.11-keV transition were also measured, butfor some anomalous El transitions in the actinide region.

  6. Materials Data on PuGe2 (SG:141) by Materials Project

    SciTech Connect (OSTI)

    Kristin Persson

    2014-11-02T23:59:59.000Z

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

  7. SYNTHESIS AND CHARACTERIZATION OF Pa(IV), Np(IV), AND Pu(IV) BOROHYDRIDES

    E-Print Network [OSTI]

    Banks, R.H.

    2010-01-01T23:59:59.000Z

    due to B and 'H (in the deuteride spectra) are resolved inresonances compared to the deuteride and a reliable g_ valuethe calculations of the deuteride data improved the fit sven

  8. Neutronics Design and Fuel Cycle Analysis of a High Conversion BWR with Pu-Th Fuel

    SciTech Connect (OSTI)

    Xu, Yunlin; Downar, T.J. [Purdue University, West Lafayette, IN 47906-1290 (United States); Takahashi, H.; Rohatgi, U.S. [Brookhaven National Laboratory, Upton, New York 11973 (United States)

    2002-07-01T23:59:59.000Z

    As part of the U.S. Department of Energy's (DOE) Nuclear Energy Research Initiative (NERI), a 'Generation IV' high conversion Boiling Water Reactor design is being investigated at Purdue University and Brookhaven National Laboratory. One of the primary innovative design features of the core proposed here is the use of Thorium as fertile material. In addition to the advantageous nonproliferation and waste characteristics of thorium fuel cycles, the use of thorium is particularly important in a tight pitch, high conversion lattice in order to insure a negative void coefficient throughout the operating life of the reactor. The principal design objective of a high conversion light water reactor is to substantially increase the conversion ratio (fissile atoms produced per fissile atoms consumed) of the reactor without compromising the safety performance of the plant. Since existing LWRs have a relatively low conversion ratio they require relatively frequent refueling which limits the economic efficiency of the plant. Also, the high volume of spent fuel can pose a burden for waste storage and the accumulation of plutonium in the uranium fuel cycle can become a materials proliferation issue. The development of Fast Breeder Reactors (FBR) as an alternative technology to alleviate some of these concerns has been delayed for various reasons. An intermediate solution has been to examine tight pitch light water reactors which can provide significant improvements in the fuel cycle performance of the existing LWRs by taking advantage of the increased conversion ratios from the harder neutron spectrum in the tight pitch lattice, as well as the by taking advantage of the waste and nonproliferation benefits of the thorium fuel cycle. Several High Conversion BWR designs have been proposed by researchers in Japan and elsewhere during the past several years. One of the more promising HCR designs is the Reduced Moderation Water Reactor (RMWR) proposed by JAERI [1]. Their design was based on a uranium fuel cycle and showed significant improvements in the fuel cycle performance compared to conventional BWRs. However, one of the drawbacks of their design was the potential for a positive void coefficient. In order to insure a negative void coefficient, the JAERI researchers designed a 'flat core' and introduced void tube assemblies in order to enhance neutron leakage in the event of core voiding. The use of thorium in the Purdue/BNL HCBWR design proposed here obviates the need for void tubes and makes it possible to increase the core height and improve neutron economy without the risk of a positive void coefficient. The principal reason for the improvement in the void coefficient is because Th-232 has a smaller fast fission cross section and resonance integral than U-238. In the design proposed here, it is possible to eliminate the void tubes in the RMWR design and replace the axial blanket with active fuel to increase the core height and further improve neutron economy. The core analyses in the work here was performed with the Purdue Fuel Management Code System [2] which is based on the Studsvik/Scandpower lattice physics code HELIOS, and the U.S. NRC core neutronics simulator, PARCS, which is coupled to the thermal-hydraulics code RELAP5. All these codes have been well assessed and benchmarked for analysis of light water reactor systems. (authors)

  9. Pb-210 and Pu-239,240 in nearshore Gulf of Mexico sediments

    E-Print Network [OSTI]

    Rotter, Richard Joseph

    1985-01-01T23:59:59.000Z

    was taken as a separate aliquot from the same bulk sediment as that used for Pb-210 analysis. Dissolved sediment samples were taken up in approximately 200 ml of 6M HC1 and loaded into gas collection bottles. These were then stripped of all radon, sealed...

  10. Quantitative NDA Measurements of Advanced Reprocessing Product Materials Containing U, NP, PU, and AM 

    E-Print Network [OSTI]

    Goddard, Braden

    2013-04-05T23:59:59.000Z

    The ability of inspection agencies and facility operators to measure powders containing several actinides is increasingly necessary as new reprocessing techniques and fuel forms are being developed. These powders are difficult to measure...

  11. Rational Ligand Design for U(VI) and Pu(IV)

    E-Print Network [OSTI]

    Szigethy, Geza

    2010-01-01T23:59:59.000Z

    V.4.024; Siemens Industrial Automation, Inc, Madison, WI,V.4.024; Siemens Industrial Automation, Inc, Madison, WI,

  12. ings regarding the emerging complexity of pu-tatively simple metals under pressure (1, 2, 5,

    E-Print Network [OSTI]

    Ballarini, Roberto

    , in Proceedings of the International School of Physics "Enrico Fermi," R. J. Hemley, G. Chiarotti, M. Bernasconi- dation. W.G. acknowledges support from REU Site for Undergraduate Research Training in Geoscience (NSF

  13. Metrology and quality assurance from surveillance of gas compositions over PuO[sub 2

    SciTech Connect (OSTI)

    Worl, L. A. (Laura A.); French, Catherine A.; Kreyer, L. S. (Lawrence S.)

    2002-01-01T23:59:59.000Z

    Until the late 1980s, a primary mission of the Department of Energy (DOE) has been the production of nuclear materials for nuclear weapons. Termination of the Cold War in 1989 and the subsequent nuclear weapons treaties dramatically decreased the inventory needs for nuclear weapons. These activities resulted in the consolidation of nuclear material inventories and activities, generating substantial amounts of surplus nuclear materials ranging from plutonium metal and pure oxides to impure plutonium residues. Packaging and storage of these materials in physically and environmentally safe configurations for significant time periods were required. In 1993 the Defense Nuclear Facility Safety Board (DNFSB) and the DOE Office of Nuclear Safety examined the storage of metal and oxides at the Rocky Flats Plant that ultimately resulted in recommendation 94-1, calling for a standard to define the processing and storage of plutonium bearing materials. This recommendation generated a standard for storage of plutonium metals and oxides, DOE-STD-3013-2000, which is now in its fourth revision. The current DOE 3013 Standard is limited to metal and oxides, which contain greater than 30 weight percent plutonium and uranium. The 3013 Standard requires that the oxide be calcined to 950 C for two hours in an oxidizing environment. Before packaging, the oxide is required to have less than 0.5 weight percent moisture. Up to five kilograms of the stabilized oxide material is subsequently sealed in a set of two-nested welded stainless steel container, which must have a power less than 19 Watts.

  14. Spin-orbit holds the heavyweight title for Pu and Am: Exchange regains it for Cm

    SciTech Connect (OSTI)

    Moore, K; der Laan, G v; Soderlind, P

    2008-01-10T23:59:59.000Z

    The conclusions of this paper are: (1) The 5f electrons in Cm are near an LS coupling scheme. (2) This coupling scheme allows for a large spin polarization of the 5f electrons, which in turn stabilizes the Cm III crystal structure. (3) Results for Cm show us the recipe for magnetic stabilization of the crystal structure of metals: (A) The metal must be near the itinerant-localized transition where multiple crystal structures have close energies; (B) The metal is just on the magnetic side of the transition; and (C) There must be a magnetic moment large enough to overcome the energy difference between crystal structures, thus dictating the atomic geometry. (4) These results solidify our understanding of magnetically-stabilized metals, showing us where to look for engineered materials with magnetic applications.

  15. Optimal power flow in microgrids using event-triggered optimization Pu Wan and Michael D. Lemmon

    E-Print Network [OSTI]

    Lemmon, Michael

    be accomplished through small micro- turbines and gas/diesel generators. Storage devices such as battery banks Abstract-- Microgrids are power generation and distribution systems in which users and generators are in close proximity. They usually have limited power generation capacity, and are networked together to meet

  16. Materials Data on PuAl3 (SG:194) by Materials Project

    SciTech Connect (OSTI)

    Kristin Persson

    2014-11-02T23:59:59.000Z

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

  17. Materials Data on PuRh3 (SG:221) by Materials Project

    SciTech Connect (OSTI)

    Kristin Persson

    2014-11-02T23:59:59.000Z

    Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

  18. Evaluation of Los Alamos National Laboratory (LANL) PU238 Waste Management

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic Plan Departmentof1-SCORECARD-09-21-11 Page5-03 EvaluationStorage SitePractices

  19. Savannah River Site: Plutonium Preparation Project (PuPP) at Savannah River

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOrigin ofEnergy atLLC - FE DKT. 10-160-LNG -Energy Proposed1-EReviewLaboratory |Site |

  20. Shenzhen Prosunpro PengSangPu Solar Industrial Products Corporation | Open

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnualProperty Edit with form HistoryRistma AGShandong LusaShelby, Ohio:ShenyuShenzhenHengEnergy

  1. Multiscale Speciation of U and Pu at Chernobyl, Hanford, Los Alamos,

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItemResearch > The EnergyCenterDioxide CaptureSee the Foundry'sMcGuire AFB, Mayak, and

  2. Table A1. Total First Use (formerly Primary Consumption) of Energy for All Pu

    U.S. Energy Information Administration (EIA) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are nowTotal" (Percent) Type: Sulfur Content API GravityDakota" "Fuel, quality", 2013,Iowa"Dakota" ,"FullWest Virginia"1 " "

  3. Table A1. Total First Use (formerly Primary Consumption) of Energy for All Pu

    U.S. Energy Information Administration (EIA) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are nowTotal" (Percent) Type: Sulfur Content API GravityDakota" "Fuel, quality", 2013,Iowa"Dakota" ,"FullWest Virginia"1 " "2"

  4. Type B Accident Investigation on the August 5, 2003, Pu-238 Multiple Uptake

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently AskedEnergyIssuesEnergyTransportation WorkDecemberInjury at theInjury on

  5. V-213: PuTTY SSH Handshake Integer Overflow Vulnerabilities | Department of

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742Energy China 2015ofDepartment of Energy Microsoft SecurityEnergy SEARCH-LAB has reported

  6. 6th US-Russian Pu Science Workshop Lawrence Livermore National Laboratory

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of ScienceandMesa del(ANL-IN-03-032) -Less isNFebruaryOctober 2, AlgeriaQ1 Q2 Q3 U . S . D

  7. Summary of Pu u O o - Kupaianaha Eruption, Kilauea Volcano, Hawaii | Open

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to:Ezfeedflag JumpID-f <Maintained By Fault Propagation AndInformation

  8. Llr. Norgnn of the St. Louis office tolepbonod Dr. ;PuAuff mcently

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group currentBradleyTableSelling CorpNewCF INDUSTRIES,L? .-I I2 m.m\Ll 1

  9. Microsoft PowerPoint - Draft HAB Pu presentation CJK 040412.pptx [Read-Only]

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville PowerCherries 82981-1cnHighand Retrievals fromprocessEstimating forcloud water distribution from 3

  10. Savannah River Site: Plutonium Preparation Project (PuPP) at Savannah River Site

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM615_CostNSARDevelopmentalEfficiency |91-51-SW State SouthTerrel J.

  11. Discovery of Pu-based superconductors and relation to other classes of

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power AdministrationField Campaign:INEA : Papers Subfolders inDiscovery of Dark

  12. **The specific contact information included in this document are current as of 11/18/2011. AppendixB:LargeNebraskaSchoolDistricts

    E-Print Network [OSTI]

    Farritor, Shane

    Schools High Schools Fire Ridge Elkhorn Middle School Elkhorn High School Hillrise Elk Ridge Middle School Emerson Glenwood Kenwood Meadowlark Northeast Park Riverdale Stone Windy Hills #12;**The specific contact Fredstrom Hartley Hill Holmes Humann Huntington Kahoa Kooser Lakeview Maxey McPhee Meadow Lane Morley

  13. Survey of Critical Biological Resources Garfield County, Colorado

    E-Print Network [OSTI]

    Survey of Critical Biological Resources Garfield County, Colorado Volume I Prepared for Garfield of the Garfield County Commissioners, the Planning Department, and the Assessor's office. We received much help and good advice from the Bureau of Land Management, especially Carla Scheck and Dan Sokal in the Glenwood

  14. Fast, Optimized Sun RPC Using Automatic Program Specialization Gilles Muller, Renaud Marlet, Calton Pu and Ashvin Goel

    E-Print Network [OSTI]

    Goel, Ashvin

    Fast, Optimized Sun RPC Using Automatic Program Specialization Gilles Muller, Renaud Marlet, Calton automatic optimization of an existing, commercial RPC implementation, namely the Sun RPC. The optimized Sun the original Sun RPC. Close examination of the specialized code does not reveal further optimization

  15. Universidad Catlica de Chile, 16 Noviembre 2006 0.05 p.u. aumento escalonado en torque

    E-Print Network [OSTI]

    Rudnick, Hugh

    auto búsqueda #12;2 1 0 -1 -2 0s 5s 10s 15s 20s AER en Gen. #4 2 1 0 -1 -2 -3 AER en Gen. #4 0s 5s 10s

  16. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    SciTech Connect (OSTI)

    Not Available

    1993-05-13T23:59:59.000Z

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  17. P.U. Honorary Degree Recipients 1970-2010 Last Name First Name Degree Year Awarded t

    E-Print Network [OSTI]

    Bou-Zeid, Elie

    Herbert Smith, Jr. LL.D. 1986 Baker William O. LL.D. 1993 Baker III James A. LL.D. 1991 Ball George W. LL Hughes Phillip Samuel LL.D. 1985 Jemison Mae D.Hum. 2000 Johnson, Jr. Frank M. LL.D. 1974 Jones James

  18. THERMAL EVALUATION OF THE CONCEPTUAL DHLW DISPOSAL CONTAINER LOADED WITH PU/CS GREENFIELD GLASS (SCPB: N/A)

    SciTech Connect (OSTI)

    T.L. Lotz

    1995-11-13T23:59:59.000Z

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8,4-11,4-24,5-1, and 5-13; Ref. 5.10) and Waste Package Plan (pp. 3-15,3-17, and 3-24; Ref. 5.9). The design data request addressed herein is: Characterize the conceptual Defense High Level Waste (DHLW) Disposal Container design to show that the design is feasible for use in the MGDS environment when loaded with a plutonium/cesium greenfield glass waste form. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the conceptual DHLW disposal container design do not preclude compatibility with the MGDS if it is loaded with alternate waste forms. The objective of this analysis is to provide thermal parameter information for the conceptual DHLW disposal container design loaded with an alternative waste form containing a plutonium/cesium mixture under nominal MGDS repository conditions. The results are intended to show that the design loaded with this alternative waste form has a reasonable chance to meet the MGDS design requirements for normal MGDS operation and to provide the required guidance to determining the major design issues for future design efforts. Future design efforts will focus on specific DHLW vendor designs and improved waste form data when they become available.

  19. Effect of 1-hydroxyethane-1,1-diphosphonic acid (HEDPA) on Partitioning of Np and Pu to Synthetic Boehmite

    E-Print Network [OSTI]

    Powell, Brian A.

    2010-01-01T23:59:59.000Z

    of plutonium IV and V on goethite. Geochim. Cosmo. Acta,reduction on synthetic goethite (?-FeOOH) and hematite (?-Fe

  20. Thermal Energy Harvesting with Thermoelectrics for Self-powered Sensors: With Applications to Implantable Medical Devices, Body Sensor Networks and Aging in Place

    E-Print Network [OSTI]

    Chen, Alic

    2011-01-01T23:59:59.000Z

    Pu-238) radioisotope and a thermoelectric generator. The Pu-to radioisotopes. In designing thermoelectric generators for

  1. I. Nuclear Production Reaction and Chemical Isolation Procedure for 240Am II. New Superheavy Element Isotopes: 242Pu(48Ca,5n)285-114

    E-Print Network [OSTI]

    Ellison, Paul Andrew

    2011-01-01T23:59:59.000Z

    material and common nuclear fission products in the 1 eV –of destroying long-lived nuclear fission products by neutronFlerov Laboratory of Nuclear Reactions Fission product Focal

  2. I. Nuclear Production Reaction and Chemical Isolation Procedure for 240Am II. New Superheavy Element Isotopes: 242Pu(48Ca,5n)285-114

    E-Print Network [OSTI]

    Ellison, Paul Andrew

    2011-01-01T23:59:59.000Z

    48 Ca, 5n) 285 114 nuclear reaction cross section. . . . .240 Am(n, f ) cross section 1.4 Nuclear properties of 2401.5 Nuclear reactions for the production of 240 Am . 2

  3. Environmental assessment operation of the HB-Line facility and frame waste recovery process for production of Pu-238 oxide at the Savannah River Site

    SciTech Connect (OSTI)

    NONE

    1995-04-01T23:59:59.000Z

    The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0948, addressing future operations of the HB-Line facility and the Frame Waste Recovery process at the Savannah River Site (SRS), near Aiken, South Carolina. Based on the analyses in the EA, DOE has determined that the proposed action is not a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, DOE has concluded that, the preparation of an environmental impact statement is not required, and is issuing this Finding of No Significant Impact.

  4. 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 (OSTI)

    D. L. Chichester; S. J. Thompson

    2013-09-01T23:59:59.000Z

    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 using cerium, which is rather easy to analyze using passive nondestructive analysis gamma-ray spectrometry, as a surrogate for plutonium in the final analysis of TMI-2 melted fuel debris. The generation of this report is motivated by the need to perform nuclear material accountancy measurements on the melted fuel debris that will be excavated from the damaged nuclear reactors at the Fukushima Daiichi nuclear power plant in Japan, which were destroyed by the Tohoku earthquake and tsunami on March 11, 2011. Lessons may be taken from prior U.S. work related to the study of the TMI-2 core debris to support the development of new assay methods for use at Fukushima Daiichi. While significant differences exist between the two reactor systems (pressurized water reactor (TMI-2) versus boiling water reactor (FD), fresh water post-accident cooing (TMI-2) versus salt water (FD), maintained containment (TMI-2) versus loss of containment (FD)) there remain sufficient similarities to motivate these comparisons.

  5. R-matrix analysis of the {sup 240}Pu neutron cross sections in the thermal to 5700 eV energy range

    SciTech Connect (OSTI)

    Derrien, H. [OECD, Paris (France). Nuclear Energy Agency Data Bank; Bouland, O. [Commissariat Energie Atomique, Saint Paul-lez-Durance (France). Centre d`Etudes; Larson, N.M.; Leal, L.C. [Oak Ridge National Lab., TN (United States)

    1997-08-01T23:59:59.000Z

    Resonance analysis of high resolution neutron transmission data and of fission cross sections were performed in the neutron energy range from the thermal regions to 5,700 eV by using the Reich-Moore Bayesian code SAMMY. The experimental data base is described and the method of analysis is given. The experimental data were carefully examined in order to identify more resonances than those found in the current evaluated data files. The statistical properties of the resonance parameters are given. A new set of the average values of the parameters is proposed, which could be used for calculation of the average cross sections in the unresolved resonance region. The resonance parameters are available IN ENDF-6 format at the national or international data centers.

  6. Preliminary Simulations for Geometric Optimization of a High-Energy Delayed Gamma Spectrometer for Direct Assay of Pu in Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Kulisek, Jonathan A.; Campbell, Luke W.; Rodriguez, Douglas C.

    2012-06-07T23:59:59.000Z

    High-energy, beta-delayed gamma-ray spectroscopy is under investigation as part of the Next Generation Safeguard Initiative effort to develop non-destructive assay instruments for plutonium mass quantification in spent nuclear fuel assemblies. Results obtained to date indicate that individual isotope-specific signatures contained in the delayed gamma-ray spectra can potentially be used to quantify the total fissile content and individual weight fractions of fissile and fertile nuclides present in spent fuel. Adequate assay precision for inventory analysis can be obtained using a neutron generator of sufficient strength and currently available detection technology. In an attempt to optimize the geometric configuration and material composition for a delayed gamma measurement on spent fuel, the current study applies MCNPX, a Monte Carlo radiation transport code, in order to obtain the best signal-to-noise ratio. Results are presented for optimizing the neutron spectrum tailoring material, geometries to maximize thermal or fast fissions from a given neutron source, and detector location to allow an acceptable delayed gamma-ray signal while achieving a reasonable detector lifetime while operating in a high-energy neutron field. This work is supported in part by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.

  7. I. Nuclear Production Reaction and Chemical Isolation Procedure for 240Am II. New Superheavy Element Isotopes: 242Pu(48Ca,5n)285-114

    E-Print Network [OSTI]

    Ellison, Paul Andrew

    2011-01-01T23:59:59.000Z

    of superheavy nuclei in cold fusion reactions. Phys. Rev. C,transfermium elements in cold fusion reactions. Phys. Rev.have been deemed “cold fusion” reactions because of the low

  8. I. Nuclear Production Reaction and Chemical Isolation Procedure for 240Am II. New Superheavy Element Isotopes: 242Pu(48Ca,5n)285-114

    E-Print Network [OSTI]

    Ellison, Paul Andrew

    2011-01-01T23:59:59.000Z

    TABLES MBS MCA MWPC n NIM NNSA ORNL p PIPS PTD Q alpha ROIxi Acknowledgments ministration (NNSA), Stewardship Sciencefunding came through a NNSA Stewardship Science Graduate

  9. I. Nuclear Production Reaction and Chemical Isolation Procedure for 240Am II. New Superheavy Element Isotopes: 242Pu(48Ca,5n)285-114

    E-Print Network [OSTI]

    Ellison, Paul Andrew

    2011-01-01T23:59:59.000Z

    library for nuclear science and technology. Nuclear DataJournal of Nuclear Science and Technology, 7(10):487–499,Journal of Nuclear Science and Technology, G.T. Seaborg,

  10. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    SciTech Connect (OSTI)

    NONE

    1993-09-15T23:59:59.000Z

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  11. Solo la crescita pu salvare un'Italia logorata e proteggere l'Euro Ulrike Sauer, Sddeutsche Zeitung 3/2/2011

    E-Print Network [OSTI]

    Angeleri Hügel, Lidia

    la torre di Pisa che pende e che pende ma non cade mai" dice la caricatura sul capo di governo invece il governo non si occupa delle condizioni in cui si trovano ad operare le imprese. Né la'Italia che la sera va a letto presto". Questo paese dimenticato sta aspettando da anni un nuovo governo che

  12. Analytical Capability of Plasma Spectrometry Team

    SciTech Connect (OSTI)

    Gallimore, David L. [Los Alamos National Laboratory

    2012-07-19T23:59:59.000Z

    Samples analyzed were: (1) Pu and U metal; (2) Pu oxide for nuclear fuel; (3) {sup 238}Pu oxide for heat source; and (4) Nuclear forensic samples - filters, swipes. Sample preparations that we did were: metal dissolution, marple filter dissolution, Pu oxide closed vessel acid digestion, and column separation to remove Pu.

  13. PONTIFICIA UNIVERSIDAD CATOLICA DE CHILE ESCUELA DE INGENIERIA

    E-Print Network [OSTI]

    Rudnick, Hugh

    conexión consume 7 MVAr y 2 MW. Los parámetros están dados en base 20 MVA. Se entregan las impedancias de parámetros en base común: G1: x1= 0,20 pu x2= 0,17 pu xo= 0,32 p.u G2: x1= 0,16 pu x2= 0,13 pu xo= 0,27 p.u T

  14. Qualification Plan for Phase One of True-MidPacific Geothermal Venture: James Campbell - Kahaualea Project, Island of Hawaii

    SciTech Connect (OSTI)

    None

    1981-06-01T23:59:59.000Z

    The objective of this project is to develop the geothermal resources of the James Campbell Estate, comprising acres in the Puna District of the Island of Hawaii. The geothermal resource is assumed to exist in the vicinity of the East Rift of the Kilauea volcano. The location of the proposed geothermal well field and the geothermal-electric power plant are shown on Dwg. No. E-04-001. Access to the project area will be provided by a new road extension from the boundary road south from Glenwood on Highway 11.

  15. Parallel Tracks: American Transcontinentalism and the Specter of Canada

    E-Print Network [OSTI]

    Eigen, Kathryn

    2010-01-01T23:59:59.000Z

    man, have great cause to grieve over the loss of Pu-pu- mox-mox." 19 In other words, for enthusiastic Far Western

  16. SUBSURFACE MOBILE PLUTONIUM SPECIATION: SAMPLING ARTIFACTS FOR GROUNDWATER COLLOIDS

    SciTech Connect (OSTI)

    Kaplan, D.; Buesseler, K.

    2010-06-29T23:59:59.000Z

    A recent review found several conflicting conclusions regarding colloid-facilitated transport of radionuclides in groundwater and noted that colloids can both facilitate and retard transport. Given these contrasting conclusions and the profound implications even trace concentrations of plutonium (Pu) have on the calculated risk posed to human health, it is important that the methodology used to sample groundwater colloids be free of artifacts. The objective of this study was: (1) to conduct a field study and measure Pu speciation, ({sup 239}Pu and {sup 240}Pu for reduced-Pu{sub aq}, oxidized-Pu{sub aq}, reduced-Pu{sub colloid}, and oxidized-Pu{sub colloid}), in a Savannah River Site (SRS) aquifer along a pH gradient in F-Area, (2) to determine the impact of pumping rate on Pu concentration, Pu speciation, and Pu isotopic ratios, (3) determine the impact of delayed sample processing (as opposed to processing directly from the well).

  17. Lattice relaxation and the stability of plutonium-based alloys and intermetallics

    SciTech Connect (OSTI)

    Becker, J.D.; Cox, L.; Wills, J.M. [Los Alamos National Lab., NM (United States); Cooper, B.R. [West Virginia Univ., Morgantown, WV (United States)

    1996-12-01T23:59:59.000Z

    The topic of this study is the electronic structure of the compounds Pu{sub 3}X [X = In, Ga, Al, or Tl], reported to have room temperature L1{sub 2} structures. The measured Pu-Pu bond length in {delta}-phase Pu is 3.28 {Angstrom}. The Pu-Pu bond lengths in Pu{sub 3}Ga and Pu{sub 3}Al are 3.18 {Angstrom}. The similarity in bonding in Pu{sub 3}Ga and Pu{sub 3}Al is especially notable as the equilibrium atomic volume in Ga is 18% higher than that of Al. Likewise, in Pu{sub 3}In and Pu{sub 3}Tl the bond lengths are 3.33 {Angstrom} and 3.32 {Angstrom}, respectively, although the atomic volumes of In and Tl differ by 10%. The electronic mechanism by which the 8 stabilizers Al and Ga contract the Pu-Pu bonds and whereby In and Tl stretch them may relate the stability of pure Pu to that of the Pu{sub 3}X compounds. For example, does the addition of Ga or Al shorten the bond length enough to make the close-packed Ll{sub 2} structure more stable than does the addition of In or Tl? The incipient localization of the f electrons and the stabilization of open structures may be indicated through LDA total energies and densities of states.

  18. Radiation-induced non-equilibrium redox chemistry of plutonium: implications for environmental migration

    SciTech Connect (OSTI)

    Haschke, J M; Siekhaus, W J

    2009-02-11T23:59:59.000Z

    Static concentrations of plutonium oxidation states in solution and at surfaces in oxide-water systems are identified as non-equilibrium steady states. These kinetically controlled systems are described by redox cycles based on irreversible disproportionation of Pu(IV), Pu(V), and Pu(VI) in OH-bridged intermediate complexes and at OH-covered oxide surfaces. Steady state is fixed by continuous redox cycles driven by radioactivity-promoted electron-transfer and energetically favorable reactions of Pu(III) and Pu(VII) disproportionation products with H2O. A model based on the redox cycles accounts for the high steady-state [Pu] coexisting with Pu(IV) hydrous oxide at pH 0-15 and for predominance of Pu(V) and Pu(VI) in solution. The steady-state [Pu] depends on pH and the surface area of oxide in solution, but not on the initial Pu oxidation state. PuO{sub 2+x} formation is attributed to high Pu(V) concentrations existing at water-exposed oxide surfaces. Results infer that migration of Pu in an aqueous environment is controlled by kinetic factors unique to that site and that the predominant oxidation states in solution are Pu(V) and Pu(VI).

  19. Corso di Laurea in Matematica Fisica Matematica Parte I (2012)

    E-Print Network [OSTI]

    Fassò, Francesco

    2012-01-01T23:59:59.000Z

    campi vettoriali e riparametrizzazioni temporali? Fare un enunciato pre- ciso e dimostrarlo. · Pu`o un

  20. MECCANICA DEI FLUIDI 0. Notazioni

    E-Print Network [OSTI]

    Bartocci, Claudio

    le due leggi della termodinamica, come si pu`o comunque verificare esplicitamente. Il bilancio della

  1. Experimental studies of actinide volatilities with application to mixed waste oxidation processors

    SciTech Connect (OSTI)

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

    1993-04-30T23:59:59.000Z

    The transpiration technique is used to measure volatilities of U from U{sub 3}O{sub 8}(s), Pu from PuO{sub 2}(s) and Pu and Am from PuO{sub 2}/2%AmO{sub 2}(s) in the presence of steam and oxygen at temperatures ranging from 900 to 1300{degree}C.

  2. Cefas contract report C5975 (MLA/2013/00269) Radiological Assessment of

    E-Print Network [OSTI]

    detected by gamma spectrometry, sediments are also known to contain activities of Pu radionuclides. The 241

  3. SUSCEPTIBILIT MAGNTIQUE DE QUELQUES SULFURES ET OXYDES DE PLUTONIUM

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    261. SUSCEPTIBILITÉ MAGNÉTIQUE DE QUELQUES SULFURES ET OXYDES DE PLUTONIUM Par GEORGES RAPHAEL et CHARLES DE NOVION, S.E.C.P.E.R., Section d'Études des Céramiques à base de Plutonium, Centre d susceptibilite magnétique des sulfures de plutonium : PuS, Pu3S4, PU2S3CXI PuS2. Ces composes non conduc- teurs

  4. surface si faible qu'elle n'a pu encore tre apprcie. Il faut en conclure, dans l'hypothse de Volta, que le rayon de la sphrc

    E-Print Network [OSTI]

    Boyer, Edmond

    - cien de Bonn croie nécessaire, pour l'explication du phénomène de Peltier, qu'une fraction (en tout cas; PAR M. A. ROSENSTIEHL. La question qui doit être résolue dans cette Note est celle de savoir si le

  5. Journal 9 : lundi 2 avril 2012 (dernier journal du 1er Pendant le transit d'hier entre la radiale 7W et Abidjan, nous avons pu encore retrouver nos

    E-Print Network [OSTI]

    navire ANTEA de l'IRD était basé ici pendant plusieurs années jusqu'en 2002 (avant son rapatriement en France suite à de gros problèmes techniques immobilisant le navire depuis 2000...) ! L'actuel commandant campagnes PIRATA faites avec l'ANTEA de 1997 à 1999 ! Tous les objectifs de cette première partie de la

  6. Investigation on the coprecipitation of transuranium elements from alkaline solutions by the method of appearing reagents. Study of the effects of waste components on decontamination from Np(IV) and Pu(IV)

    SciTech Connect (OSTI)

    Bessonov, A.A.; Budantseva, N.A.; Gelis, A.V.; Nikonov, M.V.; Shilov, V.P. [Russian Academy of Sciences, Moscow (Russian Federation). Institute of Physical Chemistry

    1997-09-01T23:59:59.000Z

    The third stage of the study on the homogeneous coprecipitation of neptunium and plutonium from alkaline high-level radioactive waste solutions by the Method of Appearing Reagents has been completed. Alkaline radioactive wastes exist at the U.S. Department of Energy Hanford Site. The recent studies investigated the effects of neptunium chemical reductants, plutonium(IV) concentration, and the presence of bulk tank waste solution components on the decontamination from tetravalent neptunium and plutonium achieved by homogeneous coprecipitation. Data on neptunium reduction to its tetravalent state in alkaline solution of different NaOH concentrations are given. Eleven reductants were tested to find those most suited to remove neptunium, through chemical reduction, from alkaline solution by homogeneous coprecipitation. Hydrazine, VOSO{sub 4}, and Na{sub 2}S{sub 2}O{sub 4} were found to be the most effective reductants. The rates of reduction with these reductants were comparable with the kinetics of carrier formation. Solution decontamination factors of about 400 were attained for 10{sup -6}M neptunium. Coprecipitation of plutonium(IV) with carriers obtained as products of thermal hydrolysis, redox transformations, and catalytic decomposition of [Co(NH{sub 3}){sub 6}]{sup 3+}, [Fe(CN){sub 5}NO]{sup 2-}, Cr(NO{sub 3}){sub 3}, KMnO{sub 4}, and Li{sub 4}UO{sub 2}(O{sub 2}){sub 3} was studied and results are described. Under optimum conditions, a 100-fold decrease of plutonium concentration was possible with each of these reagents.

  7. nature CHeMICaL BIOLOGY | vol 10 | april 2014 | www.nature.com/naturechemicalbiology 273 puBLIsHed OnLIne: 23 feBruarY 2014 | dOI: 10.1038/nCHeMBIO.1458

    E-Print Network [OSTI]

    Cai, Long

    driving force for pore opening. Desensitization has been accounted for by structural rearrangements on the functional state of a receptor and their cooperative interactions are often difficult to assess because and structural studies2,8,9 . The four ligand-binding domains (LBDs) are organized as a pair of dimers. Ligands

  8. The Magazine of The Johns hopkins BlooMBerg school of puBlic healTh www.jhsph.eduspecial death and data death and lifespan death and learning

    E-Print Network [OSTI]

    Scharfstein, Daniel

    death and data · death and lifespan · death and learning the poetry of life "Because i could not stoP for deatH" #12;MAKE A PROMISE FOR THE FUTURE Fund a Charitable Gift Annuity with a minimum gift of $10 contribution to the Johns Hopkins Bloomberg School of Public Health. Charitable Gift Annuity Rates

  9. Sorption Behavior and Morphology of Plutonium in the Presence of Goethite at 25 and 80C

    SciTech Connect (OSTI)

    Zavarin, M; Zhao, P; Dai, Z; Carroll, S A; Kersting, A B

    2012-06-11T23:59:59.000Z

    In this study, we examined the sorption behavior of Pu at elevated temperatures in the presence of one relevant mineral, goethite ({alpha}-FeOOH), over a range of concentrations that span solubility-controlled to adsorption-controlled concentrations. We focused on the sorptive behavior of two common forms of Pu: aqueous Pu(IV) and intrinsic Pu(IV) nano-colloids at 25 and 80 C in a dilute pH 8 NaCl/NaHCO{sub 3} solution. The morphology of Pu sorbed to goethite was characterized using transmission electron microscopy (TEM). We examined the relative stability of PuO{sub 2} precipitates, PuO{sub 2} nano-colloids, Pu{sub 4}O{sub 7} surface precipitates, and monomeric sorbed Pu as a function of temperature and over a time scale of months.

  10. PONTIFICIA UNIVERSIDAD CATOLICA DE CHILE ESCUELA DE INGENIERIA

    E-Print Network [OSTI]

    Rudnick, Hugh

    interconectado de tensión 0,96 pu. Su turbina le permite entregar una potencia activa máxima de 1,05 pu. La el generador simultáneamente con entregar potencia activa máxima. Si se puede sobrecargar la turbina

  11. ACTIVITY COORDINATOR HOST LOCATION 1.1. Observation session for management level European partners UA Spain

    E-Print Network [OSTI]

    UA Spain 1.2. Observation session for Staff level European partners UPMF France 2.1. Development.3. Management meetings UA, PU UA, PU Spain, Jordan 4. Networking and sustainability (network) 3. Capacity

  12. EA-0534: Radioisotope Heat Source Fuel Processing and Fabrication, Los Alamos, New Mexico

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of a proposal to operate existing Pu-238 processing facilities at Savannah River Site, and fabricate a limited quantity of Pu-238 fueled heat sources at...

  13. Electronic structure and ionicity of actinide oxides from first principles L. Petit,1,2,* A. Svane,1 Z. Szotek,2 W. M. Temmerman,2 and G. M. Stocks3

    E-Print Network [OSTI]

    Svane, Axel Torstein

    . A mixture of UO2 and PuO2, where Pu is blended with either natural or depleted uranium, constitutes. INTRODUCTION Actinide oxides play a dominant role in the nuclear fuel cycle.1 For many years, uranium dioxide

  14. SCUOLA NORMALE SUPERIORE DI PISA Laboratorio di Storia, Archeologia e Topografia del Mondo Antico

    E-Print Network [OSTI]

    Abbondandolo, Alberto

    anfore, non può fornire altro che ipotesi di lavoro. Infatti, la perdita del vecchio numero d'inventario

  15. Tri-Gate Bulk CMOS Technology for Improved SRAM Scalability Changhwan Shin, Borivoje Nikoli, Tsu-Jae King Liu

    E-Print Network [OSTI]

    Nikolic, Borivoje

    ] is an example of such a design; it utilizes a gate electrode that is physically wrapped around the top portion along a poly-Si gate electrode in an SRAM array, for 15nm nominal STI recess depth. The sequence), pull- 570nm 263nm 570nm 263nm PG1 PD1 PU1 PU2 PD2 PG2 (a) 570nm 263nm 570nm 263nm PG1 PD1 PU1 PU2 PD2

  16. Universit degli Studi di Roma "La Sapienza" Presidenza della Facolt di Scienze MFN

    E-Print Network [OSTI]

    Guidoni, Leonardo

    Termodinamica e laboratorio caratterizzante FIS/01 9 90 II La supplenza/affidamento può essere conferita

  17. South American Journal of Herpetology, 2(3), 2007, 201-205 2007 Brazilian Society of Herpetology

    E-Print Network [OSTI]

    Bermingham, Eldredge

    commission of INGEOMINAS (Instituto Colombiano de Geología y Minería) Pu- benza locality, Department

  18. Nuclear Physics A559 (1993) l-41 ~ortb-~ojland

    E-Print Network [OSTI]

    Ponomarev, Vladimir

    Institute for Nucleur Research, Dubna, Head Post Cl&e, P.U. Box 79, Moscow, Russian Federation Received

  19. Collective excitations of transactinide nuclei in a self-consistent mean field theory

    E-Print Network [OSTI]

    L. Prochniak

    2007-12-03T23:59:59.000Z

    The ATDHFB approach has been applied for a study of properties of collective quadrupole states in several transactinide nuclei: 238U, 240Pu, 242Pu, 246Cm, 248Cm, 250Cf and 252Cf. Calculated energies and B(E2) transition probabilities are in a reasonable agreement with experimental data. Results concerning superdeformed collective states in the 240Pu nucleus are also presented.

  20. Natural and anthropogenic radionuclides in the marginal seas of Siberia: implications for the fate and removal of pollutants

    E-Print Network [OSTI]

    Schwantes, Jon Michael

    1996-01-01T23:59:59.000Z

    understand scavenging within the water column. To elucidate the sources of Pu and 131CS to the study area, Pu and "'Cs concentrations and 238PU/239,240puactivity ratios were measured in water samples. Concentrations of Ra isotopes were also determined...

  1. FY12 Final Report for PL10-Mod Separations-PD12: Electrochemically Modulated Separation of Plutonium from Dilute and Concentrated Dissolver Solutions for Analysis by Gamma Spectroscopy

    SciTech Connect (OSTI)

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

    2013-05-01T23:59:59.000Z

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

  2. Electrochemically Modulated Separation for Plutonium Safeguards

    SciTech Connect (OSTI)

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

    2013-12-31T23:59:59.000Z

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

  3. Coordination and Hydrolysis of Plutonium Ions in Aqueous Solution using Car-Parrinello Molecular Dynamics Free Energy Simulations

    SciTech Connect (OSTI)

    Odoh, Samuel O.; Bylaska, Eric J.; De Jong, Wibe A.

    2013-11-27T23:59:59.000Z

    Car-Parrinello molecular dynamics (CPMD) simulations have been used to examine the hydration structures, coordination energetics and the first hydrolysis constants of Pu3+, Pu4+, PuO2+ and PuO22+ ions in aqueous solution at 300 K. The coordination numbers and structural properties of the first shell of these ions are in good agreement with available experimental estimates. The hexavalent PuO22+ species is coordinated to 5 aquo ligands while the pentavalent PuO2+ complex is coordinated to 4 aquo ligands. The Pu3+ and Pu4+ ions are both coordinated to 8 water molecules. The first hydrolysis constants obtained for Pu3+ and PuO22+ are 6.65 and 5.70 respectively, all within 0.3 pH units of the experimental values (6.90 and 5.50 respectively). The hydrolysis constant of Pu4+, 0.17, disagrees with the value of -0.60 in the most recent update of the Nuclear Energy Agency Thermochemical Database (NEA-TDB) but supports recent experimental findings. The hydrolysis constant of PuO2+, 9.51, supports the experimental results of Bennett et al. (Radiochim. Act. 1992, 56, 15). A correlation between the pKa of the first hydrolysis reaction and the effective charge of the plutonium center was found.

  4. An investigation into the spectral evolution of turbulent mixing by Rayleigh-Taylor Instability

    E-Print Network [OSTI]

    Wilson, Peter Nixon

    1998-01-01T23:59:59.000Z

    ?dx ?dx, dx, (2-2) The mass-weighted average of the velocity vector separates as u, = u, + u, ", defined as, p" i u p (2-3) and the relationship, pu, " = 0, is shown from the following: 14 pu, " = p(u, ? u, ) = piu, ? =' = pu, ? pu, = 0. p (2-4... TABLE OF CONTENTS, LIST OF FIGURES . . LIST OF TABLES. . xn1 NOMENCLATURE . 1. INTRODUCTION. xtv 1. 1 Description of Turbulence. 1. 2 Motivation for Research. 1. 3 Previous Investigations Into Vanable-Density Turbulence. . . . 1. 4 Spectral...

  5. Plutonium and minor actinides utilization in Thorium molten salt reactor

    SciTech Connect (OSTI)

    Waris, Abdul; Aji, Indarta K.; Novitrian,; Kurniadi, Rizal; Su'ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jalan Ganesa 10 Bandung 40132 (Indonesia)

    2012-06-06T23:59:59.000Z

    FUJI-12 reactor is one of MSR systems that proposed by Japan. The original FUJI-12 design considers Th/{sup 233}U or Th/Pu as main fuel. In accordance with the currently suggestion to stay away from the separation of Pu and minor actinides (MA), in this study we evaluated the utilization of Pu and MA in FUJI-12. The reactor grade Pu was employed in the present study as a small effort of supporting THORIMS-NES scenario. The result shows that the reactor can achieve its criticality with the Pu and MA composition in the fuel of 5.96% or more.

  6. Criteria for safe storage of plutonium metals and oxides

    SciTech Connect (OSTI)

    Not Available

    1994-12-01T23:59:59.000Z

    This standard establishes safety criteria for safe storage of plutonium metals and plutonium oxides at DOE facilities; materials packaged to meet these criteria should not need subsequent repackaging to ensure safe storage for at least 50 years or until final disposition. The standard applied to Pu metals, selected alloys (eg., Ga and Al alloys), and stabilized oxides containing at least 50 wt % Pu; it does not apply to Pu-bearing liquids, process residues, waste, sealed weapon components, or material containing more than 3 wt % {sup 238}Pu. Requirements for a Pu storage facility and safeguards and security considerations are not stressed as they are addressed in detail by other DOE orders.

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

    SciTech Connect (OSTI)

    Nakahara, Masaumi; Koma, Yoshikazu; Nakajima, Yasuo [Japan Atomic Energy Agency: 4-33 Muramatsu, Naka-gun, Tokai-mura, Ibaraki 319-1194 (Japan)

    2013-07-01T23:59:59.000Z

    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)

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

    SciTech Connect (OSTI)

    G. Youinou; S. Bays

    2009-05-01T23:59:59.000Z

    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).

  9. Plutonium Metallurgy

    SciTech Connect (OSTI)

    Freibert, Franz J. [Los Alamos National Laboratory

    2012-08-09T23:59:59.000Z

    Due to its nuclear properties, Pu will remain a material of global interest well into the future. Processing, Structure, Properties and Performance remains a good framework for discussion of Pu materials science Self-irradiation and aging effects continue to be central in discussions of Pu metallurgy Pu in its elemental form is extremely unstable, but alloying helps to stabilize Pu; but, questions remain as to how and why this stabilization occurs. Which is true Pu-Ga binary phase diagram: US or Russian? Metallurgical issues such as solute coring, phase instability, crystallographic texture, etc. result in challenges to casting, processing, and properties modeling and experiments. For Ga alloyed FCC stabilized Pu, temperature and pressure remain as variables impacting phase stability.

  10. PLUTONIUM SOLUBILITY IN HIGH-LEVEL WASTE ALKALI BOROSILICATE GLASS

    SciTech Connect (OSTI)

    Marra, J.; Crawford, C.; Fox, K.; Bibler, N.

    2011-01-04T23:59:59.000Z

    The solubility of plutonium in a Sludge Batch 6 (SB6) reference glass and the effect of incorporation of Pu in the glass on specific glass properties were evaluated. A Pu loading of 1 wt % in glass was studied. Prior to actual plutonium glass testing, surrogate testing (using Hf as a surrogate for Pu) was conducted to evaluate the homogeneity of significant quantities of Hf (Pu) in the glass, determine the most appropriate methods to evaluate homogeneity for Pu glass testing, and to evaluate the impact of Hf loading in the glass on select glass properties. Surrogate testing was conducted using Hf to represent between 0 and 1 wt % Pu in glass on an equivalent molar basis. A Pu loading of 1 wt % in glass translated to {approx}18 kg Pu per Defense Waste Processing Facility (DWPF) canister, or about 10X the current allowed limit per the Waste Acceptance Product Specifications (2500 g/m{sup 3} of glass or about 1700 g/canister) and about 30X the current allowable concentration based on the fissile material concentration limit referenced in the Yucca Mountain Project License Application (897 g/m{sup 3}3 of glass or about 600 g Pu/canister). Based on historical process throughput data, this level was considered to represent a reasonable upper bound for Pu loading based on the ability to provide Pu containing feed to the DWPF. The task elements included evaluating the distribution of Pu in the glass (e.g. homogeneity), evaluating crystallization within the glass, evaluating select glass properties (with surrogates), and evaluating durability using the Product Consistency Test -- Method A (PCT-A). The behavior of Pu in the melter was evaluated using paper studies and corresponding analyses of DWPF melter pour samples.The results of the testing indicated that at 1 wt % Pu in the glass, the Pu was homogeneously distributed and did not result in any formation of plutonium-containing crystalline phases as long as the glass was prepared under 'well-mixed' conditions. The incorporation of 1 wt % Pu in the glass did not adversely impact glass viscosity (as assessed using Hf surrogate) or glass durability. Finally, evaluation of DWPF glass pour samples that had Pu concentrations below the 897 g/m{sup 3} limit showed that Pu concentrations in the glass pour stream were close to targeted compositions in the melter feed indicating that Pu neither volatilized from the melt nor stratified in the melter when processed in the DWPF melter.

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

    SciTech Connect (OSTI)

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

    2009-01-01T23:59:59.000Z

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

  12. Wind resuspension of trace amounts of plutonium particles from soil in a semi-arid climate

    SciTech Connect (OSTI)

    Langer, G.

    1984-01-01T23:59:59.000Z

    This study of resuspension of soil containing minute amounts of plutonium (Pu-239) has been in progress at the Rocky Flats (RF) Plant since 1978. It is one of several studies initiated after wind relocated small amounts of soil-borne Pu-239 during cleanup of an outdoor storage area. The Pu-239-settled field is now sparsely covered with prairie grass typical of the area. Past studies were limited to comparisons of bulk soil activity with total activity in the airborne dust. This work covers the physics of the particle resuspension process. This report covers the following: (1) Pu-239 resuspension rate versus wind speed, (2) mechanisms of soil particle resuspension, (3) vertical concentration profile of Pu-239 particles, (4) Pu-239 and host particle size distribution and activity concentration. 5 references, 1 table.

  13. Standard test method for the determination of impurities in plutonium metal: acid digestion and inductively coupled plasma-mass spectroscopy (ICP-MS) analysis

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2006-01-01T23:59:59.000Z

    1.1 This Test Method covers the determination of 58 trace elements in plutonium (Pu) metal. The Pu sample is dissolved in acid, and the concentration of the trace impurities are determined by Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS). 1.2 This Test Method is specific for the determination of trace impurities in Pu metal. It may be applied to other types of Pu materials, such as Pu oxides, if the samples are dissolved and oxidized to the Pu(IV) state. However, it is the responsibility of the user to evaluate the performance of other matrices. 1.3 This standard does not purport to address all of the safety concerns associated with its use. It is the responsibility of the user of this method to establish appropriate safety and health practices and to determine the applicability of regulatory limitations prior to use of this standard.

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

    DOE Patents [OSTI]

    Crawford, Douglas C. (Idaho Falls, ID); Porter, Douglas L. (Idaho Falls, ID); Hayes, Steven L. (Idaho Falls, ID); Hill, Robert N. (Bolingbrook, IL)

    1999-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1997-12-01T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

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

    1999-03-23T23:59:59.000Z

    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.

  17. X-ray absorption fine structure spectroscopic determination of plutonium speciation at the Rocky Flats environmental technology

    SciTech Connect (OSTI)

    Lezama-pacheco, Juan S [Los Alamos National Laboratory; Conradson, Steven D [Los Alamos National Laboratory; Clark, David L [Los Alamos National Laboratory

    2008-01-01T23:59:59.000Z

    X-ray Absorption Fine Structure spectroscopy was used to probe the speciation of the ppm level Pu in thirteen soil and concrete samples from the Rocky Flats Environmental Technology Site in support of the site remediation effort that has been successfully completed since these measurements. In addition to X-ray Absorption Near Edge Spectra, two of the samples yielded Extended X-ray Absorption Fine Structure spectra that could be analyzed by curve-fits. Most of these spectra exhibited features consistent with PU(IV), and more specificaJly, PuO{sub 2+x}-type speciation. Two were ambiguous, possibly indicating that Pu that was originally present in a different form was transforming into PuO{sub 2+x}, and one was interpreted as demonstrating the presence of an unusual Pu(VI) compound, consistent with its source being spills from a PUREX purification line onto a concrete floor and the resultant extreme conditions. These experimental results therefore validated models that predicted that insoluble PuO{sub 2+x} would be the most stable form of Pu in equilibrium with air and water even when the source terms were most likely Pu metal with organic compounds or a Pu fire. A corollary of these models' predictions and other in situ observations is therefore that the minimal transport of Pu that occurred on the site was via the resuspension and mobilization of colloidal particles. Under these conditions, the small amounts of diffusely distributed Pu that were left on the site after its remediation pose only a negligible hazard.

  18. Evaluation and compilation of fission product yields 1993

    SciTech Connect (OSTI)

    England, T.R.; Rider, B.F.

    1995-12-31T23:59:59.000Z

    This document is the latest in a series of compilations of fission yield data. Fission yield measurements reported in the open literature and calculated charge distributions have been used to produce a recommended set of yields for the fission products. The original data with reference sources, and the recommended yields axe presented in tabular form. These include many nuclides which fission by neutrons at several energies. These energies include thermal energies (T), fission spectrum energies (F), 14 meV High Energy (H or HE), and spontaneous fission (S), in six sets of ten each. Set A includes U235T, U235F, U235HE, U238F, U238HE, Pu239T, Pu239F, Pu241T, U233T, Th232F. Set B includes U233F, U233HE, U236F, Pu239H, Pu240F, Pu241F, Pu242F, Th232H, Np237F, Cf252S. Set C includes U234F, U237F, Pu240H, U234HE, U236HE, Pu238F, Am241F, Am243F, Np238F, Cm242F. Set D includes Th227T, Th229T, Pa231F, Am241T, Am241H, Am242MT, Cm245T, Cf249T, Cf251T, Es254T. Set E includes Cf250S, Cm244S, Cm248S, Es253S, Fm254S, Fm255T, Fm256S, Np237H, U232T, U238S. Set F includes Cm243T, Cm246S, Cm243F, Cm244F, Cm246F, Cm248F, Pu242H, Np237T, Pu240T, and Pu242T to complete fission product yield evaluations for 60 fissioning systems in all. This report also serves as the primary documentation for the second evaluation of yields in ENDF/B-VI released in 1993.

  19. Method of immobilizing weapons plutonium to provide a durable, disposable waste product

    DOE Patents [OSTI]

    Ewing, Rodney C. (Albuquerque, NM); Lutze, Werner (Albuquerque, NM); Weber, William J. (Richland, WA)

    1996-01-01T23:59:59.000Z

    A method of atomic scale fixation and immobilization of plutonium to provide a durable waste product. Plutonium is provided in the form of either PuO.sub.2 or Pu(NO.sub.3).sub.4 and is mixed with and SiO.sub.2. The resulting mixture is cold pressed and then heated under pressure to form (Zr,Pu)SiO.sub.4 as the waste product.

  20. 30 IE E E S o f t wa r E | pu blI S hE d by thE IE E E c o mpu tE r S o c IE t y 074 0 -74 5 9/12 / $ 31.0 0 2 012 I E E E cover image here

    E-Print Network [OSTI]

    Xie, Tao

    . Effective use of criteria such as structural-code cov- erage can help reveal faults. Manually writing test the manual effort, they can employ auto- mated test generation tools that use dynamic symbolic execution (DSE cloud envi- ronment with a fake stub that provides default or user-defined return values to cloud

  1. alarm border monitoring: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    alarm). Pu Plutonium. ROC Receiver operating characteristic. RPM Radiation portal monitor. WGPu Weapons grade plutonium. I. INTRODUCTION Parasuraman, Raja 54 Fire Alarm...

  2. Computer News, Volume 32

    E-Print Network [OSTI]

    MATH DEPT Computer News, Volume 32. The new TeX and how to create the new PU Math letterhead stationery. with help from Brad Lucier, Rodrigo Bañuelos

  3. Spectrum Sharing in Cognitive Radio Systems Under Outage Probablility Constraint

    E-Print Network [OSTI]

    Cai, Pei Li

    2011-02-22T23:59:59.000Z

    (SU) link with multiple transmitting an- tennas and a single receiving antenna, coexisting with a primary user (PU) link with a single receiving antenna. At the SU transmitter (SU-Tx), the channel state infor- mation (CSI) of the SU link is assumed... to be perfectly known; while the interference channel from the SU-Tx to the PU receiver (PU-Rx) is not perfectly known due to less cooperation between the SU and the PU. As such, the SU-Tx is only assumed to know that the interference channel gain can take values...

  4. A Close in Place Option for Buried Transuranic Waste at the Nevada...

    National Nuclear Security Administration (NNSA)

    and the number of boreholes per intrusion event. The Pu-239 TRU inventory shows a non-linear rapidly increasing relationship with the cumulative release. The marginal dependence...

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

    SciTech Connect (OSTI)

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

    2012-07-26T23:59:59.000Z

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

  6. analytique du plutonium: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    to workers conducting planned plutonium (Pu) stabilization processes at the Hanford Site Plutonium Finishing Plant (PFP). The report is based on a time and motion dose study...

  7. Influence of Iron Redox Transformations on Plutonium Sorption...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    and Pu(V) reduction demonstrates the potential impact of changing iron mineralogy on plutonium subsurface transport through redox transition areas. These findings...

  8. Enduring Stockpile CMM Shell Inspection Plan (U)

    SciTech Connect (OSTI)

    Montano, Joshua D. [Los Alamos National Laboratory; Flores, Randy A. [Los Alamos National Laboratory

    2012-06-13T23:59:59.000Z

    The slides are intended to serve as a high level summary of the CMM Shell Inspection Plan as presented to Pu Sustainment Legacy Pit Production IPT.

  9. Accident Investigation Report Plutonium Contamination in the...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    one of the storage containers, the workers discovered a Pu fuel plate wrapped in plastic and tape. When the workers attempted to remove the wrapping material, an uncontrolled...

  10. Plutonium(IV) precipitates formed in alkaline media in the presence of various anions

    SciTech Connect (OSTI)

    Krot, N.N.; Shilov, V.P.; Yusov, A.B.; Tananaev, I.G.; Grigoriev, M.S.; Garnov, A.Yu.; Perminov, V.P.; Astafurova, L.N.

    1998-09-01T23:59:59.000Z

    The tendency of Pu(IV) to hydrolyze and form true solutions, colloid solutions, or insoluble precipitates has been known since the Manhattan Project. Since then, specific studies have been performed to examine in detail the equilibria of Pu(IV) hydrolytic reactions in various media. Great attention also has been paid to the preparation, structure, and properties of Pu(IV) polymers or colloids. These compounds found an important application in sol-gel technology for the preparation of nuclear fuel materials. A most important result of these works was the conclusion that Pu(IV) hydroxide, after some aging, consists of very small PuO{sub 2} crystallites and should therefore be considered to be Pu(IV) hydrous oxide. However, studies of the properties and behavior of solid Pu(IV) hydroxide in complex heterogeneous systems are rare. The primary goal of this investigation was to obtain data on the composition and properties of Pu(IV) hydrous oxide or other compounds formed in alkaline media under different conditions. Such information is important to understand Pu(IV) behavior and the forms of its existence in the Hanford Site alkaline tank waste sludge. This knowledge then may be applied in assessing plutonium criticality hazards in the storage, retrieval, and treatment of Hanford Site tank wastes as well as in understanding its contribution to the transuranic waste inventory (threshold at 100 nCi/g or about 5 {times} 10{sup {minus}6} M) of the separate solution and solid phases.

  11. absolute radionuclide activity: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Sciences and Ecology Websites Summary: series radionuclides Fuel Fabrication Facility 99 Tc (reprocessing only) Enriched uranium Transuranics (e.g., 237 Np, 239 Pu) (reprocessing...

  12. DOE Annual NEPA Planning Summary report templates 2011

    Broader source: Energy.gov (indexed) [DOE]

    power systems for space and national security missions. These systems provide heat and electricity in remote harsh environments by converting the decay heat from Pu-238...

  13. ARQ_29MAY2013.indd

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Pu-238 project of the Nuclear Nonproliferation and Security Program, is investing in preventive maintenance to further sustain reliability of equipment. Some more recent...

  14. Application of Neutron-Absorbing Structural-Amorphous metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Controls

    E-Print Network [OSTI]

    2006-01-01T23:59:59.000Z

    241 Pu, etc. ). To prevent nuclear criticality in spent fuelto enhance criticality safety for spent nuclear fuel inSpent Nuclear Fuel (SNF) Container to Enhance Criticality

  15. UO

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    success in this work will deepen our fundamental understanding of the nuclear energy materials. Background: The actinide (U, Np, Pu) oxides, nitrides, and carbides are of...

  16. Self-regulating neutron coincidence counter

    DOE Patents [OSTI]

    Baron, N.

    1980-06-16T23:59:59.000Z

    A device for accurately measuring the mass of /sup 240/Pu and /sup 239/Pu in a sample having arbitrary moderation and mixed with various contaminants. The device utilizes a thermal neutron well counter which has two concentric rings of neutron detectors separated by a moderating material surrounding the well. Neutron spectroscopic information derived by the two rings of detectors is used to measure the quantity of /sup 239/Pu and /sup 240/Pu in device which corrects for background radiation, deadtime losses of the detector and electronics and various other constants of the system.

  17. al settore delle: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    stampantiscanner Computer Technologies and Information Sciences Websites Summary: 7232, che si trova presso la portineria di ingresso al Palazzo San Niccol pu...

  18. al regolamento emas: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    stampantiscanner Computer Technologies and Information Sciences Websites Summary: 7232, che si trova presso la portineria di ingresso al Palazzo San Niccol pu...

  19. acque utilizzo di: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    stampantiscanner Computer Technologies and Information Sciences Websites Summary: 7232, che si trova presso la portineria di ingresso al Palazzo San Niccol pu...

  20. TH`ESE DE DOCTORAT DE L'ECOLE NORMALE SUPERIEURE DE CACHAN

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    ´ebastien Marcille, No¨el Tchidjo-Moyo et Luca Rose. Merci `a tous ceux que j'ai pu c^otoyer au cours de mes

  1. SCKCEN/4126769 dinsdag 22 juli 2014Copyright MYRRHA Spallation Target Design and

    E-Print Network [OSTI]

    McDonald, Kirk

    MTBF > 250 h 5 Reactor power ~85 MWth keff 0.95 spectrum fast (flexible) fuel 30 to 35% Pu MOX coolant

  2. Anticipated dose to workers for Plutonium Stabilization and Handling at PFP Project W-460

    SciTech Connect (OSTI)

    LILLY, J.T.

    1999-11-30T23:59:59.000Z

    Report provides estimates of expected whole body and extremity radiological dose to workers conducting planned Pu stabilization and packaging operations at PFP.

  3. WebPlug: A Framework for the Web of Things Benedikt Ostermaier, Fabian Schlup, Kay Rmer

    E-Print Network [OSTI]

    resources, which can be accessed using lightweight APIs based on the REST principle [1]. Exposing real API for the same pu

  4. Structural Characterization of and Plutonium Sorption on Mesoporous and Nanoparticulate Ferrihydrite

    E-Print Network [OSTI]

    Brogan, Luna Kestrel Schwaiger

    2012-01-01T23:59:59.000Z

    known that nanoparticle photocatalysis will increase withThe discovery of the photocatalysis for Pu reduction wasnoteworthy that the photocatalysis can be greatly affected

  5. Self-irradiation damage to the local structure of plutonium and plutonium intermetallics

    SciTech Connect (OSTI)

    Booth, C. H.; Jiang Yu; Medling, S. A. [Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, California 94720 (United States); Wang, D. L. [Nuclear Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, California 94720 (United States); Costello, A. L.; Schwartz, D. S.; Mitchell, J. N.; Tobash, P. H. [Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Bauer, E. D. [Materials Physics and Applications Division, Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); McCall, S. K.; Wall, M. A.; Allen, P. G. [Condensed Matter and Materials Division, Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

    2013-03-07T23:59:59.000Z

    The effect of self-irradiation damage on the local structure of {delta}-Pu, PuAl{sub 2}, PuGa{sub 3}, and other Pu intermetallics has been determined for samples stored at room temperature using the extended x-ray absorption fine-structure (EXAFS) technique. These measurements indicate that the intermetallic samples damage at a similar rate as indicated in previous studies of PuCoGa{sub 5}. In contrast, {delta}-Pu data indicate a much slower damage accumulation rate. To explore the effect of storage temperature and possible room temperature annealing effects, we also collected EXAFS data on a {delta}-Pu sample that was held at less than 32 K for a two month period. This sample damaged much more quickly. In addition, the measurable damage was annealed out at above only 135 K. Data from samples of {delta}-Pu with different Ga concentrations and results on all samples collected from different absorption edges are also reported. These results are discussed in terms of the vibrational properties of the materials and the role of Ga in {delta}-Pu as a network former.

  6. asa cssa sssa: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    ,CSSA,andSSSA.Allcopyrightsreserved. Simulation of Phytoremediation of a TNT-Contaminated Soil Environmental Sciences and Ecology Websites Summary: JournalofEnvironmentalQuality.Pu...

  7. Analysis of a Nuclear Accident: Fission and Activation Product Releases from the Fukushima Daiichi Nuclear Facility as Remote Indicators of Source Identification, Extent of Release, and State of Damaged Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Schwantes, Jon M.; Orton, Christopher R.; Clark, Richard A.

    2011-12-05T23:59:59.000Z

    Evidence of the release Pu from the Fukushima Daiichi nuclear power station to the local environment and surrounding communities and estimates on fraction of total fuel inventory released

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

    SciTech Connect (OSTI)

    Fuller, Kenneth S. Jr.; Smith, Robert H. Jr.; Goergen, Charles R. [Savannah River Nuclear Solutions, LLC, Savannah River Site, Aiken, SC 29802 (United States)] [Savannah River Nuclear Solutions, LLC, Savannah River Site, Aiken, SC 29802 (United States)

    2013-07-01T23:59:59.000Z

    Plutonium-238 is used in Radioisotope Thermoelectric Generators (RTGs) to generate electrical power and in Radioisotope Heater Units (RHUs) to produce heat for electronics and environmental control for deep space missions. The domestic supply of Pu-238 consists of scrap material from previous mission production or material purchased from Russia. Currently, the United States has no significant production scale operational capability to produce and separate new Pu-238 from irradiated neptunium-237 targets. The Department of Energy - Nuclear Energy is currently evaluating and developing plans to reconstitute the United States capability to produce Pu-238 from irradiated Np-237 targets. The Savannah River Site had previously produced and/or processed all the Pu-238 utilized in Radioisotope Thermoelectric Generators (RTGs) for deep space missions up to and including the majority of the plutonium for the Cassini Mission. The previous full production cycle capabilities included: Np- 237 target fabrication, target irradiation, target dissolution and Np-237 and Pu-238 separation and purification, conversion of Np-237 and Pu-238 to oxide, scrap recovery, and Pu-238 encapsulation. The capability and equipment still exist and could be revitalized or put back into service to recover and purify Pu-238/Np-237 or broken General Purpose Heat Source (GPHS) pellets utilizing existing process equipment in HB-Line Scrap Recovery, and H-Canyon Frame Waste Recovery processes. The conversion of Np-237 and Pu-238 to oxide can be performed in the existing HB-Line Phase-2 and Phase- 3 Processes. Dissolution of irradiated Np-237 target material, and separation and purification of Np-237 and Pu-238 product streams would be possible at production rates of ?2 kg/month of Pu-238 if the existing H-Canyon Frames Process spare equipment were re-installed. Previously, the primary H-Canyon Frames equipment was removed to be replaced: however, the replacement project was stopped. The spare equipment is stored and still available for installation. Out of specification Pu-238 scrap material can be purified and recovered by utilizing the HB-Line Phase- 1 Scrap Recovery Line and the Phase-3 Pu-238 Oxide Conversion Line along with H-Canyon Frame Waste Recovery process. In addition, it also covers and describes utilizing the Phase-2 Np-237 Oxide Conversion Line, in conjunction with the H-Canyon Frames Process to restore the H-Canyon capability to process and recover Np-237 and Pu-238 from irradiated Np-237 targets and address potential synergies with other programs like recovery of Pu-244 and heavy isotopes of curium from other target material. (authors)

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

    SciTech Connect (OSTI)

    Fuller, K.; Smith, Robert H. Jr.; Goergen, Charles R.

    2013-01-09T23:59:59.000Z

    Plutonium-238 is used in Radioisotope Thermoelectric Generators (RTGs) to generate electrical power and in Radioisotope Heater Units (RHUs) to produce heat for electronics and environmental control for deep space missions. The domestic supply of Pu-238 consists of scrap material from previous mission production or material purchased from Russia. Currently, the United States has no significant production scale operational capability to produce and separate new Pu-238 from irradiated neptunium-237 targets. The Department of Energy - Nuclear Energy is currently evaluating and developing plans to reconstitute the United States capability to produce Pu-238 from irradiated Np-237 targets. The Savannah River Site had previously produced and/or processed all the Pu-238 utilized in Radioisotope Thermoelectric Generators (RTGs) for deep space missions up to and including the majority of the plutonium for the Cassini Mission. The previous full production cycle capabilities included: Np-237 target fabrication, target irradiation, target dissolution and Np-237 and Pu-238 separation and purification, conversion of Np-237 and Pu-238 to oxide, scrap recovery, and Pu-238 encapsulation. The capability and equipment still exist and could be revitalized or put back into service to recover and purify Pu-238/Np-237 or broken General Purpose Heat Source (GPHS) pellets utilizing existing process equipment in HB-Line Scrap Recovery, and H-anyon Frame Waste Recovery processes. The conversion of Np-237 and Pu-238 to oxide can be performed in the existing HB-Line Phase-2 and Phase-3 Processes. Dissolution of irradiated Np-237 target material, and separation and purification of Np-237 and Pu-238 product streams would be possible at production rates of ~ 2 kg/month of Pu-238 if the existing H-Canyon Frames Process spare equipment were re-installed. Previously, the primary H-Canyon Frames equipment was removed to be replaced: however, the replacement project was stopped. The spare equipment is stored and still available for installation. Out of specification Pu-238 scrap material can be purified and recovered by utilizing the HB-Line Phase-1 Scrap Recovery Line and the Phase-3 Pu-238 Oxide Conversion Line along with H-Canyon Frame Waste Recovery process. In addition, it also covers and describes utilizing the Phase-2 Np-237 Oxide Conversion Line, in conjunction with the H-Canyon Frames Process to restore the H-Canyon capability to process and recover Np-237 and Pu-238 from irradiated Np-237 targets and address potential synergies with other programs like recovery of Pu-244 and heavy isotopes of curium from other target material.

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

    SciTech Connect (OSTI)

    Permana, Sidik; Novitrian,; Waris, Abdul [Nuclear Physics and Biophysics Research Division, Physics Department, Institut Teknologi Bandung (Indonesia); Ismail [Center for Technical Assessment of Nuclear Installation and Materials, Indonesian Nuclear Energy Regulatory (Indonesia); Suzuki, Mitsutoshi [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA) (Japan); Saito, Masaki [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan)

    2014-09-30T23:59:59.000Z

    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.

  11. Estimation of the formation rates of polyatomic species of heavy metals in plutonium analyses using a multicollector ICP-MS with a desolvating nebulizer

    SciTech Connect (OSTI)

    Mitroshkov, Alexandre V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Olsen, Khris B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Thomas, Linda M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-01-01T23:59:59.000Z

    The analyses of IAEA and environmental samples for Plutonium isotopic content are conducted normally at very low concentrations of Pu–usually in the range of part per trillion level and even more often at the parts per quadrillion level. To analyze such low concentrations, the interferences in the analytical solution must be reduced as much as possible. Polyatomic interferences (PIs), formed by the heavy metals (HMs) from Hf to Bi are known to create the problems for Pu isotopic analyses, because even the relatively high resolution of a modern multicollector ICP-MS is not enough to separate Pu isotopes from this PIs in most of the cases. Desolvating nebulizers (DSN) (e.g. APEX and AridusII) reduce significantly the formation of PIs compare to the use of wet plasma. The purpose of this work was to investigate the rate of formation of PIs, produced by HMs, when high resolution MC ICP-MS with desolvating nebulizer was used for Pu isotopic analyses and to estimate the influence of the metals present in the sample on the results of analyses. The NU Plasma HR Multicollector and AridusII desolvating nebulizer were used in this investigation. This investigation was done for all Pu isotopes normally analyzed by ICP-MS, including ²??Pu, with the exception of ²³?Pu, which most of the time can’t be analyzed by ICP-MS, because of the overwhelming presence of ²³?U in the solutions. The PI formation rates were determined and reported for all 12 HMs from Hf to Bi. Selected IAEA samples were scanned for the presence of HMs and the influence of HMs on the results of Pu isotopic analyses was evaluated. It was found that the implemented separation procedure provides sufficient separation of HM from Pu, although the effect of PIs on the measurement of low level isotopes like ²?¹Pu and ²?²Pu in some cases can still be observed.

  12. New approaches for MOX multi-recycling

    SciTech Connect (OSTI)

    Gain, T.; Bouvier, E.; Grosman, R.; Senentz, G.H.; Lelievre, F.; Bailly, F.; Brueziere, J. [AREVA NC, 1 place Jean Millier, Paris La Defense, 92084 (France); Murray, P. [AREVA Federal Services LLC, 4800 Hampden Lane, Bethesda, MD 20814 (United States)

    2013-07-01T23:59:59.000Z

    Due to its low fissile content after irradiation, Pu from used MOX fuel is considered by some as not recyclable in LWR (Light Water Reactors). The point of this paper is hence to go back to those statements and provide a new analysis based on AREVA extended experience in the fields of fissile and fertile material management and optimized waste management. This is done using the current US fuel inventory as a case study. MOX Multi-recycling in LWRs is a closed cycle scenario where U and Pu management through reprocessing and recycling leads to a significant reduction of the used assemblies to be stored. The recycling of Pu in MOX fuel is moreover a way to maintain the self-protection of the Pu-bearing assemblies. With this scenario, Pu content is also reduced repetitively via a multi-recycling of MOX in LWRs. Simultaneously, {sup 238}Pu content decreases. All along this scenario, HLW (High-Level Radioactive Waste) vitrified canisters are produced and planned for deep geological disposal. Contrary to used fuel, HLW vitrified canisters do not contain proliferation materials. Moreover, the reprocessing of used fuel limits the space needed on current interim storage. With MOX multi-recycling in LWR, Pu isotopy needs to be managed carefully all along the scenario. The early introduction of a limited number of SFRs (Sodium Fast Reactors) can therefore be a real asset for the overall system. A few SFRs would be enough to improve the Pu isotopy from used LWR MOX fuel and provide a Pu-isotopy that could be mixed back with multi-recycled Pu from LWRs, hence increasing the Pu multi-recycling potential in LWRs.

  13. Non-destructive assay of EBR-II blanket elements using resonance transmission analysis.

    SciTech Connect (OSTI)

    Klann, R.T.; Poenitz, W.P.

    1998-09-11T23:59:59.000Z

    Resonance transmission analysis utilizing a faltered reactor beam was examined as a means of determining the {sup 239}Pu content in Experimental Breeder Reactor-II depleted uranium blanket elements. The technique uses cadmium and gadolinium falters along with a {sup 239}Pu fission chamber to isolate the 0.3 eV resonance in {sup 239}Pu. In the energy range of this resonance (0.1 eV to 0.5 ev), the total microscopic cross-section of {sup 239}Pu is significantly greater than the cross-sections of {sup 238}U and {sup 235}U. This large difference allows small changes in the {sup 239}Pu content of a sample to result in large changes in the mass signal response. Tests with small stacks of depleted uranium and {sup 239}Pu foils indicate a significant change in response based on the {sup 239}Pu content of the foil stack. In addition, the tests indicate good agreement between the measured and predicted values of {sup 239}Pu up to approximately two weight percent.

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

    SciTech Connect (OSTI)

    Hikaru Hiruta; Gilles Youinou

    2013-09-01T23:59:59.000Z

    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.

  15. 7/1/2003 00101 Marcare con una crocetta le risposte ritenute corrette e consegnare la scheda al termine della prima

    E-Print Network [OSTI]

    Frosini, Patrizio

    del tipo n Ã? (n + 1) A) ha infinite soluzioni. B) pu`o essere impossibile. C) ha una sola soluzione. D) se `e omogeneo ha una sola soluzione. #12;7/1/2003 00110 Marcare con una crocetta le risposte omogeneo ha una sola soluzione. B) ha una sola soluzione. C) pu`o essere impossibile. D) ha infinite

  16. Cognome: Nome: Matricola: Matematica e Statistica -Docente: P. Causin

    E-Print Network [OSTI]

    Causin, Paola

    alghe secondo tale modello per t = 4 mesi? ESERCIZIO 4. Contrassegnare la sola risposta esatta. Si riguardo le radici della funzione tra 0 e 2? A c'`e una sola radice in tale intervallo; B non possiamo pu`o dare una sola risposta tra 4 alternative proposte, in quanti modi diversi uno studente pu

  17. Microsoft Outlook - Memo Style

    National Nuclear Security Administration (NNSA)

    Pu-240 8.02E-02 Pu-241 2.45E+00 2.83E+00 Total: kgm3 Parameter Iron-based MetalsAlloys 2.00E+01 Aluminum-based MetalsAlloys 3.00E+00 Other Metals 1.00E+00 Other...

  18. NSTX Program Governance, Research Support and Facility Operation

    E-Print Network [OSTI]

    Princeton Plasma Physics Laboratory

    of DPP, 3 PU faculty) · Graduate students & post-doc (from PU) · Engineering expertise: designNSTX Program Governance, Research Support and Facility Operation Office of Science M.G. Bell, PPPL for the NSTX Research Team NSTX 5 Year Plan Review for 2009-13 Princeton Plasma Physics Laboratory July 28

  19. PERSPECTIVES IN OPERATOR THEORY BANACH CENTER PUBLICATIONS, VOLUME 75

    E-Print Network [OSTI]

    Latushkin, Yuri

    is invertible in 2 (Z, Cd ), and by a direct computation, its inverse is a difference operator, K = (Kjk)j,kZ, with kernel defined by (1.2) Kjk = Uj(I - P)U-1 k+1 for j > k and Kjk = -UjPU-1 k+1 for j k. If AÃ? = (AÃ? j )j

  20. Caustic Precipitation of Plutonium and Uranium with Gadolinium as a Neutron Poison

    SciTech Connect (OSTI)

    VISSER, ANN E.; BRONIKOWSKI, MICHAEL G.; RUDISILL, TRACY S.

    2005-10-18T23:59:59.000Z

    The caustic precipitation of plutonium (Pu) and uranium (U) from Pu and U-containing waste solutions has been investigated to determine whether gadolinium (Gd) could be used as a neutron poison for precipitation with greater than a fissile mass containing both Pu and enriched U. Precipitation experiments were performed using both process solution samples and simulant solutions with a range of 2.6-5.16 g/L U and 0-4.3:1 U:Pu. Analyses were performed on solutions at intermediate pH to determine the partitioning of elements for accident scenarios. When both Pu and U were present in the solution, precipitation began at pH 4.5 and by pH 7, 99% of Pu and U had precipitated. When complete neutralization was achieved at pH > 14 with 1.2 M excess OH{sup -}, greater than 99% of Pu, U, and Gd had precipitated. At pH > 14, the particles sizes were larger and the distribution was a single mode. The ratio of hydrogen:fissile atoms in the precipitate was determined after both settling and centrifuging and indicates that sufficient water was associated with the precipitates to provide the needed neutron moderation for Gd to prevent a criticality in solutions containing up to 4.3:1 U:Pu and up to 5.16 g/L U.

  1. On Optimal Sequential Prediction for General Processes Andrew B. Nobel

    E-Print Network [OSTI]

    Nobel, Andrew

    -stationary processes under p-th power loss p(u, v) = |u-v|p , 1 'th power loss. Andrew Nobel is with the Department of Statistics, University of North Carolina, Chapel'th power loss of the form p(u, v) = |u - v|p, with 1

  2. Subscriber access provided by WOODS HOLE OCEANOGRAPHIC INST Environmental Science & Technology is published by the American Chemical

    E-Print Network [OSTI]

    Buesseler, Ken

    , Savannah River National Laboratory, Aiken, South Carolina 29808, and State Key Laboratory of Marine-Area of the Savannah River Site and compared to similar samples collected six years earlier. Two sources of Pu were of Pu geochemistry in the laboratory are understood, its fate in the natural environment is far more di

  3. Nordisk kernesikkerhedsforskning Norrnar kjarnryggisrannsknir

    E-Print Network [OSTI]

    in the Nordic lake sedi- ment, pore-water and lake water; (7) Sequential extraction of Pu in soil, sediment sediment, pore-water and lake water; (7) Sequential extraction of Pu in soil, sediment and concrete samples in partners labs, which in- cludes: (1) Further development on the speciation of 129I and 127I in water sam

  4. DISSOLUTION OF FISSILE MATERIALS CONTAINING TANTALUM METAL

    SciTech Connect (OSTI)

    Rudisill, T; Mark Crowder, M; Michael Bronikowski, M

    2007-05-29T23:59:59.000Z

    The dissolution of composite materials containing plutonium (Pu) and tantalum (Ta) metals is currently performed in Phase I of the HB-Line facility. The conditions for the present flowsheet are the dissolution of 500 g of Pu metal in the 15 L dissolver using a 4 M nitric acid (HNO{sub 3}) solution containing 0.2 M potassium fluoride (KF) at 95 C for 4-6 h.[1] The Ta metal, which is essentially insoluble in HNO{sub 3}/fluoride solutions, is rinsed with process water to remove residual acid, and then burned to destroy classified information. During the initial dissolution campaign, the total mass of Pu and Ta in the dissolver charge was limited to nominally 300 g. The reduced amount of Pu in the dissolver charge coupled with significant evaporation of solution during processing of several dissolver charges resulted in the precipitation of a fluoride salt contain Pu. Dissolution of the salt required the addition of aluminum nitrate (Al(NO{sub 3}){sub 3}) and a subsequent undesired 4 h heating cycle. As a result of this issue, HB-Line Engineering requested the Savannah River National Laboratory (SRNL) to optimize the dissolution flowsheet to reduce the cycle time, reduce the risk of precipitating solids, and obtain hydrogen (H{sub 2}) generation data at lower fluoride concentrations.[2] Using samples of the Pu/Ta composite material, we performed three experiments to demonstrate the dissolution of the Pu metal using HNO{sub 3} solutions containing 0.15 and 0.175 M KF. When 0.15 M KF was used in the dissolving solution, 95.5% of the Pu in the sample dissolved in approximately 6 h. The undissolved material included a small amount of Pu metal and plutonium oxide (PuO{sub 2}) solids. Complete dissolution of the metal would have likely occurred if the dissolution time had been extended. This assumption is based on the steady increase in the Pu concentration observed during the last several hours of the experiment. We attribute the formation of PuO{sub 2} to the complexation of fluoride by the Pu. The fluoride became unavailable to catalyze the dissolution of PuO{sub 2} as it formed on the surface of the metal. The mass of Pu dissolved is equivalent to the dissolution of 343 g of Pu in the HB-Line dissolvers. In the initial experiment with 0.175 M KF in the solution, we achieved complete dissolution of the Pu in 6 h. The mass of Pu dissolved scales to the dissolution of 358 g of Pu in the HB-Line dissolvers. The second experiment using 0.175 M KF was terminated after approximately 6 h following the dissolution of 92.7% of the Pu in the sample; however, dissolution of additional Pu was severely limited due to the slow dissolution rate observed beyond approximately 4 h. A small amount of PuO{sub 2} was also produced in the solution. The slow rate of dissolution was attributed to the diminishing surface area of the Pu and a reduction in the fluoride activity due to complexation with Pu. Given time (>4 h), the Pu metal may have dissolved using the original solution or a significant portion may have oxidized to PuO{sub 2}. If the metal oxidized to PuO{sub 2}, we expect little of the material would have dissolved due to the fluoride complexation and the low HNO{sub 3} concentration. The mass of Pu dissolved in the second experiment scales to the dissolution of 309 g of Pu in the HB-Line dissolvers. Based on the data from the Pu/Ta dissolution experiments we recommend the use of 4 M HNO{sub 3} containing 0.175 M KF for the dissolution of 300 g of Pu metal in the 15 L HB-Line dissolver. A dissolution temperature of nominally 95 C should allow for essentially complete dissolution of the metal in 6 h. Although the H{sub 2} concentration in the offgas from the experiments was at or below the detection limit of the gas chromatograph (GC) used in these experiments, small concentrations (<3 vol %) of H{sub 2} are typically produced in the offgas during Pu metal dissolutions. Therefore, appropriate controls must be established to address the small H{sub 3} generation rates in accordance with this work and the earlier flowsheet demonstrated

  5. Neutron scattering studies in the actinide region. Progress report, August 1, 1992--July 31, 1993

    SciTech Connect (OSTI)

    Kegel, G.H.R.; Egan, J.J.

    1993-09-01T23:59:59.000Z

    This report discusses the following topics: Prompt fission neutron energy spectra for {sup 235}U and {sup 239}Pu; Two-parameter measurement of nuclear lifetimes; ``Black`` neutron detector; Data reduction techniques for neutron scattering experiments; Inelastic neutron scattering studies in {sup 197}Au; Elastic and inelastic scattering studies in {sup 239}Pu; and neutron induced defects in silicon dioxide MOS structures.

  6. First Semester Philadelphia University

    E-Print Network [OSTI]

    Groups. Classification of Subgroups of Cyclic Groups. 6 Permutation Groups: Definition and Notation/ Cole 2010. Call number in PU library: 512.02 GAL. - John B. Fraleigh, A First Course in Abstract Algebra, 7th Edition, Pearson 2003. Call number in PU library: 512.02 FRA. - Amin Witno, Group Theory

  7. J PHYS TV FRANCE 7 ( 1997) Colloque C3, Supplement au Journal de Physique 111d'aoiit 1997

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Behaviour of Polycarbonate/Polyurethane Multi-Layer for Transparent Armor W. Toqueboeuf, B. Mortaigne and C of layered Polycarbonate (PC)IPolyurethane (PU) polymers used in lightweight transparent armor bilayer feuilletages de Polycarbonate (PC)IPolyurCthanne (PU) utilisis en face arriere de concept de blindages legers

  8. SYNTHESE EN FRANAIS TITRE: NEUTRONIC STUDY OF THE MONO-RECYCLING OF AMERICIUM IN PWR

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    retraitement standard consiste à séparer le plutonium afin de fabriquer un combustible MOX sur base d'uranium appauvri. La concentration du Pu dans le MOX est déterminée pour atteindre un taux d'irradiation du MOX de refroidissement de 30 ans demande à augmenter la teneur en Pu dans le MOX. L'241Am, avec une durée de vie de 432

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

    SciTech Connect (OSTI)

    Sonat Sen; Gilles Youinou

    2013-02-01T23:59:59.000Z

    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. Open Archive TOULOUSE Archive Ouverte (OATAO) OATAO is an open access repository that collects the work of Toulouse researchers and

    E-Print Network [OSTI]

    Mailhes, Corinne

    : staff-oatao@listes-diff.inp-toulouse.fr #12;Recovery of actinides from actinide­aluminium alloys A chlorination route is being investigated for recovery of actinides from actinide­aluminium alloys, which experimentally tested using U­Pu­Al alloy prepared by electrodeposition of U and Pu on solid aluminium plate

  11. Role of self-irradiation defects on the ageing of 239 , J.-C. Griveau1

    E-Print Network [OSTI]

    Boyer, Edmond

    and critical current density measurements, are used to study the ageing of the actinide superconductor PuCoGa5. These measurements reveal that 2-nm sized non- superconducting point-like regions are the main damage formed during. Introduction. ­ The discovery of superconductivity in PuCoGa5 [1], with characteristics intermediate between

  12. Hycon2 Benchmark: Power Network System Stefano Riverso

    E-Print Network [OSTI]

    Ferrari-Trecate, Giancarlo

    steps and produce the required power. We consider thermal power stations with single-stage turbines). In particular, we note that Dss preserves the input-decoupled structure of C [i] while D does not. 2 #12;i nominal value (p.u.) Pvi Deviation of the steam valve position from nominal value (p.u.) Prefi Deviation

  13. Development of a Real-Time Detection Strategy for Material Accountancy and Process Monitoring During Nuclear Fuel Reprocessing Using the Urex+3A Method

    E-Print Network [OSTI]

    Goddard, Braden

    2010-07-14T23:59:59.000Z

    radioactive activation products and transuranic (TRU) actinide elements, such as Pu, Np, Am, and Cm. The FP, U, and TRU isotopes emit gamma rays with unique energy spectra that are characteristic to each of the individual isotopes. These unique spectra.../Pu UREX+1 U Tc Cs/Sr TRUs/Ln FPs UREX+1a U Tc Cs/Sr TRUs FPs/Ln UREX+1b U Tc Cs/Sr U/TRUs FPs/Ln UREX+2 U Tc Cs/Sr Pu/Np Am/Cm/Ln FPs UREX+2a U Tc Cs/Sr U/Pu/Np Am/Cm/Ln FPs UREX+3 U Tc Cs/Sr Pu/Np Am/Cm FPs...

  14. BAR-CODE BASED WEIGHT MEASUREMENT STATION FOR PHYSICAL INVENTORY TAKING OF PLUTONIUM OXIDE CONTAINERS AT THE MINING AND CHEMICAL COMBINE RADIOCHEMICAL REPROCESSING PLANT NEAR KRASNOYARSK, SIBERIA.

    SciTech Connect (OSTI)

    SUDA,S.

    1999-09-20T23:59:59.000Z

    This paper describes the technical tasks being implemented to computerize the physical inventory taking (PIT) at the Mining and Chemical Combine (Gorno-Khimichesky Kombinat, GKhK) radiochemical plant under the US/Russian cooperative nuclear material protection, control, and accounting (MPC and A) program. Under the MPC and A program, Lab-to-Lab task agreements with GKhK were negotiated that involved computerized equipment for item verification and confirmatory measurement of the Pu containers. Tasks under Phase I cover the work for demonstrating the plan and procedures for carrying out the comparison of the Pu container identification on the container with the computerized inventory records. In addition to the records validation, the verification procedures include the application of bar codes and bar coded TIDs to the Pu containers. Phase II involves the verification of the Pu content. A plan and procedures are being written for carrying out confirmatory measurements on the Pu containers.

  15. MANTRA: An Integral Reactor Physics Experiment to Infer Actinide Capture Cross-sections from Thorium to Californium with Accelerator Mass Spectrometry

    SciTech Connect (OSTI)

    G. Youinou; C. McGrath; G. Imel; M. Paul; R. Pardo; F. Kondev; M. Salvatores; G. Palmiotti

    2011-08-01T23:59:59.000Z

    The principle of the proposed experiment is to irradiate very pure actinide samples in the Advanced Test Reactor at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation will allow inference of effective neutron capture cross-sections. This approach has been used in the past and the novelty of this experiment is that the atom densities of the different transmutation products will be determined using the Accelerator Mass Spectrometry technique at the ATLAS facility located at ANL. It is currently planned to irradiate the following isotopes: 232Th, 235U, 236U, 238U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 244Cm and 248Cm.

  16. ~,,{--Oz ~~~,,,,,,{{{--OOOzzz

    E-Print Network [OSTI]

    Zhang, Li-Xin

    U = pU (t) ÷veã·§, PpL (T t) = /2, PpU (T t) = 1-PpU (T t+1) = /2. K[pL(T), pU (T)]´p~&Y²·1 - ~&«m, ·§ du #12;1o�!«m O §ÏéͶþ ~,,·{--ò�OþͶþz pL = pL(t), pU = pU (t) ÷veã·§, PpL (T t) = /2, PpU (T t·§, PpL (T t) = /2, PpU (T t) = 1-PpU (T t+1) = /2. K[pL(T), pU (T)]´p~&Y²·1 - ~&«m, ·§ du P

  17. OBES "One Pager"

    SciTech Connect (OSTI)

    Tobin, J G

    2008-11-07T23:59:59.000Z

    We are developing and utilizing photon dichroic and spin resolved techniques to investigate electron correlation in complex systems. These materials include potential spintronic device sources such as Fe/GaAs and f electronic materials such as non-magnetic {delta}-Pu. We are pursuing Double Polarization Photoelectron Dichroism measurements of the Fano Effect, using spin resolving detection in photoelectron spectroscopy, to test the nature of electron correlation in Pu. (See Pubs 2, 4 & 5.) If successful, we will solve the riddle of Pu electronic structure that has remained unresolved for the last 60 years. We are also developing a Bremstrahlung Isochromat Spectroscopy (BIS) capability to permit the direct measurement of the unoccupied electronic structure of Pu, which is another missing piece in the puzzle of Pu electronic structure.

  18. Delocalization and occupancy effects of 5f orbitals in plutonium intermetallics using L3-edge resonant X-ray emission spectroscopy

    SciTech Connect (OSTI)

    Booth, C. H.; Medling, S. A.; Jiang, Yu; Bauer, E. D.; Tobash, P. H.; Mitchell, J. N.; Veirs, D. K.; Wall, M. A.; Allen, P. G.; Kas, J. J.; Sokaras, D.; Nordlund, D.; Weng, T.-C.

    2014-06-24T23:59:59.000Z

    Although actinide (An) L3 -edge X-ray absorption near-edge structure (XANES) spectroscopy has been very effective in determining An oxidation states in insulating, ionically bonded materials, such as in certain coordination compounds and mineral systems, the technique fails in systems featuring more delocalized 5f orbitals, especially in metals. Recently, actinide L3-edge resonant X-ray emission spec- troscopy (RXES) has been shown to be an effective alternative. This technique is further demonstrated here using a parameterized partial unoccupied density of states method to quantify both occupancy and delocalization of the 5f orbital in ?-Pu, ?-Pu, PuCoGa5 , PuCoIn5 , and PuSb2. These new results, supported by FEFF calculations, highlight the effects of strong correlations on RXES spectra and the technique?s ability to differentiate between f-orbital occupation and delocalization.

  19. Comparison of Spectroscopic Data with Cluster Calculations of Plutonium, Plutonium Dioxide and Uranium Dioxide

    SciTech Connect (OSTI)

    Tobin, J G; Yu, S W; Chung, B W; Ryzhkov, M V; Mirmelstein, A

    2012-05-15T23:59:59.000Z

    Using spectroscopic data produced in the experimental investigations of bulk systems, including X-Ray Absorption Spectroscopy (XAS), Photoelectron Spectroscopy (PES) and Bremstrahlung Isochromat Spectroscopy (BIS), the theoretical results within for UO{sub 2}{sup 6}, PuO{sub 2}{sup 6} and Pu{sup 7} clusters have been evaluated. The calculations of the electronic structure of the clusters have been performed within the framework of the Relativistic Discrete-Variational Method (RDV). The comparisons between the LLNL experimental data and the Russian calculations are quite favorable. The cluster calculations may represent a new and useful avenue to address unresolved questions within the field of actinide electron structure, particularly that of Pu. Observation of the changes in the Pu electronic structure as a function of size suggests interesting implications for bulk Pu electronic structure.

  20. LITERATURE REVIEW OF PUO2 CALCINATION TIME AND TEMPERATURE DATA FOR SPECIFIC SURFACE AREA

    SciTech Connect (OSTI)

    Daniel, G.

    2012-03-06T23:59:59.000Z

    The literature has been reviewed in December 2011 for calcination data of plutonium oxide (PuO{sub 2}) from plutonium oxalate Pu(C{sub 2}O{sub 4}){sub 2} precipitation with respect to the PuO{sub 2} specific surface area (SSA). A summary of the literature is presented for what are believed to be the dominant factors influencing SSA, the calcination temperature and time. The PuO{sub 2} from Pu(C{sub 2}O{sub 4}){sub 2} calcination data from this review has been regressed to better understand the influence of calcination temperature and time on SSA. Based on this literature review data set, calcination temperature has a bigger impact on SSA versus time. However, there is still some variance in this data set that may be reflecting differences in the plutonium oxalate preparation or different calcination techniques. It is evident from this review that additional calcination temperature and time data for PuO{sub 2} from Pu(C{sub 2}O{sub 4}){sub 2} needs to be collected and evaluated to better define the relationship. The existing data set has a lot of calcination times that are about 2 hours and therefore may be underestimating the impact of heating time on SSA. SRNL recommends that more calcination temperature and time data for PuO{sub 2} from Pu(C{sub 2}O{sub 4}){sub 2} be collected and this literature review data set be augmented to better refine the relationship between PuO{sub 2} SSA and its calcination parameters.

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

    SciTech Connect (OSTI)

    Crowder, M.; Pierce, R.

    2012-08-22T23:59:59.000Z

    H-Canyon and HB-Line are tasked with the production of PuO{sub 2} from a feed of plutonium metal. The PuO{sub 2} will provide feed material for the MOX Fuel Fabrication Facility. After dissolution of the Pu metal in H-Canyon, the solution will be transferred to HB-Line for purification by anion exchange. Subsequent unit operations include Pu(IV) oxalate precipitation, filtration and calcination to form PuO{sub 2}. This report details the results from SRNL anion exchange, precipitation, filtration, calcination, and characterization tests, as requested by HB-Line1 and described in the task plan. This study involved an 80-g batch of Pu and employed test conditions prototypical of HB-Line conditions, wherever feasible. In addition, this study integrated lessons learned from earlier anion exchange and precipitation and calcination studies. H-Area Engineering selected direct strike Pu(IV) oxalate precipitation to produce a more dense PuO{sub 2} product than expected from Pu(III) oxalate precipitation. One benefit of the Pu(IV) approach is that it eliminates the need for reduction by ascorbic acid. The proposed HB-Line precipitation process involves a digestion time of 5 minutes after the time (44 min) required for oxalic acid addition. These were the conditions during HB-line production of neptunium oxide (NpO{sub 2}). In addition, a series of small Pu(IV) oxalate precipitation tests with different digestion times were conducted to better understand the effect of digestion time on particle size, filtration efficiency and other factors. To test the recommended process conditions, researchers performed two nearly-identical larger-scale precipitation and calcination tests. The calcined batches of PuO{sub 2} were characterized for density, specific surface area (SSA), particle size, moisture content, and impurities. Because the 3013 Standard requires that the calcination (or stabilization) process eliminate organics, characterization of PuO{sub 2} batches monitored the presence of oxalate by thermogravimetric analysis-mass spectrometry (TGA-MS). To use the TGA-MS for carbon or oxalate content, some method development will be required. However, the TGA-MS is already used for moisture measurements. Therefore, SRNL initiated method development for the TGA-MS to allow quantification of oxalate or total carbon. That work continues at this time and is not yet ready for use in this study. However, the collected test data can be reviewed later as those analysis tools are available.

  2. t h i s i s s u e D e a n s e n e c a , B a ' 9 0 | t a m a r a B r o w n , m e ' 0 3 | G r e G o r y m i c h a e l i D i s , m a ' 9 4 & B a ' 9 2 a puBlication of the universit y at Buffalo alumni association

    E-Print Network [OSTI]

    Krovi, Venkat

    on dining in the dorms #12;firstlook HayesHallRedux The original granite block walls, timber frame students Startup savvy 32 Young entrepreneurs share what it was like to launch successful companies

  3. 2 I E E E S o f t wa r E Pu b l i s h e d b y t h e I EEE C o m p u t e r S o c i e t y 0 74 0 -74 5 9 / 0 9 / $ 2 5 . 0 0 2 0 0 9 I E E E The bazaar has three major advantages over

    E-Print Network [OSTI]

    and Prosumers End-user development (EUD) aims to enable end users "at some point to create, modify, or extend engineering superscript #12;September/October 2009 I E E E S o f t wa r E 3 the bazaar: the prosumer. A prosumer serves as producer and consumer. If appropriate, the pro- sumer produces the goods he or she and

  4. 42 I E E E S O F T WA R E Pu b l i s h e d b y t h e I EEE C o m p u t e r S o c i e t y 0 74 0 -74 5 9 / 0 9 / $ 2 5 . 0 0 2 0 0 9 I E E E To overcome such obstacles, we developed a

    E-Print Network [OSTI]

    demonstration is important when developing safety-critical software such as a nuclear power plant's reactor-methods-based process that supports de- velopment, verification, and safety analysis. We also developed CASE tools to let nuclear engineers apply formal methods without having to know the underlying formalism in depth

  5. 42 I E E E S o f t wa r E Pu b l i s h e d b y t h e I EEE C o m p u t e r S o c i e t y 0 74 0 -74 5 9 / 0 9 / $ 2 5 . 0 0 2 0 0 9 I E E E To overcome such obstacles, we developed a

    E-Print Network [OSTI]

    demonstration is important when developing safety-critical software such as a nuclear power plant's reactor-methods-based process that supports de- velopment, verification, and safety analysis. We also developed CASE tools to let nuclear engineers apply formal methods without having to know the underlying formalism in depth

  6. Deployment Scenario of Heavy Water Cooled Thorium Breeder Reactor

    SciTech Connect (OSTI)

    Mardiansah, Deby; Takaki, Naoyuki [Course of Applied Science, School of Engineering, Tokai University (Japan)

    2010-06-22T23:59:59.000Z

    Deployment scenario of heavy water cooled thorium breeder reactor has been studied. We have assumed to use plutonium and thorium oxide fuel in water cooled reactor to produce {sup 233}U which will be used in thorium breeder reactor. The objective is to analysis the potential of water cooled Th-Pu reactor for replacing all of current LWRs especially in Japan. In this paper, the standard Pressurize Water Reactor (PWR) has been designed to produce 3423 MWt; (i) Th-Pu PWR, (ii) Th-Pu HWR (MFR = 1.0) and (iii) Th-Pu HWR (MFR 1.2). The properties and performance of the core were investigated by using cell and core calculation code. Th-Pu PWR or HWR produces {sup 233}U to introduce thorium breeder reactor. The result showed that to replace all (60 GWe) LWR by thorium breeder reactor within a period of one century, Th-Pu oxide fueled PWR has insufficient capability to produce necessary amount of {sup 233}U and Th-Pu oxide fueled HWR has almost enough potential to produce {sup 233}U but shows positive void reactivity coefficient.

  7. Properties of Liquid Plutonium

    SciTech Connect (OSTI)

    Freibert, Franz J. [Los Alamos National Laboratory; Mitchell, Jeremy N. [Los Alamos National Laboratory; Schwartz, Daniel S. [Los Alamos National Laboratory; Saleh, Tarik A. [Los Alamos National Laboratory; Migliori, Albert [Los Alamos National Laboratory

    2012-08-02T23:59:59.000Z

    Unalloyed polycrystalline Pu displays extreme thermal expansion behavior, i.e., {alpha} {yields} {beta} {yields} {gamma} {yields} {delta} increases by 25% in volume and {delta} {yields} {var_epsilon} {yields} liquid decreases by 4.5% in volume. Thus, making it difficult to measure density into the liquid state. Dilatometer outfitted with CaF molten metal cell offers a proven capability to measure thermal expansion in molten metals, but has yet to be proven for Pu. Historic data from the liquid nuclear fuels program will prove extremely useful as a guide to future measurements. 3.3at% Ga changes Pu molten metal properties: 50% increase in viscosity and {approx}3% decrease in density. Fe may decrease the density by a small amount assuming an averaging of densities for Pu-Ga and Pu-Fe liquids. More recent Boivineau (2009) work needs some interpretation, but technique is being employed in (U,Pu)O{sub 2} nuclear fuels program (Pu Futures, 2012).

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

    SciTech Connect (OSTI)

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

    2014-01-01T23:59:59.000Z

    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.

  9. Results of the Excreta Bioassay Quality Control Program for April 1, 2009 through March 31, 2010

    SciTech Connect (OSTI)

    Antonio, Cheryl L.

    2012-07-19T23:59:59.000Z

    A total of 58 urine samples and 10 fecal samples were submitted during the report period (April 1, 2009 through March 31, 2010) to General Engineering Laboratories, South Carolina by the Hanford Internal Dosimetry Program (IDP) to check the accuracy, precision, and detection levels of their analyses. Urine analyses for Sr, 238Pu, 239Pu, 241Am, 243Am 235U, 238U, elemental uranium and fecal analyses for 241Am, 238Pu and 239Pu were tested this year as well as four tissue samples for 238Pu, 239Pu, 241Am and 241Pu. The number of QC urine samples submitted during the report period represented 1.3% of the total samples submitted. In addition to the samples provided by IDP, GEL was also required to conduct their own QC program, and submit the results of analyses to IDP. About 33% of the analyses processed by GEL during the third year of this contract were quality control samples. GEL tested the performance of 21 radioisotopes, all of which met or exceeded the specifications in the Statement of Work within statistical uncertainty (Table 4).

  10. Dust transport: Wind blown and mechanical resuspension, July 1983 to December 1984

    SciTech Connect (OSTI)

    Langer, G.

    1986-09-20T23:59:59.000Z

    This study defines the processes that resuspend plutonium (Pu) particles from Pu-contaminated soil at Rocky Flats. Such knowledge can predict the transport of Pu particles from the site and the population dose. A vertical dust flux tower profiled the plume of Pu particles from the site. The data show a 70% reduction between 1 and 10 m in the concentration of coarse and inhalable Pu particles. The respirable particle concentration remained steady at both heights, slightly above background levels. High winds visually resuspend large amounts of dust for short periods, but we suspected that present sampling devices do not function properly above 50 km/h. During a windstorm reaching 80 km/h, the size-selective sampler used seriously underestimated the dust(Pu) concentration. Wind tunnel studies measured resuspension versus wind speed from our prairie grass covered, arid soil. We failed to find a good correlation between resuspension and wind speed. This led to a search for alternative mechanisms of resuspension besides wind erosion. Resuspension of dust(Pu) from grass proved to be important, as well as resuspension from rain splash.

  11. High-level assessment of LANL ABC Design

    SciTech Connect (OSTI)

    Not Available

    1994-04-15T23:59:59.000Z

    An annual weapon`s grade Pu disposition goal should be stated and related to the amount of Pu that needs to be disposed of. It needs to be determined to what extent it is possible to destroy Pu without building up any new Pu, i.e., how realistic this goal is. The strong positive Doppler coefficient for a Pu core might require the addition of some fertile material to ensure a negative Doppler coefficient. This in turn will affect the net Pu disposition rate. If a fertile material is required throughout the life of the ABC to ensure a negative Doppler coefficient, the difference between the molten salt ABC and other reactors in regard to Pu disposition is not a principled difference anymore but one of degree. A rationale has then to be developed that explains why {open_quotes}x{close_quotes} kg production of fissile material are acceptable but {open_quotes}y{close_quotes} kg are not. It is important to determine how a requirement for electricity production will impact on the ABC design choices. It is conceivable that DOE will not insist on electricity generation. In this case advantage has to be taken in terms of design simplifications and relaxed operating conditions.

  12. Final Report for Plutonium and Quantum Criticality LDRD 03-ERD-077

    SciTech Connect (OSTI)

    Fluss, M J; McCall, S K; Chung, B W; Chapline, G F; Jackson, D D; Heffner, R H; Haire, R G

    2008-02-11T23:59:59.000Z

    Plutonium possesses the most complicated phase diagram in the periodic table, driven by the complexities of overlapping 5f electron orbitals. Despite the importance of the 5f electrons in defining the structure and physical properties, there is no experimental evidence that these electrons localize to form magnetic moments in pure Pu and the {sup +}{mu}SR measurements included here place an upper limit of <0.001{micro}{sub B} for the magnetic moment on Pu. Instead, a large temperature independent Pauli susceptibility indicates they form narrow conduction bands. Radiation damage from the {alpha}-particle decay of Pu creates numerous defects in the crystal structure which produce a significant temperature dependent magnetic susceptibility {chi}(T), in {alpha}-Pu, {delta}-Pu(4.3at%Ga), and Pu{sub 1-x}Am{sub x} alloys ({delta}-Pu phase). This effect can be removed by thermal annealing above room temperature. By contrast, below 35K the radiation damage is frozen in place permitting the evolution in {chi}(T) with increasing damage to be studied systematically. This leads to a two component model consisting of a Curie-Weiss term and a short-ranged interaction term consistent with disorder induced local moment models. Thus it is shown that self-damage creates localized magnetic moments in previously nonmagnetic plutonium. This effect is greatly magnified in some Pu{sub 1-x}Am{sub x} alloys where an apparent damage-induced phase transition occurs at low temperatures near Stage I annealing which results local moments on the order of 1 {micro}{sub B}/Pu. The phase is metastable, and anneals away at higher temperatures.

  13. Preconceptual Feasibility Study to Evaluate Alternative Means to Produce Plutonium-238

    SciTech Connect (OSTI)

    John D. Bess; Matthew S. Everson

    2013-02-01T23:59:59.000Z

    There is currently no large-scale production of 238Pu in the United States. Feasibility studies were performed at the Idaho National Laboratory to assess the capability of developing alternative 238Pu production strategies. Initial investigations indicate potential capability to provision radioisotope-powered systems for future space exploration endeavors. For the short term production of 238Pu, sealed canisters of dilute 237Np solution in nitric acid could be irradiated in the Advanced Test Reactor (ATR). Targets in the large and medium “I” positions of the ATR were irradiated over a simulated period of 306 days and analyzed using MCNP5 and ORIGEN2.2. Approximately 0.5 kg of 238Pu could be produced annually in the ATR with purity greater than 92%. Optimization of the irradiation cycles could further increase the purity to greater than 98%. Whereas the typical purity of space batteries is between 80 to 85%, the higher purity 238Pu produced in the ATR could be blended with existing lower-purity inventory to produce useable material. Development of irradiation methods in the ATR provides the fastest alterative to restart United States 238Pu production. The analysis of 238Pu production in the ATR provides the technical basis for production using TRIGA® (Training, Research, Isotopes, General Atomics) nuclear reactors. Preliminary analyses envisage a production rate of approximately 0.7 kg annually using a single dedicated 5-MW TRIGA reactor with continuous flow loops to achieve high purity product. Two TRIGA reactors represent a robust means of providing at over 1 kg/yr of 238Pu annually using dilute solution targets of 237Np in nitric acid. Further collaboration and optimization of reactor design, radiochemical methods, and systems analyses would further increase annual 238Pu throughput, while reducing the currently evaluated reactor requirements.

  14. Reactor Physics Assessment of the Inclusion of Unseparated Neptunium in MOX Reactor Fuel

    SciTech Connect (OSTI)

    Ellis, Ronald James [ORNL

    2009-01-01T23:59:59.000Z

    Reducing the number of actinide separation streams in a spent fuel recovery process would reduce the cost and complexity of the process, and lower the quantity and numbers of solvents needed. It is more difficult and costly to separate Np and recombine it with Am-Cm prior to co-conversion than to simply co-strip it with the U-Pu-Np. Inclusion of the Np in mixed oxide (MOX) fuel for light water reactor (LWR) applications should not seriously affect the operating behavior of the reactor, nor should it pose insurmountable fuel design issues. In this work, the U, Pu, and Np from typical discharged and cooled PWR spent nuclear fuel are assumed to be used together in the preparation of MOX fuel for use in a pressurized water reactor (PWR). The reactor grade Pu isotopic vector is used in the model and the relative mass ratio of the Pu and Np content (Np/Pu mass is 0.061) from the cooled spent fuel is maintained but the overall Pu-Np MOX wt% is adjusted with respect to the U content (assumed to be at 0.25 wt% 235U enrichment) to offset reactivity and cycle length effects. The SCALE 5.1 scientific package (especially modules TRITON, NEWT, ORIGEN-S, ORIGEN-ARP) was used for the calculations presented in this paper. A typical Westinghouse 17x17 fuel assembly design was modeled at nominal PWR operating conditions. It was seen that U-Pu-Np MOX fuel with NpO2 and PuO2 representing 11.5wt% of the total MOX fuel would be similar to standard MOX fuel in which PuO2 is 9wt% of the fuel. The reactivity, isotopic composition, and neutron and ? sources, and the decay heat details for the discharged MOX fuel are presented and discussed in this paper.

  15. Global plutonium management: A security option

    SciTech Connect (OSTI)

    Sylvester, K.W.B.

    1998-12-31T23:59:59.000Z

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

  16. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    SciTech Connect (OSTI)

    Wishau, R.

    1998-05-01T23:59:59.000Z

    Molten salt oxidation (MSO) is proposed as a {sup 238}Pu waste treatment technology that should be developed for volume reduction and recovery of {sup 238}Pu and as an alternative to the transport and permanent disposal of {sup 238}Pu waste to the WIPP repository. In MSO technology, molten sodium carbonate salt at 800--900 C in a reaction vessel acts as a reaction media for wastes. The waste material is destroyed when injected into the molten salt, creating harmless carbon dioxide and steam and a small amount of ash in the spent salt. The spent salt can be treated using aqueous separation methods to reuse the salt and to recover 99.9% of the precious {sup 238}Pu that was in the waste. Tests of MSO technology have shown that the volume of combustible TRU waste can be reduced by a factor of at least twenty. Using this factor the present inventory of 574 TRU drums of {sup 238}Pu contaminated wastes is reduced to 30 drums. Further {sup 238}Pu waste costs of $22 million are avoided from not having to repackage 312 of the 574 drums to a drum total of more than 4,600 drums. MSO combined with aqueous processing of salts will recover approximately 1.7 kilograms of precious {sup 238}Pu valued at 4 million dollars (at $2,500/gram). Thus, installation and use of MSO technology at LANL will result in significant cost savings compared to present plans to transport and dispose {sup 238}Pu TRU waste to the WIPP site. Using a total net present value cost for the MSO project as $4.09 million over a five-year lifetime, the project can pay for itself after either recovery of 1.6 kg of Pu or through volume reduction of 818 drums or a combination of the two. These savings show a positive return on investment.

  17. Discovery of plutonium-based superconductivity

    SciTech Connect (OSTI)

    Sarrao, John L.,; Thompson, J. D. (Joe David); Moreno, N. O.; Morales, L. A. (Luis A.); Wastin, F. (Franck); Rebizant, J.; Boulet, P.; Colineau, E.; Lander, G. H.

    2002-01-01T23:59:59.000Z

    The discovery of superconductivity in single crystals of PuCoGa{sub 5} with transition temperature T{sub c}=18.5 K is discussed. The existing data lead to the speculation that the superconductivity in PuCoGa{sub 5} may be unconventional. In such a scenario the properties of PuCoGa{sub 5} would be intermediate between those of isostructural UCoGa{sub 5} and CeCoIn{sub 5}, more heavily studied f-electron materials.

  18. Radionuclide concentrations in honey bees from Area G at TA-54 during 1997. Progress report

    SciTech Connect (OSTI)

    Haarmann, T.K.; Fresquez, P.R.

    1998-07-01T23:59:59.000Z

    Honey bees were collected from two colonies located at Los Alamos National Laboratory`s Area G, Technical Area 54, and from one control (background) colony located near Jamez Springs, NM. Samples were analyzed for the following: cesium ({sup 137}Cs), americium ({sup 241}Am), plutonium ({sup 238}Pu and {sup 239,240}Pu), tritium ({sup 3}H), total uranium, and gross gamma activity. Area G sample results from both colonies were higher than the upper (95%) level background concentration for {sup 238}Pu and {sup 3}H.

  19. Seaborg's Plutonium ?

    E-Print Network [OSTI]

    Norman, Eric B; Telhami, Kristina E

    2014-01-01T23:59:59.000Z

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

  20. Seaborg's Plutonium?

    E-Print Network [OSTI]

    Eric B. Norman; Keenan J. Thomas; Kristina E. Telhami

    2015-02-17T23:59:59.000Z

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

  1. PLUTONIUM SOLUBILITY IN SIMULATED SAVANNAH RIVER SITE WASTE SOLUTIONS

    SciTech Connect (OSTI)

    Rudisill, T.; Hobbs, D.; Edwards, T.

    2010-09-27T23:59:59.000Z

    To address the accelerated disposition of the supernate and salt portions of Savannah River Site (SRS) high level waste (HLW), solubility experiments were performed to develop a predictive capability for plutonium (Pu) solubility. A statistically designed experiment was used to measure the solubility of Pu in simulated solutions with salt concentrations and temperatures which bounded those observed in SRS HLW solutions. Constituents of the simulated waste solutions included: hydroxide (OH{sup -}), aluminate (Al(OH){sub 4}{sup -}), sulfate (SO{sub 4}{sup 2-}), carbonate (CO{sub 3}{sup 2-}), nitrate (NO{sub 3}{sup -}), and nitrite (NO{sub 2}{sup -}) anions. Each anion was added to the waste solution in the sodium form. The solubilities were measured at 25 and 80 C. Five sets of samples were analyzed over a six month period and a partial sample set was analyzed after nominally fifteen months of equilibration. No discernable time dependence of the measured Pu concentrations was observed except for two salt solutions equilibrated at 80 C which contained OH{sup -} concentrations >5 mol/L. In these solutions, the Pu solubility increased with time. This observation was attributed to the air oxidation of a portion of the Pu from Pu(IV) to the more soluble Pu(V) or Pu(VI) valence states. A data driven approach was subsequently used to develop a modified response surface model for Pu solubility. Solubility data from this study and historical data from the literature were used to fit the model. The model predicted the Pu solubility of the solutions from this study within the 95% confidence interval for individual predictions and the analysis of variance indicated no statistically significant lack of fit. The Savannah River National Laboratory (SRNL) model was compared with predicted values from the Aqueous Electrolyte (AQ) model developed by OLI Systems, Inc. and a solubility prediction equation developed by Delegard and Gallagher for Hanford tank waste. The agreement between measured or values predicted by the SRNL model and values predicted by the OLI AG model was very poor. The much higher predicted concentrations by the OLI AQ model appears to be the result of the model predicting the predominate Pu oxidation state is Pu(V) which is reported as unstable below sodium hydroxide (NaOH) concentrations of 6 M. There was very good agreement between the predicted Pu concentrations using the SRNL model and the model developed by Delegard and Gallagher with the exception of solutions that had very high OH{sup -} (15 M) concentrations. The lower Pu solubilities in these solutions were attributed to the presence of NO{sub 3}{sup -} and NO{sub 2}{sup -} which limit the oxidation of Pu(IV) to Pu(V).

  2. Relation between two twisted inverse image pseudofunctors in ...

    E-Print Network [OSTI]

    2014-10-17T23:59:59.000Z

    Srikanth B. Iyengar, Joseph Lipman and Amnon Neeman ..... compactification theorem, to wit, that any map f in E factors as pu where p is proper and u is ...... and its interaction with algebraic geometry (Grenoble-Lyon, 2001), Contemp. Math.

  3. State-of-the-Art Highly Insulating Window Frames - Research and Market Review

    E-Print Network [OSTI]

    Gustavsen, Arild

    2008-01-01T23:59:59.000Z

    analysis investigation of a PVC window frame naturally agedThermix / TGI-wave 1.23 x 1.48 PVC profile with PUR (? =TOPLINE Plus Rahmenmaterial: PVC- Profile, Kammern mit PU

  4. Synthesis and characterization of Magnetic Nanoparticles and Their Reinforcement in Polyurethane Film

    E-Print Network [OSTI]

    Zheng, Yufeng

    in biomedical field, like coatings on cardiovascular stents. Introduction Magnetic nanoparticles show remarkableSynthesis and characterization of Magnetic Nanoparticles and Their Reinforcement in Polyurethane: magnetite nanoparticles, synthesis, PU composite films Abstract: Magnetic nanoparticles have attracted

  5. http://www.arias.cnrs.fr/ http://www.univ-lyon1.fr/ Fanny Lignon

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    for Mollie », regroupe les informations qu'il a pu réunir sur un soldat extraordinaire, qu'il n'a jamais vu'utilisa comme interprète et réussit à décourager 568 soldats italiens qui se rendirent sans condition. Notes

  6. COMIT FRANAIS D'HISTOIRE DE LA GOLOGIE -Troisime srie -

    E-Print Network [OSTI]

    Boyer, Edmond

    uniquement inspirée par des suintements d'hydrocarbures qui s'étaient manifestés à plusieurs reprises aux connue qui aurait pu fonctionner comme un piège à hydrocarbures. Poursuivant son tour de France, l

  7. Dynamic homology and phylogenetic systematics: a unified approach using POY

    E-Print Network [OSTI]

    Wheeler, Ward C.; Aagesen, Lone; Arango, Claudia P.; Faivovich, Juliá n; Grant, Taran; D’ Haese, Cyrille; Janies, Daniel; Smith, William Leo; Varó n, André s; Giribet, Gonzalo

    2006-01-01T23:59:59.000Z

    the relative 9 G ui de lin es 1 Da ta 11 In sta lla tio n 3 A na ly sis 13 In pu t/ O ut pu t 14 T ut or ia ls 15 C om m an ds 7 Ev al ua tio n 8 Pa ra lle l8 Chapter 1 merits of alternative historical expl anations. All observations are consis- tent... ta 11 In sta lla tio n 3 A na ly sis 13 In pu t/ O ut pu t 14 T ut or ia ls 15 C om m an ds 7 Ev al ua tio n 8 Pa ra lle lData 9 and can include behavior, while molecular data consists of nucleic acids, proteins, and genomes. This distinction...

  8. Investigations on the sediment chronology and trace metal accumulation in Sabine-Neches estuary, Beaumont, Texas

    E-Print Network [OSTI]

    Ravichandran, Mahalingam

    1994-01-01T23:59:59.000Z

    geochronology of sediments and reconstruction of the history of trace metal inputs into this shallow estuarine environment was possible because the 239,240pu profiles closely tracked the bomb fallout history into the environment. The sedimentation rate...

  9. Molten salt fuels with high plutonium solubility

    DOE Patents [OSTI]

    Moir, Ralph W; Turchi, Patrice E.A.; Shaw, Henry F; Kaufman, Larry

    2013-08-13T23:59:59.000Z

    The present invention includes a composition of LiF--ThF.sub.4--UF.sub.4--PuF.sub.3 for use as a fuel in a nuclear engine.

  10. Pour obtenir le grade de DOCTEUR DE L'UNIVERSIT DE GRENOBLE

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    intégrée. Parmi eux, notre cher Hervé ACHARD, qui a su assurer la partie filière de notre projet. Merci rencontre très prolifique où j'ai pu me faire de nombreux amis et collaborateurs tels que Jean-Pierre

  11. Reasoning with cause and effect

    E-Print Network [OSTI]

    Judea Pearl

    2011-01-01T23:59:59.000Z

    and R . A . Kowalski. Abduction compared with nega- tion asthe following three steps: 1. Abduction—update P(u) by thenecessary for supporting abduction; if we were to use one-

  12. Design and fabrication of physiologic tissue scaffolds using projection-micro-stereolithography

    E-Print Network [OSTI]

    Brickman Raredon, Micha Sam

    2014-01-01T23:59:59.000Z

    Recent advances in material processing are presenting groundbreaking opportunities for biomedical engineers. Projection-micro-stereolithography, or PuSL, is an additive manufacturing technique in which complex parts are ...

  13. Microsoft PowerPoint - MOX Adventure_Reactor Subcommittee_Tamara...

    National Nuclear Security Administration (NNSA)

    MOX Fuel at Duke Energy MOX Fuel and NMMSS Page 3 MOX Fuel - General MOX fuel pellets from former weapons plutonium Blend of 5% PuO 2 with 95% depleted UO 2 Like...

  14. Updated September 1, 2013 ?David Reide Corbett

    E-Print Network [OSTI]

    Corbett, D. Reide

    and Ecosystem Sciences, Dept. of Geology, Tulane University · Uranium cycling and sediment transport, and 137Cs, extraction of 239Pu from seawater, and hydrological studies of coastal regions; supervise

  15. Plutonium dissolution process

    DOE Patents [OSTI]

    Vest, M.A.; Fink, S.D.; Karraker, D.G.; Moore, E.N.; Holcomb, H.P.

    1994-01-01T23:59:59.000Z

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

  16. ARTICLE doi:10.1038/nature10423 Deep sequencing reveals 50 novel genes

    E-Print Network [OSTI]

    Cai, Long

    Zecha3 , Marzieh Mohseni1 , Lucia Pu¨ttmann3 , Leyla Nouri Vahid1 , Corinna Jensen3 , Lia Abbasi Moheb1,3 , Melanie Bienek3 , Farzaneh Larti1 , Ines Mueller3 , Robert Weissmann3 , Hossein Darvish1 , Klaus Wrogemann

  17. antigen-receptor interaction requirement: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    than those with two positive alleles (FY*B A Maestre; Carlos Muskus; Victoria Duque; Olga Agudelo; Pu Liu; Akihide Takagi; Francis B; John H. Adams; Kim Lee Sim; Stephen L....

  18. Diego Pérez de Luján, Relación de la expedición de Antonio de Espejo a Nuevo México, 1582-1583

    E-Print Network [OSTI]

    Craddock, Jerry R.; De Marco, Barbara

    2013-01-01T23:59:59.000Z

    poblado | el Rio del | Norte §que pu- | sieron por nombre |los rios 15 que diçen de el Norte con aquel por donde ybamostres leguas hasta el Rio del Norte y en el ca- 30 mino nos

  19. Advanced DOE-2 Calibration Procedures: A Technical Reference Manual

    E-Print Network [OSTI]

    Bou-Saada, T. E.

    1994-01-01T23:59:59.000Z

    toolkit). 675 Massachusetts Ave., Cambridge, MA 02139. Feuermann, D. and W. Kempton. 1987. ARCHIVE: Software for management of field data. The Center for Energy and Environmental Studies. Princeton University, PU/CEES Report No. 216. (June). LBL. 1980. DOE...

  20. Class Year Last Name First Name Project Title 2008 Avetisyan Marina Genomics Analysis of Sexually Dimorphic Gene Expression in the Developing and Adult Murine

    E-Print Network [OSTI]

    Weaver, Harold A. "Hal"

    , and Russian Modernism 2008 Keyvan Nina PU.1 and GATA Factor Inter-regulation in Mast Cell Differentiation 2008-Cameroon Pipeline: A Case Study on Perceived Doubts and Benefits in the Ngalaba Village 2008 Lu Sophie Olympian

  1. 2.

    E-Print Network [OSTI]

    SOA

    2007-05-23T23:59:59.000Z

    May 1, 2007 ... Let p be the true probability of the stock going up. Thus,. puS + (1 – p)dS ..... which, by taking logarithms, are equivalent to. 0.0571. (2) (2) 0.04 ?.

  2. advanced non-destructive assay: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    such as 235U and 239Pu. The neutrons from a pulsed, 14-MeV ne... American Society for Testing and Materials. Philadelphia 2009-01-01 3 A non-destructive transformer oil tester MIT...

  3. Fusion-breeder program

    SciTech Connect (OSTI)

    Moir, R.W.

    1982-11-19T23:59:59.000Z

    The various approaches to a combined fusion-fission reactor for the purpose of breeding /sup 239/Pu and /sup 233/U are described. Design aspects and cost estimates for fuel production and electricity generation are discussed. (MOW)

  4. Data completion method for the characterization of sound source in confined domain

    E-Print Network [OSTI]

    Boyer, Edmond

    formulation [4]. Then, vectors p and np are defined by: p = pm pu np = npm npu , where pm and mpm is a vector which is related to the geometry and to the measured acous- tic quantities pm and npm .

  5. MATHEMATICAL MODELING OF THE BEHAVIOR OF GEOTHERMAL SYSTEMS UNDER EXPLOITATION

    E-Print Network [OSTI]

    Bodvarsson, G.S.

    2010-01-01T23:59:59.000Z

    U. S. Department of Energy, Geothermal direct h e a t a p pU S Department of Energy, Geothermal Energy Division, 87,homes are heated by geothermal energy, and there are plans t

  6. aqueous waste treatment: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    z 20og (pendant 15 mn). On a pu ainsi montrer OF THE POLLUTING LOAD OF SWINE MANURE ACCORDING TO ITS VARIOUS PHYSICAL COMPONENTS. STUDY OF A SCREENING DEVICE for...

  7. advanced waste treatment: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    z 20og (pendant 15 mn). On a pu ainsi montrer OF THE POLLUTING LOAD OF SWINE MANURE ACCORDING TO ITS VARIOUS PHYSICAL COMPONENTS. STUDY OF A SCREENING DEVICE for...

  8. aerox waste treatment: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    z 20og (pendant 15 mn). On a pu ainsi montrer OF THE POLLUTING LOAD OF SWINE MANURE ACCORDING TO ITS VARIOUS PHYSICAL COMPONENTS. STUDY OF A SCREENING DEVICE for...

  9. Feasibility of breeding in hard spectrum boiling water reactors with oxide and nitride fuels

    E-Print Network [OSTI]

    Feng, Bo, Ph. D. Massachusetts Institute of Technology

    2011-01-01T23:59:59.000Z

    This study assesses the neutronic, thermal-hydraulic, and fuel performance aspects of using nitride fuel in place of oxides in Pu-based high conversion light water reactor designs. Using the higher density nitride fuel ...

  10. Summary - Small Column Ion Exchange (SCIX)Technology at the SRS

    Office of Environmental Management (EM)

    ETR R Un Baseline The Sm being The SC operat which Sr, and waste critical the SC deploy Specif exchan CST) CST, a (mono and so (RMF) maturi readin design moving The pu techni...

  11. RTDS-Based Design and Simulation of Distributed P-Q Power Resources in Smart Grid

    E-Print Network [OSTI]

    Taylor, Zachariah David

    2014-01-01T23:59:59.000Z

    1.2 MW grid-connected solar panel at bus 8. The power outputis as in Figure 5.11 [104]. Solar Panel Power Injection (pu)The daily output of the solar panel during a cloudy day.

  12. Analysis of resuspension source area impacts at Rocky Flats surveillance air samplers S-7 and S-8, July 25-August 25, 1983 and September 8-October 4, 1983

    SciTech Connect (OSTI)

    Hammer, R.J.

    1984-01-01T23:59:59.000Z

    An on-going study at the Rocky Flats Plant is being used to evaluate resuspension source area contributions to Pu-239 concentrations at 2 of the samplers in the Plants air sampling network. Early results from the study indicate that Pu-239 concentration levels are being affected primarily by resuspension from a zone 150 meters east and west of the study samplers. Initial results have also shown that net transport of Pu-239 during the sampling period has been from the east toward the west, onto the plant proper. These early findings show that sources immediately east of the 2 samplers are responsible for most of the Pu-239 exposure at the samplers. 2 references, 1 figure, 4 tables.

  13. The Role of Colloids in the Transport of Plutonium and Americium: Implications for

    SciTech Connect (OSTI)

    Kersting, A B

    2003-09-17T23:59:59.000Z

    Colloids are small particulates (ranging in size from 1 to 0.001 micron) composed of inorganic and organic material and found in all natural water. Due to their small size, they have the ability to remain suspended in water and transported. Small amounts of plutonium (Pu) and americium (Am) can adsorb (attach) to colloids, and/or form colloidal-sized polymers and migrate in water. At Rocky Flats Environmental Technology Site (RFETS) sedimentation and resuspension of particulates and colloids in surface waters represent the dominant process for Pu and Am migration. The amount of Pu and Am that can be transported at RFETS has been quantified in the Pathway Analysis Report. The Pathway Analysis Report shows that the two dominant pathways for Pu and Am transport at RFETS are air and surface water. Shallow groundwater and biological pathways are minor.

  14. RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01T23:59:59.000Z

    2000. Total fuel mix is 11% MOX + 89% U0 fuel with PuRadionuclide H U0 Fuel U0 + MOX Fuel 14C Kr I llO Other

  15. Radial Power Profile of MOX and LEU Fuel Pellet Versus Burnup

    SciTech Connect (OSTI)

    Chang, Gray S.; Pedersen, Robert C. [INEEL - Idaho National Engineering and Environmental Laboratory, Idaho Falls, ID 83415 (United States)

    2002-07-01T23:59:59.000Z

    One of challenge to burn the WG-Pu in Mixed Oxide (MOX) fuel in light water reactors (LWR) is to demonstrate that the differences between WG-MOX, RG-MOX, and LWR LEU fuel are minimal, and therefore, the commercial MOX and LEU fuel experience base is applicable. The MCWO-calculated Radial Power Profile of LEU, Weapons Grade-MOX and Reactor Grade-MOX fuel pellets at various burnups are similar toward the end of life (50 GWd/t). Therefore, the LEU fuel performance evaluation code - FRAPCON-3 with modifications, such as, the detailed fission power profiles versus burnup, can be used in the MOX fuel pellet performance analysis. MCWO also calculated the {sup 240}Pu/Pu ratio in WG-MOX versus burnup, which reaches an average of 31.25% at discharged burnup of 50 GWd/t. It meets the spent fuel standard for WG-Pu disposition in LWR. (authors)

  16. IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

    SciTech Connect (OSTI)

    Gilles Youinou; Andrea Alfonsi

    2012-03-01T23:59:59.000Z

    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.

  17. Selective broadcast of fenoxycarb bait on fire ant infested prairie: effect on native ant community

    E-Print Network [OSTI]

    Morris, John Robert

    1994-01-01T23:59:59.000Z

    . Sample sites within managemmzt units are numbered. Letters indicate where treatment plot samples were located (PI = infested, PM = managed, PU = uninfested). 23 100 o IL COLLECTION LITE 10 12 13 14 " 16 CLYMER MEADOW Fig. 2. Distribution...

  18. Feasibility Study of a Portable Coupled 3He Detector with LaBr3 Gamma Scintillator for Field Identification and Quantification of Nuclear Material

    E-Print Network [OSTI]

    Strohmeyer, Daniel C.

    2010-07-14T23:59:59.000Z

    spectroscopy and neutron coincidence counting in a single instrument allows for direct measurement of plutonium (Pu) mass without need for assumptions or operator declarations. A combined neutron-gamma instrument was designed for use in characterizing...

  19. Microsoft PowerPoint - 5_IRENE_WU_NMMSS_2014_NSTS Update.ppt...

    National Nuclear Security Administration (NNSA)

    Nationally Tracked Sources Nationally Tracked Sources Examples Neutron sources for reactor startup (e.g., Pu-238Be) Calibration sources (e.g., Cs-137) Where We Are Today...

  20. The phonon density of states of (alpha) and (delta)-Plutonium by inelastic x-ray scattering

    SciTech Connect (OSTI)

    Manley, M E; Said, A; Fluss, M J; Wall, M; Lashley, J C; Alatas, A; Moore, K T

    2008-10-08T23:59:59.000Z

    Inelastic x-ray scattering measurements of the phonon density of states (DOS) were performed on polycrystalline samples of pure {alpha}-Pu and {delta}-Pu{sub 0.98}Ga{sub 0.02} at room temperature. The heat capacity of {alpha}-Pu is well reproduced by contributions calculated from the measured phonon DOS plus conventional thermal expansion and electronic contributions, showing that {alpha}-Pu is a 'well-behaved' metal in this regard. A comparison of the phonon DOS of the two phases at room temperature surprised us in that the vibrational entropy difference between them is only a quarter of the total entropy difference expected from known thermodynamic measurements. The missing entropy is too large to be accounted for by conventional electronic entropy and evidence from the literature rules out a contribution from spin fluctuations. Possible alternative sources for the missing entropy are discussed.

  1. EVALUATION OF FLOWSHEET CHANGES FOR THE HIGHLY ENRICHED URANIUM BLENDDOWN PROGRAM

    SciTech Connect (OSTI)

    Crowder, M.; Rudisill, T.; Laurinat, J.; Mickalonis, J.

    2007-10-22T23:59:59.000Z

    H Canyon is considering a flowsheet change for Plutonium (Pu) Contaminated Scrap (PuCS) material. The proposed change is to route dissolved PuCS material directly to a uranium (U) storage tank. As a result, the PuCS solution will bypass Head End and First U Cycle, and will be purified by solvent extraction in Second U Cycle. The PuCS solution contains appreciable amounts of boron (B) and fluoride (F{sup -}), which are currently at trace levels in the U storage tank. Though unlikely, if the B concentration in the U storage tank were to reach 1.8 g B/g U, the entire contents of the U storage tank would likely require a second pass through Second U Cycle to provide sufficient decontamination to meet the Tennessee Valley Authority (TVA) Blend Grade Highly Enriched Uranium (HEU) specification for B, which is 30 {micro}g/g U. In addition, Second U Cycle is expected to provide sufficient decontamination of F{sup -} and Pu regardless of the amount of PuCS solution sent to the storage tank. Though aluminum (Al) is not present in the PuCS solution, B can be credited as a complexant of F{sup -}. Both stability constants from the literature and Savannah River National Laboratory (SRNL) corrosion studies were documented to demonstrate that B complexation of F{sup -} in nitric acid solutions is sufficient to prevent excessive corrosion. Though B and Al complex F{sup -} to a similar degree, neither completely eliminates the presence of free F{sup -} in solution. Therefore, a limited amount of corrosion is expected even with complexed F{sup -} solutions. Tanks maintained at ambient temperature are not expected to experience significant corrosion. However, the Low Activity Waste (LAW) evaporators may be subjected to a corrosion rate of about 25 mils per year (mpy) as they reach their highest F{sup -} concentrations. The feed adjustment evaporator would only be subjected to the corrosion rate of about 25 mpy in the latter stages of the PuCS campaign. An issue that must be addressed as part of the proposed PuCS flowsheet change is that B has limited solubility in concentrated nitric acid solutions. As the proposed PuCS campaign progresses, the B concentration will increase in the U storage tank, in Second U Cycle feed, and in the 1DW stream sent to the LAW evaporators. Limitations on the B concentration in the LAW evaporators will be needed to prevent formation of boron-containing solids.

  2. Annual progress Report on research related to our research project “Stabilization of Plutonium in Subsurface Environments via Microbial Reduction and Biofilm Formation” funded by the Environmental Remediation Sciences Division (ERSD)

    SciTech Connect (OSTI)

    New, Mary

    2006-06-01T23:59:59.000Z

    The overarching goal of this research project is to investigate and optimize the mechanisms for in situ immobilization of Pu species by naturally-occurring bacteria. Specific research objectives are: (a) investigate the mechanism of bacterial accumulation and immobilization of plutonium species by biofilm formation under aerobic conditions and (b) to demonstrate the direct and indirect stabilization of Pu via dissimilatory reduction by Geobacter metallireducens.

  3. Selected papers for global `95 concerning plutonium

    SciTech Connect (OSTI)

    Sutcliffe, W.G. [ed.

    1996-06-14T23:59:59.000Z

    This report contains selected papers from the Global `95 Conference ``Evaluation of Emerging Nuclear Fuel Cycle Systems,`` held in Versailles, Sept. 11-14, 1995. The 11 papers in Part I are from ``Benefits and Risks of Reprocessing`` sessions. The 7 papers in Part II are some of the more interesting poster papers that relate to the use of Pu for power generation. Finally, the 3 papers are on the topic of management and disposition of Pu from retired nuclear weapons.

  4. Fluorescence Imaging for Nuclear Arms Control Verification 

    E-Print Network [OSTI]

    Feener, Jessica S

    2014-08-14T23:59:59.000Z

    into a supercritical configuration. A thermonuclear device combines an implosion device, known as the primary, with a secondary fusion device composed of uranium and lithium deuteride. Most modern weapons contain a thermonuclear warhead. Gun... grade Pu typically contains more than 90% 239Pu.10 Additionally, the fissile material must be in metallic form. There are many different nuclear warhead designs, but there are three general warhead types: gun-type, implosion and thermonuclear. In a...

  5. Characterization of Polyurethane at Multiple Scales for Erosion Mechanisms Under Sand Particle Impact

    E-Print Network [OSTI]

    Sigamani, Nirmal

    2010-07-14T23:59:59.000Z

    B (501). ................ 9 Figure 1.10 Schematic of the micro structure of the polyurethane ................................ 9 Figure 1.11 Network structure in PU when the chain extender has: (a) even number...-scale, multi-phase separation in PU. In the micro scale, the soft segment matrix is mixed in the hard segment spherulites whereas in the nano scale the hard segment domains are dispersed in the soft segment matrix. 9 Figure 1.9 AFM images...

  6. DKCM0OO4 2. D RisO-M-2856

    E-Print Network [OSTI]

    was successfully used for collecting 239 »240 Pu from 200 litres seawater by coprecipitation with 16 g FeS04.7H20 fresh NaN02 to keep Pu** valence for uranium decontamination. The system of the column is changed from 8 of plutonium for a 200 litres-seawater sample is 60-80%. The resolution of the electroplated thin source

  7. EE 581 Power Systems Admittance Matrix: Development, Direct and Iterative

    E-Print Network [OSTI]

    Wedeward, Kevin

    of calculations, admittance is used (Y) Ohm's Law: V=IR Complex: V=IZ = 1 = (Siemens) is symmetric with respect to the reference bus. Step 2a: Source transform any voltage sources in parallel to an equivalent: Transformer Data: Name From Bus To Bus Z' (p.u.) Y'/2 (p.u.) TL1 3 4 0.077 + 0.31j 0.16j TL2 3 5 0.039 + 0.15j

  8. Nuclear Data Sheets for A=228

    SciTech Connect (OSTI)

    Abusaleem, Khalifeh

    2014-02-01T23:59:59.000Z

    The evaluated spectroscopic data are presented for known nuclides of mass 228 (Ac, At, Fr, Np, Pa, Pu, Ra, Rn, Th, and U). Excited states in {sup 228}At, {sup 228}Rn, {sup 228}Fr, {sup 228}Np, and {sup 228}Pu have not been identified as yet. Significant amounts of new data have been added since the last evaluation of A=228 nuclides. This work supersedes earlier full evaluations of A=228 published by 1997Ar08.

  9. Light-water reactors: preliminary safety and environmental information document. Volume I

    SciTech Connect (OSTI)

    Not Available

    1980-01-01T23:59:59.000Z

    Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle (Pu/ThO/sub 2/ Burner).

  10. SRS vitrification studies in support of the U.S. program for disposition of excess plutonium

    SciTech Connect (OSTI)

    Wicks, G.G.; McKibben, J.M.; Plodinec, M.J.; Ramsey, W.G.

    1995-09-01T23:59:59.000Z

    Many thousands of nuclear weapons are being retired in the U.S. and Russian as a result of nuclear disarmament activities. These efforts are expected to produce a surplus of about 50 MT of weapons grade plutonium (Pu) in each country. In addition to this inventory, the U.S. Department of Energy (DOE) has more than 20 MT of Pu scrap, residue, etc., and Russian is also believed to have at least as much of this type of material. The entire surplus Pu inventories in the U.S. and Russian present a clear and immediate danger to national and international security. It is important that a solution be found to secure and manage this material effectively and that such an effort be implemented as quickly as possible. One option under consideration is vitrification of Pu into a safe, durable, accountable and proliferation-resistant form. As a result of decades to experience within the DOE community involving vitrification of a variety of hazardous and radioactive wastes, this existing technology can now be expanded to include mobilization of large amounts of Pu. This technology can then be implemented rapidly using the many existing resources currently available. An overall strategy to vitrify many different types of Pu will be already developed throughout the waste management community can be used in a staged Pu vitrification effort. This approach uses the flexible vitrification technology already available and can even be made portable so that it may be brought to the source and ultimately, used to produce a consistent and common borosilicate glass composition for the vitrified Pu. The final composition of this product can be made similar to nationally and internationally accepted HLW glasses.

  11. ICDERS July 23-29, 2011 UC-Irvine, CA * Corresponding author: kuhl2@llnl.gov 1

    E-Print Network [OSTI]

    Bell, John B.

    ) combustion code [3,4]. 2 Model Conservation Laws The Model is based on the Eulerian multi-phase conservation-dynamic conservation laws: Mass: t + ( u) = s (1) Momentum: t u+ (uu+ p) = sv - fs (2) Energy: t E + (uE + pu) = - qs + sEs - fs v (3) Where , p,U represent the gas density, pressure and specific internal energy, u

  12. Reliability Engineering Approach to Probabilistic Proliferation Resistance Analysis of the Example Sodium Fast Reactor Fuel Cycle Facility

    E-Print Network [OSTI]

    Cronholm, Lillian Marie

    2012-10-19T23:59:59.000Z

    HEU (235U ? 20%) 25 kg Indirect use nuclear material U (235U depleted U) Th 20 tons a For Pu containing less than 80% 238Pu. b Including low enriched, natural and depleted uranium... Highly Enriched Uranium HM Heavy Metal IAEA International Atomic Energy Agency IC Product Prep Injection Caster Furnace IS&NP International Safeguards and Nonproliferation KMP Key Measurement Point MBA Material Balance Area MCNP Monte...

  13. Development of a Phosphate Ceramic as a Host for Halide-contaminated Plutonium Pyrochemical Reprocessing Wastes

    SciTech Connect (OSTI)

    Metcalfe, Brian; Fong, Shirley K.; Gerrard, Lee A.; Donald, Ian W.; Strachan, Denis M.; Scheele, Randall D.

    2007-03-31T23:59:59.000Z

    The presence of halide anions in four types of wastes arising from the pyrochemical reprocessing of plutonium required an immobilization process to be developed in which not only the actinide cations but also the halide anions were immobilized in a durable waste form. At AWE, we have developed such a process using Ca3(PO4)2 as the host material. Successful trials of the process with actinide- and Cl-bearing Type I waste were carried out at PNNL where the immobilization of the waste in a form resistant to aqueous leaching was confirmed. Normalized mass losses determined at 40°C and 28 days were 12 x 10-6 g?m-2 and 2.7 x 10-3 g?m-2 for Pu and Cl, respectively. Accelerated radiation-induced damage effects are being determined with specimens containing 238Pu. No changes in the crystalline lattice have been detected with XRD after the 239Pu equivalent of 400 years ageing. Confirmation of the process for Type II waste (a oxyhydroxide-based waste) is currently underway at PNNL. Differences in the ionic state of Pu in the four types of waste have required different surrogates to be used. Samarium chloride was used successfully as a surrogate for both Pu(III) and Am(III) chlorides. Initial investigations into the use of HfO2 as the surrogate for Pu(IV) oxide in Type II waste indicated no significant differences.

  14. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    SciTech Connect (OSTI)

    Wishau, R.; Ramsey, K.B.; Montoya, A.

    1998-12-31T23:59:59.000Z

    This paper presents the technical and economic feasibility of molten salt oxidation technology as a volume reduction and recovery process for {sup 238}Pu contaminated waste. Combustible low-level waste material contaminated with {sup 238}Pu residue is destroyed by oxidation in a 900 C molten salt reaction vessel. The combustible waste is destroyed creating carbon dioxide and steam and a small amount of ash and insoluble {sup 2328}Pu in the spent salt. The valuable {sup 238}Pu is recycled using aqueous recovery techniques. Experimental test results for this technology indicate a plutonium recovery efficiency of 99%. Molten salt oxidation stabilizes the waste converting it to a non-combustible waste. Thus installation and use of molten salt oxidation technology will substantially reduce the volume of {sup 238}Pu contaminated waste. Cost-effectiveness evaluations of molten salt oxidation indicate a significant cost savings when compared to the present plans to package, or re-package, certify and transport these wastes to the Waste Isolation Pilot Plant for permanent disposal. Clear and distinct cost advantages exist for MSO when the monetary value of the recovered {sup 238}Pu is considered.

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

    SciTech Connect (OSTI)

    Forsberg, C. [Massachusetts Institute of Technology, 77 Massachusetts Ave. Cambridge, MA 20129 (United States)

    2013-07-01T23:59:59.000Z

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

  16. Corrosion testing of a plutonium-loaded lanthanide borosilicate glass made with Frit B.

    SciTech Connect (OSTI)

    Ebert, W. L.; Chemical Engineering

    2006-09-30T23:59:59.000Z

    Laboratory tests were conducted with a lanthanide borosilicate (LaBS) glass made with Frit B and added PuO2 (the glass is referred to herein as Pu LaBS-B glass) to measure the dependence of the glass dissolution rate on pH and temperature. These results are compared with the dependencies used in the Defense HLW Glass Degradation Model that was developed to account for HLW glasses in total system performance assessment (TSPA) calculations for the Yucca Mountain repository to determine if that model can also be used to represent the release of radionuclides from disposed Pu LaBS glass by using either the same parameter values that are used for HLW glasses or parameter values specific for Pu LaBS glass. Tests were conducted by immersing monolithic specimens of Pu LaBS-B glass in six solutions that imposed pH values between about pH 3.5 and pH 11, and then measuring the amounts of glass components released into solution. Tests were conducted at 40, 70, and 90 C for 1, 2, 3, 4, and 5 days at low glass-surface-area-to-solution volume ratios. As intended, these test conditions maintained sufficiently dilute solutions that the impacts of solution feedback effects on the dissolution rates were negligible in most tests. The glass dissolution rates were determined from the concentrations of Si and B measured in the test solutions. The dissolution rates determined from the releases of Si and B were consistent with the 'V' shaped pH dependence that is commonly seen for borosilicate glasses and is included in the Defense HLW Glass Degradation Model. The rate equation in that model (using the coefficients determined for HLW glasses) provides values that are higher than the Pu LaBS-B glass dissolution rates that were measured over the range of pH and temperature values that were studied (i.e., an upper bound). Separate coefficients for the rate expression in acidic and alkaline solutions were also determined from the test results to model Pu LaBS-B glass dissolution directly. The releases of Gd, Hf, and Pu from the glass were also measured. The release of Pu was significantly less than Si at all temperatures and pH values (on a normalized basis). More Gd than Pu or Hf was released from the glass in acidic solutions, but more Pu than Gd or Hf was released in alkaline solutions. Almost all of the released Gd remained in solution in tests conducted in Teflon vessels, whereas about half of the released Pu and Hf became fixed to the Teflon. In tests conducted in Type 304L stainless steel vessels, most of the released Gd, Hf, and Pu became fixed to the steel. The aqueous concentrations of Gd, Hf, and Pu decreased from about 2 x 10{sup -5}, 2 x 10{sup -8}, and 1 x 10{sup -7} M in tests solutions near pH 3.7 to about 1 x 10{sup -9}, 8 x 10{sup -10}, and 1 x 10{sup -8} M in test solutions near pH 10.8, respectively, in the 90 C tests in Teflon vessels (the solutions were not filtered prior to analysis). Vapor hydration tests (VHTs) were conducted at 120 and 200 C with Pu LaBS-B glass and SRL 418 glass, which was made to represent the HLW glass that will be used to macro-encapsulate LaBS glass within the waste form. Some VHTs were conducted with specimens of Pu LaBS-B and SRL 418 glasses that were in contact to study the effect of the solution generated as HLW glass dissolves on the corrosion behavior of Pu LaBS-B glass. Other VHTs were conducted in which the glasses were not in contact. The Pu LaBS-B glass is more durable than the HLW glass under these accelerating test conditions, even when the glasses are in contact. The presence of the SRL 418 glass did not promote the dissolution of the Pu LaBS-B glass significantly. However, Gd, Hf, and Pu were detected in alteration phases formed on the Pu LaBS-B glass surface and in (or on) phases formed by SRL 418 glass degradation, such as analcime. This indicates that Gd, Hf, and Pu were transported from the LaBS glass, through the water film formed on the specimens, and to the SRL 418 glass during the test. The disposition of the PuO{sub 2} inclusion phases as the Pu LaBS-B glass dissolved was not det

  17. Radionuclide contaminant analysis of small mammals at Area G, Technical Area 54, 1996 (with cumulative summary for 1994--1996)

    SciTech Connect (OSTI)

    Biggs, J.R.; Bennett, K.D.; Fresquez, P.R.

    1997-07-01T23:59:59.000Z

    Small mammals were sampled at two waste burial sites at Area G, Technical Area (TA) 54 and a control site within the proposed Area G expansion area in 1996 to (1) identify radionuclides that are present within rodent tissues at waste burial sites, (2) to compare the amount of radionuclide uptake by small mammals at waste burial sites to a control site, and (3) to identify the primary mode of contamination to small mammals, either through surface contact or ingestion/inhalation. Three composite samples of approximately five animals per sample were collected at each site. Pelts and carcasses of each animal were separated and analyzed independently. Samples were analyzed for {sup 241}Am, {sup 90}Sr, {sup 238}Pu, {sup 239}Pu, total U, {sup 137}Cs, and {sup 3}H. Higher levels of total U, {sup 241}Am, {sup 238}Pu, and {sup 239}Pu were detected in pelts as compared to the carcasses of small mammals at TA-54. Concentrations of other measured radionuclides in carcasses were nearly equal to or exceeded the mean concentrations in the pelts. Due to low sample sizes in total number of animals captured, statistical analysis to compare site to site could not be conducted. However, mean concentrations of total U, {sup 238}Pu, {sup 239}Pu, and {sup 137}Cs in rodent carcasses were higher at Site 1 than site 2 or the Control Site and {sup 241}Am was higher at Site 2 than Site 1 or the Control Site.

  18. INTERPRETATION OF AT-LINE SPECTRA FROM AFS-2 BATCH #3 FERROUS SULFAMATE TREATMENT

    SciTech Connect (OSTI)

    Kyser, E.; O'Rourke, P.

    2013-12-10T23:59:59.000Z

    Spectra from the “at-line” spectrometer were obtained during the ferrous sulfamate (FS) valence adjustment step of AFS-2 Batch #3 on 9/18/2013. These spectra were analyzed by mathematical principal component regression (PCR) techniques to evaluate the effectiveness of this treatment. Despite the complications from Pu(IV), we conclude that all Pu(VI) was consumed during the FS treatment, and that by the end of the treatment, about 85% was as Pu(IV) and about 15% was as Pu(III). Due to the concerns about the “odd” shape of the Pu(IV) peak and the possibility of this behavior being observed in the future, a follow-up sample was sent to SRNL to investigate this further. Analysis of this sample confirmed the previous results and concluded that it “odd” shape was due to an intermediate acid concentration. Since the spectral evidence shows complete reduction of Pu(VI) we conclude that it is appropriate to proceed with processing of this the batch of feed solution for HB-Line including the complexation of the fluoride with aluminum nitrate.

  19. MA transmutation performance in the optimized MYRRHA

    SciTech Connect (OSTI)

    Malambu, E.; Van den Eynde, G.; Fernandez, R.; Baeten, P.; Ait Abderrahim, H. [SCK-CEN, Boeretang 200, BE-2400 Mol (Belgium)

    2013-07-01T23:59:59.000Z

    MYRRHA (multi-purpose hybrid research reactor for high-tech applications) is a multipurpose research facility currently being developed at SCK-CEN. It will be able to work in both critical and subcritical modes and, cooled by lead-bismuth eutectic. In this paper the minor actinides (MA) transmutation capabilities of MYRRHA are investigated. (Pu + Am, U) MOX fuel and (Np + Am + Cm, Pu) Inert Matrix Fuel test samples have been loaded in the central channel of the MYRRHA critical core and have been irradiated during five cycles, each one consisting of 90 days of operation at 100 MWth and 30 days of shutdown. The reactivity worth of the test fuel assembly was about 1.1 dollar. A wide range of burn-up level has been achieved, extending from 42 to 110 MWd/kg HM, the samples with lower MA-to-Pu ratios reaching the highest burn-up. This study has highlighted the importance of the initial MA content, expressed in terms of MA/Pu ratio, on the transmutation rate of MA elements. For (Pu + Am, U) MOX fuel samples, a net build-up of MA is observed when the initial content of MA is very low (here, 1.77 wt% MA/Pu) while a net decrease in MA is observed in the sample with an initial content of 5 wt%. This suggests the existence of some 'equilibrium' initial MA content value beyond which a net transmutation is achievable.

  20. Lawrence Livermore National Laboratory Working Reference Material Production Pla

    SciTech Connect (OSTI)

    Amy Wong; Denise Thronas; Robert Marshall

    1998-11-04T23:59:59.000Z

    This Lawrence Livermore National Laboratory (LLNL) Working Reference Material Production Plan was written for LLNL by the Los Alamos National Laboratory to address key elements of producing seven Pu-diatomaceous earth NDA Working Reference Materials (WRMS). These WRMS contain low burnup Pu ranging in mass from 0.1 grams to 68 grams. The composite Pu mass of the seven WRMS was designed to approximate the maximum TRU allowable loading of 200 grams Pu. This document serves two purposes: first, it defines all the operations required to meet the LLNL Statement of Work quality objectives, and second, it provides a record of the production and certification of the WRMS. Guidance provided in ASTM Standard Guide C1128-89 was used to ensure that this Plan addressed all the required elements for producing and certifying Working Reference Materials. The Production Plan was written to provide a general description of the processes, steps, files, quality control, and certification measures that were taken to produce the WRMS. The Plan identifies the files where detailed procedures, data, quality control, and certification documentation and forms are retained. The Production Plan is organized into three parts: a) an initial section describing the preparation and characterization of the Pu02 and diatomaceous earth materials, b) middle sections describing the loading, encapsulation, and measurement on the encapsulated WRMS, and c) final sections describing the calculations of the Pu, Am, and alpha activity for the WRMS and the uncertainties associated with these quantities.

  1. Plutonium Isotopes in the Terrestrial Environment at the Savannah River Site, USA: A Long-Term Study

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

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

    2015-02-03T23:59:59.000Z

    This work presents the findings of a long term plutonium study at Savannah River Site (SRS) conducted between 2003 and 2013. Terrestrial environmental samples were obtained at Savannah River National Laboratory (SRNL) in A-area. Plutonium content and isotopic abundances were measured over this time period by alpha spectrometry and three stage thermal ionization mass spectrometry (3STIMS). Here we detail the complete sample collection, radiochemical separation, and measurement procedure specifically targeted to trace plutonium in bulk environmental samples. Total plutonium activities were determined to be not significantly above atmospheric global fallout. However, the 238Pu/239+240Pu activity ratios attributed to SRS are above atmospheric global fallout ranges. The 240Pu/239Pu atom ratios are reasonably consistent from year to year and are lower than fallout, while the 242Pu/239Pu atom ratios are higher than fallout values. Overall, the plutonium signatures obtained in this study reflect a mixture of weapons-grade, higher burn-up, and fallout material. This study provides a blue print for long term low level monitoring of plutonium in the environment.

  2. Melting temperatures of the ZrO{sub 2}-MOX system

    SciTech Connect (OSTI)

    Uchida, T.; Hirooka, S.; Kato, M.; Morimoto, K. [Japan Atomic Energy Agency, 4-33, Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1194 (Japan); Sugata, H.; Shibata, K.; Sato, D. [Inspection Development Company, 4-33, Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1194 (Japan)

    2013-07-01T23:59:59.000Z

    Severe accidents occurred at the Fukushima Daiichi Nuclear Power Plant Units 1-3 on March 11, 2011. MOX fuels were loaded in the Unit 3. For the thermal analysis of the severe accident, melting temperature and phase state of MOX corium were investigated. The simulated coriums were prepared from 4%Pu-containing MOX, 8%Pu-containing MOX and ZrO{sub 2}. Then X-ray diffraction, density and melting temperature measurements were carried out as a function of zirconium and plutonium contents. The cubic phase was observed in the 25%Zr-containing corium and the tetragonal phase was observed in the 50% and 75%Zr-containing coria. The lattice parameter and density monotonically changed with Pu content. Melting temperature increased with increasing Pu content; melting temperature were estimated to be 2932 K for 4%Pu MOX corium and 3012 K for 8%Pu MOX corium in the 25%ZrO{sub 2}-MOX system. The lowest melting temperature was observed for 50%Zr-containing corium. (authors)

  3. Uncertainty Measurement for Trace Element Analysis of Uranium and Plutonium Samples by Inductively Coupled Plasma-Atomic Emission Spectrometry (ICP-AES) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS)

    SciTech Connect (OSTI)

    Gallimore, David L. [Los Alamos National Laboratory

    2012-06-13T23:59:59.000Z

    The measurement uncertainty estimatino associated with trace element analysis of impurities in U and Pu was evaluated using the Guide to the Expression of Uncertainty Measurement (GUM). I this evalution the uncertainty sources were identified and standard uncertainties for the components were categorized as either Type A or B. The combined standard uncertainty was calculated and a coverage factor k = 2 was applied to obtain the expanded uncertainty, U. The ICP-AES and ICP-MS methods used were deveoped for the multi-element analysis of U and Pu samples. A typical analytical run consists of standards, process blanks, samples, matrix spiked samples, post digestion spiked samples and independent calibration verification standards. The uncertainty estimation was performed on U and Pu samples that have been analyzed previously as part of the U and Pu Sample Exchange Programs. Control chart results and data from the U and Pu metal exchange programs were combined with the GUM into a concentration dependent estimate of the expanded uncertainty. Comparison of trace element uncertainties obtained using this model was compared to those obtained for trace element results as part of the Exchange programs. This process was completed for all trace elements that were determined to be above the detection limit for the U and Pu samples.

  4. Studies of Plutonium Aerosol Resuspension at the Time of the Maralinga Cleanup

    SciTech Connect (OSTI)

    Shinn, J

    2003-08-01T23:59:59.000Z

    At the former nuclear test site at Maralinga, South Australia, soil cleanup began in October 1996 with the objective to remove the potential for residual plutonium (Pu) exposures to the public. In this case the cleanup was to restore access to the closed test site. The proposed long-term land use was primarily to be a hunting area for Pitjantjatjara (Aboriginal) people, but also presumably to be available to the public who might have an interest in the history of the site. The long-term management objective for the site was to allow casual use, but to prohibit habitation. The goal of this study is to provide an evaluation of the Maralinga soil cleanup in terms of potential long-term public inhalation exposures to particulate Pu, and in terms of a contribution to planning and conducting any such soil Pu-cleanup. Such cleanups might be carried out for example, on the Nevada Test Site in the United States. For Pu that has been deposited on the soil by atmospheric sources of finely divided particles, the dominant exposure pathway to humans is by inhalation. Other exposure pathways are less important because the Pu particles become oxidized into a nearly insoluble form, do not easily enter into the food chain, nor are they significantly transferred through the intestine to the bloodstream should Pu become ingested. The purpose of this report is to provide results of the Pu resuspension measurements made before, during, and after the Pu cleanup at Maralinga, to compare these against similar measurements made elsewhere, and to interpret the results as they relate to potential long-term public exposures. (Exposures to Pu in dust plumes produced by mechanical disturbance during cleanup are considered short-term, unlikely to be significant for purposes of this report, and are not included). A considerable amount of research had been conducted at Maralinga by the Australian Radiation Laboratory, now the Australian Radiation Protection and Nuclear Safety Agency (ARPANSA), prior to the cleanup (Johnston et al, 1992, Williams 1993, Johnston et al 1993, Burns et al 1994, Burns et al 1995). ARPANSA staff made major contributions to delineate the areas with Pu in the soil, to determine the degree of secondary soil contamination by fission products from nuclear testing, to measure Pu resuspension by wind erosion of the undisturbed soil, and to prepare assessments of the human health risk from residual soil Pu. In addition, ARPANSA supported the Maralinga cleanup to assure compliance with criteria set by an independent technical advisory committee. During the cleanup ARPANSA monitored the residual Pu in the soil and certified that the cleanup was complete according to the criteria. It was not the reduction in potential inhalation exposure that usually was the main driver of the cleanup, but the requirement to also remove individual hot particles and fragments. It is the residual microscopic particles of Pu in the soil, however, that have the potential for long-term human exposure. The resuspension of respirable-size Pu particles has been studied with specialized equipment at the Nevada Test Site (Gilbert et al 1988a, Gilbert et al 1988b, Shinn et al 1989, and Shinn 1992), and at Bikini and Enewetak in the Marshall Islands (Shinn et al 1997). These efforts were in large part contributed by the Health and Ecological Assessment Division, University of California, Lawrence Livermore National Laboratory (LLNL). The study reported here is a collaboration between ARPANSA and LLNL, and was jointly supported by the United States Department of Energy, and the Commonwealth of Australia Department of Primary Industry and Energy.

  5. GLASS FABRICATION AND PRODUCT CONSISTENCY TESTING OF LANTHANIDE BOROSHILICATE FRIT X COMPOSITION FOR PLUTONIUM DISPOSITION

    SciTech Connect (OSTI)

    Marra, J

    2006-11-21T23:59:59.000Z

    The Department of Energy Office of Environmental Management (DOE/EM) plans to conduct the Plutonium Disposition Project at the Savannah River Site (SRS) to disposition excess weapons-usable plutonium. A plutonium glass waste form is the preferred option for immobilization of the plutonium for subsequent disposition in a geologic repository. A reference glass composition (Lanthanide Borosilicate (LaBS) Frit B) was developed during the Plutonium Immobilization Program (PIP) to immobilize plutonium in the late 1990's. A limited amount of performance testing was performed on this baseline composition before efforts to further pursue Pu disposition via a glass waste form ceased. Recent FY05 studies have further investigated the LaBS Frit B formulation as well as development of a newer LaBS formulation denoted as LaBS Frit X. The objectives of this present task were to fabricate plutonium loaded LaBS Frit X glass and perform corrosion testing to provide near-term data that will increase confidence that LaBS glass product is suitable for disposal in the Yucca Mountain Repository. Specifically, testing was conducted in an effort to provide data to Yucca Mountain Project (YMP) personnel for use in performance assessment calculations. Plutonium containing LaBS glass with the Frit X composition with a 9.5 wt% PuO{sub 2} loading was prepared for testing. Glass was prepared to support Product Consistency Testing (PCT) at Savannah River National Laboratory (SRNL). The glass was thoroughly characterized using x-ray diffraction (XRD) and scanning electron microscopy coupled with energy dispersive spectroscopy (SEM/EDS) prior to performance testing. A series of PCTs were conducted at SRNL using quenched Pu Frit X glass with varying exposed surface areas. Effects of isothermal and can-in-canister heat treatments on the Pu Frit X glass were also investigated. Another series of PCTs were performed on these different heat-treated Pu Frit X glasses. Leachates from all these PCTs were analyzed to determine the dissolved concentrations of key elements. Acid stripping of leach vessels was performed to determine the concentration of the glass constituents that may have sorbed on the vessels during leach testing. Additionally, the leachate solutions were ultrafiltered to quantify colloid formation. Characterization of the quenched Pu Frit X glass prior to testing revealed that some crystalline plutonium oxide was present in the glass. The crystalline particles had a disklike morphology and likely formed via coarsening of particles in areas compositionally enriched in plutonium. Similar results had also been observed in previous Pu Frit B studies. Isothermal 1250 C heat-treated Pu Frit X glasses showed two different crystalline phases (PuO{sub 2} and Nd{sub 2}Hf{sub 2}O{sub 7}), as well as a peak shift in the XRD spectra that is likely due to a solid solution phase PuO{sub 2}-HfO{sub 2} formation. Micrographs of this glass showed a clustering of some of the crystalline phases. Pu Frit X glass subjected to the can-in-canister heating profile also displayed the two PuO{sub 2} and Nd{sub 2}Hf{sub 2}O{sub 7} phases from XRD analysis. Additional micrographs indicate crystalline phases in this glass were of varying forms (a spherical PuO{sub 2} phase that appeared to range in size from submicron to {approx}5 micron, a dendritic-type phase that was comprised of mixed lanthanides and plutonium, and a minor phase that contained Pu and Hf), and clustering of the phases was also observed.

  6. Transmutation of high-level radioactive waste and production of {sup 233}U using an accelerator-driven reactor

    SciTech Connect (OSTI)

    Takahashi, Hiroshi; Takashita, Hirofumi; Chen, Xinyi

    1994-08-01T23:59:59.000Z

    Reactor safety, the disposal of high-level nuclear waste, and nonproliferation of nuclear material for military purposes are the problems of greatest concern for nuclear energy. Technologies for accelerators developed in the field of high-energy physics can contribute to solving these problems. For reactor safety, especially for that of a Na-cooled fast reactor, the use of an accelerator, even a small one, can enhance the safety using a slightly subcritical reactor. There is growing concern about how we can deal with weapons-grade Pu, and about the large amount of Pu accumulating from the operation of commercial reactors. It has been suggested that this Pu could be incinerated, using the reactor and a proton accelerator. However, because Pu is a very valuable material with future potential for generating nuclear energy, we should consider transforming it into a proliferation-resistant material that cannot be used for making bombs, rather than simply eliminating the Pu. An accelerator-driven fast reactor (700 MWt), run in a subcritical condition, and fueled with MOX can generate {sup 233}U more safely and efficiently than can a critical reactor. We evaluate the production of {sup 233}U, {sup 239}Pu, and the transmutation of the long-lived fission products of {sup 99}Tc and {sup 129}I, which are loaded with YH{sub 1.7} between the fast core and blanket, by reducing the conversion factor of Pu to {sup 233}U. And we assessed the rates of radiation damage, hydrogen production, and helium production in a target window and in the surrounding vessel.

  7. Evaluation of technologies for volume reduction of plutonium-contaminated soils from the Nevada Test Site

    SciTech Connect (OSTI)

    Papelis, C.; Jacobson, R.L.; Miller, F.L.; Shaulis, L.K.

    1996-06-01T23:59:59.000Z

    Nuclear testing at and around the Nevada Test Site (NTS) resulted in plutonium (Pu) contamination of the soil over an area of several thousands of acres. The objective of this project was to evaluate the potential of five different processes to reduce the volume of Pu-contaminated soil from three different areas, namely Areas 11, 13, and 52. Volume reduction was to be accomplished by concentrating the Pu into a small but highly contaminated soil fraction, thereby greatly reducing the volume of soil requiring disposal. The processes tested were proposed by Paramag Corp. (PARAMAG), Advanced Processing Technologies Inc. (APT), Lockheed Environmental Systems and Technologies (LESAT), Nuclear Remediation Technologies (NRT), and Scientific Ecology Group (SEG). Because of time and budgetary restraints, the NRT and SEG processes were tested with soil from Area 11 only. These processes typically included a preliminary soil conditioning step (e.g., attrition scrubbing, wet sieving), followed by a more advanced process designed to separate Pu from the soil, based on physiochemical properties of Pu compounds (e.g., magnetic susceptibility, specific gravity). Analysis of the soil indicates that a substantial fraction of the total Pu contamination is typically confined in a relatively narrow and small particle size range. Processes which were able to separate this highly contaminated soil fraction (using physical methods, e.g., attrition scrubbing, wet sieving), from the rest of the soil achieved volume (mass) reductions on the order of 70%. The advanced, more complex processes tested did not enhance volume reduction. The primary reason why processes that rely on the dependence of settling velocity on density differences failed was the very fine grain size of the Pu-rich particles.

  8. DISTRIBUTION OF ACTINIDES BETWEEN THE AQUEOUS AND ORGANIC PHASES IN THE TALSPEAK PROCESS

    SciTech Connect (OSTI)

    Rudisill, T.; Kyser, E.

    2010-09-02T23:59:59.000Z

    One objective of the US Department of Energy's Office of Nuclear Energy (DOE-NE) is the development of sustainable nuclear fuel cycles which improve uranium resource utilization, maximize energy generation, minimize waste generation, improve safety, and complement institutional measures limiting proliferation risks. Activities in progress which support this objective include the development of advanced separation technologies to recover the actinides from used nuclear fuels. With the increased interest in the development of technology to allow closure of the nuclear fuel cycle, the TALSPEAK process is being considered for the separation of Am and Cm from the lanthanide fission products in a next generation reprocessing plant. However, at this time, the level of understanding associated with the chemistry and the control of the process variables is not acceptable for deployment of the process on an industrial scale. To address this issue, DOE-NE is supporting basic scientific studies focused on the TALSPEAK process through its Fuel Cycle Research and Development (R&D) program. One aspect of these studies is an experimental program at the Savannah River National Laboratory (SRNL) in which temperature-dependent distribution coefficients for the extraction of actinide elements in the TALSPEAK process were measured. The data were subsequently used to calculate conditional enthalpies and entropies of extraction by van't Hoff analysis to better understand the thermodynamic driving forces for the TALSPEAK process. In the SRNL studies, the distribution of Pu(III) in the TALSPEAK process was of particular interest. A small amount of Pu(III) would be present in the feed due to process losses and valence adjustment in prior recovery operations. Actinide elements such as Np and Pu have multiple stable oxidation states in aqueous solutions; therefore the oxidation state for these elements must be controlled in the TALSPEAK process, as the extraction chemistry is dependent upon the actinide's valence. Since our plans included the measurement of Pu(III) distribution coefficients using a Np(V) solution containing small amounts of {sup 238}Pu, it was necessary to demonstrate that the desired oxidation states of Np and Pu are produced and could be stabilized in a buffered lactate solution containing diethylenetriaminepentaacetic (DTPA). The stability of Np(V) and Pu(III) in lactic acid/DTPA solutions was evaluated by ultraviolet-visible (UV-vis) spectroscopy. To perform the evaluation, Np and Pu were added to solutions containing either hydroxylamine nitrate (HAN) or ferrous sulfamate (FS) as the reductant and nominally 1.5 M lactic acid/0.05 M DTPA. The pH of the solution was subsequently adjusted to nominally 2.8 as would be performed in the TALSPEAK process. In the valence adjustment study, we found that it was necessary to reduce Pu to Pu(III) prior to combining with the lactic acid and DTPA. The Pu reduction was performed using either HAN or FS. When FS was used, Np was reduced to Np(IV). The spectroscopic studies showed that Np(V) and Pu(III) are not stable in lactic acid/DTPA solutions. The stability of Np(IV)- and Pu(IV)-DTPA complexes are much greater than the stability of the Np(V)- and Pu(III)-DTPA complexes, and as a result, Np is slowly reduced to Np(IV) and Pu is slowly oxidized to Pu(IV) due to the reduced activity of the more stable complexes. When Np(V) was added to a solution containing a 1.5 M lactic acid/ammonium lactate buffer and 0.05 M DTPA, approximately 50% of the Np was reduced to Np(IV) in the first day. The fraction of Np(V) in the solution continued to diminish with time and was essentially reduced to Np(IV) after one week. When Pu(III) was added to a lactic acid/DTPA solution of the same composition, the spectrum recorded following at least two days after preparation of the solution continued to show some sign of Pu(III). The Pu(III) was completely oxidized to Pu(IV) after 3-4 days. The UV-vis spectroscopy demonstrated that Np(V) and Pu(III) were the predominate valences in the lactic acid/DTPA solution for th

  9. DISSOLUTION OF PLUTONIUM METAL IN 8-10 M NITRIC ACID

    SciTech Connect (OSTI)

    Rudisill, T.; Pierce, R.

    2012-02-21T23:59:59.000Z

    The H-Canyon facility will be used to dissolve Pu metal for subsequent purification and conversion to plutonium dioxide (PuO{sub 2}) using Phase II of HB-Line. To support the new mission, the development of a Pu metal dissolution flowsheet which utilizes concentrated (8-10 M) nitric acid (HNO{sub 3}) solutions containing potassium fluoride (KF) is required. Dissolution of Pu metal in concentrated HNO{sub 3} is desired to eliminate the need to adjust the solution acidity prior to purification by anion exchange. The preferred flowsheet would use 8-10 M HNO{sub 3}, 0.015-0.07 M KF, and 0.5-1.0 g/L Gd to dissolve the Pu up to 6.75 g/L. An alternate flowsheet would use 8-10 M HNO{sub 3}, 0.1-0.2 M KF, and 1-2 g/L B to dissolve the Pu. The targeted average Pu metal dissolution rate is 20 mg/min-cm{sup 2}, which is sufficient to dissolve a 'standard' 2250-g Pu metal button in 24 h. Plutonium metal dissolution rate measurements showed that if Gd is used as the nuclear poison, the optimum dissolution conditions occur in 10 M HNO{sub 3}, 0.04-0.05 M KF, and 0.5-1.0 g/L Gd at 112 to 116 C (boiling). These conditions will result in an estimated Pu metal dissolution rate of {approx}11-15 mg/min-cm{sup 2} and will result in dissolution times of 36-48 h for standard buttons. The recommended minimum and maximum KF concentrations are 0.03 M and 0.07 M, respectively. The maximum KF concentration is dictated by a potential room-temperature Pu-Gd-F precipitation issue at low Pu concentrations. The purpose of the experimental work described in this report was two-fold. Initially a series of screening experiments was performed to measure the dissolution rate of Pu metal as functions of the HNO{sub 3}, KF, and Gd or B concentrations. The objective of the screening tests was to propose optimized conditions for subsequent flowsheet demonstration tests. Based on the rate measurements, this study found that optimal dissolution conditions in solutions containing 0.5-1.0 g/L Gd occurred in 8-10 M HNO{sub 3} with 0.04-0.05 M KF at 112 to 116 C (boiling). The testing also showed that solutions containing 8-10 M HNO{sub 3}, 0.1-0.2 M KF, and 1-2 g/L B achieved acceptable dissolution rates in the same temperature range. To confirm that conditions identified by the dissolution rate measurements for solutions containing Gd or B can be used to dissolve Pu metal up to 6.75 g/L in the presence of Fe, demonstration experiments were performed using concentrations in the optimal ranges. In two of the demonstration experiments using Gd and in one experiment using B, the offgas generation during the dissolution was measured and samples were analyzed for H{sub 2}. The experimental methods used to perform the dissolution rate measurements and flowsheet demonstrations and a discussion of the results are presented.

  10. TECHNICAL BASIS FOR DOE STANDARD 3013 EQUIVALENCY SUPPORTING REDUCED TEMPERATURE STABILIZATION OF OXALATE-DERIVED PLUTONIUM OXIDE PRODUCED BY THE HB-LINE FACILITY AT SAVANNAH RIVER SITE

    SciTech Connect (OSTI)

    Duffey, J.; Livingston, R.; Berg, J.; Veirs, D.

    2012-07-02T23:59:59.000Z

    The HB-Line (HBL) facility at the Savannah River Site (SRS) is designed to produce high-purity plutonium dioxide (PuO{sub 2}) which is suitable for future use in production of Mixed Oxide (MOX) fuel. The MOX Fuel Fabrication Facility (MFFF) requires PuO{sub 2} feed to be packaged per the U.S. Department of Energy (DOE) Standard 3013 (DOE-STD-3013) to comply with the facility's safety basis. The stabilization conditions imposed by DOE-STD-3013 for PuO{sub 2} (i.e., 950 C for 2 hours) preclude use of the HBL PuO{sub 2} in direct fuel fabrication and reduce the value of the HBL product as MFFF feedstock. Consequently, HBL initiated a technical evaluation to define acceptable operating conditions for production of high-purity PuO{sub 2} that fulfills the DOE-STD-3013 criteria for safe storage. The purpose of this document is to demonstrate that within the defined operating conditions, the HBL process will be equivalent for meeting the requirements of the DOE-STD-3013 stabilization process for plutonium-bearing materials from the DOE complex. The proposed 3013 equivalency reduces the prescribed stabilization temperature for high-purity PuO{sub 2} from oxalate precipitation processes from 950 C to 640 C and places a limit of 60% on the relative humidity (RH) at the lowest material temperature. The equivalency is limited to material produced using the HBL established flow sheet, for example, nitric acid anion exchange and Pu(IV) direct strike oxalate precipitation with stabilization at a minimum temperature of 640 C for four hours (h). The product purity must meet the MFFF acceptance criteria of 23,600 {micro}g/g Pu (i.e., 2.1 wt %) total impurities and chloride content less than 250 {micro}g/g of Pu. All other stabilization and packaging criteria identified by DOE-STD-3013-2012 or earlier revisions of the standard apply. Based on the evaluation of test data discussed in this document, the expert judgment of the authors supports packaging the HBL product under a 3013 equivalency. Under the defined process conditions and associated material specifications, the high-purity PuO{sub 2} produced in HBL presents no unique safety concerns for packaging or storage in the 3013 required configuration. The PuO{sub 2} produced using the HBL flow sheet conditions will have a higher specific surface area (SSA) than PuO{sub 2} stabilized at 950 C and, consequently, under identical conditions will adsorb more water from the atmosphere. The greatest challenge to HBL operators will be controlling moisture content below 0.5 wt %. However, even at the 0.5 wt % moisture limit, the maximum acceptable pressure of a stoichiometric mixture of hydrogen and oxygen in the 3013 container is greater than the maximum possible pressure for the HBL PuO{sub 2} product.

  11. Durability of Actinide Ceramic Waste Forms Under Conditions of Granitoid Rocks

    SciTech Connect (OSTI)

    Burakov, B. E.; Anderson, E. B.

    2002-02-26T23:59:59.000Z

    Three samples of {sup 239}Pu-{sup 241}Am-doped ceramics obtained from previous research were used for alteration experiments simulating corrosion of waste forms in ion-saturated solutions. These were ceramics based on: pyrochlore, (Ca,Hf,Pu,U,Gd){sub 2}Ti{sub 2}O{sub 7}, containing 10 wt.% Pu and 0.1 wt.% Am; zircon, (Zr,Pu)SiO{sub 4}, containing 5-6 wt.% Pu and 0.05 wt.% Am; cubic zirconia, (Zr,Gd,Pu)O{sub 2}, containing 10 wt.% Pu and 0.1 wt.% Am. All these samples were milled in an agate mortar to obtain powder with particle sizes less than 30 micron. Sample of granite taken from the depth 500-503 m was studied and then used for preparing ion-saturated water solutions. A rock sample was ground, washed and classified. A fraction with particle size 0.10-0.25 mm was selected for alteration experiments. Powdered ceramic samples were separately placed into deionized water together with ground granite (approximately 1gram granite per 12-ml water) in special Teflon{trademark} vessels and set at 90 C in the oven for 3 months. After alteration experiments, the ceramic powders were studied by precise XRD analysis. Aqueous solutions and granite grains were analyzed for Am and Pu contents. The results show that alteration did not cause significant phase transformation in all ceramic samples. For all altered samples, the Am contents in aqueous solutions after experiments were similar (approximately n x 10{sup 2} Bq/ml) as well as Am amounts absorbed on granite grains (approximately n x 10{sup 5} Bq/g). Results on Pu contents were varied: for the solutions--from 60 Bq/ml for pyrochlore ceramic to 2.1 x 10{sup 3} Bq/ml for zircon ceramic; and for the absorption on granite--from 2.6 x 10{sup 4} Bq/g for zirconia ceramic to 1.4-6.8 x 10{sup 5} Bq/g for pyrochlore and zircon ceramics.

  12. ANALYSIS OF 2H-EVAPORATOR SCALE WALL [HTF-13-82] AND POT BOTTOM [HTF-13-77] SAMPLES

    SciTech Connect (OSTI)

    Oji, L.

    2013-06-21T23:59:59.000Z

    Savannah River Remediation (SRR) is planning to remove a buildup of sodium aluminosilicate scale from the 2H-evaporator pot by loading and soaking the pot with heated 1.5 M nitric acid solution. Sampling and analysis of the scale material has been performed so that uranium and plutonium isotopic analysis can be input into a Nuclear Criticality Safety Assessment (NCSA) for scale removal by chemical cleaning. Historically, since the operation of the Defense Waste Processing Facility (DWPF), silicon in the DWPF recycle stream combines with aluminum in the typical tank farm supernate to form sodium aluminosilicate scale mineral deposits in the 2Hevaporator pot and gravity drain line. The 2H-evaporator scale samples analyzed by Savannah River National Laboratory (SRNL) came from the bottom cone sections of the 2H-evaporator pot [Sample HTF-13-77] and the wall 2H-evaporator [sample HTF-13-82]. X-ray diffraction analysis (XRD) confirmed that both the 2H-evaporator pot scale and the wall samples consist of nitrated cancrinite (a crystalline sodium aluminosilicate solid) and clarkeite (a uranium oxy-hydroxide mineral). On “as received” basis, the bottom pot section scale sample contained an average of 2.59E+00 ± 1.40E-01 wt % total uranium with a U-235 enrichment of 6.12E-01 ± 1.48E-02 %, while the wall sample contained an average of 4.03E+00 ± 9.79E-01 wt % total uranium with a U-235 enrichment of 6.03E-01% ± 1.66E-02 wt %. The bottom pot section scale sample analyses results for Pu-238, Pu-239, and Pu-241 are 3.16E- 05 ± 5.40E-06 wt %, 3.28E-04 ± 1.45E-05 wt %, and <8.80E-07 wt %, respectively. The evaporator wall scale samples analysis values for Pu-238, Pu-239, and Pu-241 averages 3.74E-05 ± 6.01E-06 wt %, 4.38E-04 ± 5.08E-05 wt %, and <1.38E-06 wt %, respectively. The Pu-241 analyses results, as presented, are upper limit values. These results are provided so that SRR can calculate the equivalent uranium-235 concentrations for the NCSA. Results confirm that the uranium contained in the scale remains depleted with respect to natural uranium. SRNL did not calculate an equivalent U-235 enrichment, which takes into account other fissionable isotopes U-233, Pu-239 and Pu-241. The applicable method for calculation of equivalent U-235 will be determined in the NCSA.

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

    SciTech Connect (OSTI)

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

    2010-12-07T23:59:59.000Z

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

  14. Spent nuclear fuel recycling with plasma reduction and etching

    DOE Patents [OSTI]

    Kim, Yong Ho

    2012-06-05T23:59:59.000Z

    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.

  15. Differential die-away technique for determination of the fissile contents in spent fuel assembly

    SciTech Connect (OSTI)

    Lee, Tachoon [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, Nartyn T [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

    2010-01-01T23:59:59.000Z

    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.

  16. Whole-Organism Concentration Ratios for Plutonium in Wildlife from Past US Nuclear Research Data

    SciTech Connect (OSTI)

    johansen, M.; Kamboj; Kuhne, W.

    2012-07-26T23:59:59.000Z

    Whole-organism concentration ratios (CR{sub wo-media}) for plutonium (Pu) in wildlife were calculated using data from the broad range of organism types and environmental settings of the US nuclear research program. Original sources included site-specific reports and scientific journal articles typically from 1960s to 80s research. Most of the calculated CR{sub wo-media} values are new to existing data sets, and, for some wildlife categories, serve to fill gaps or add to sparse data including those for terrestrial reptile; freshwater bird, crustacean and zooplankton; and marine crustacean and zooplankton. Ratios of Pu concentration in the whole-organism to that in specific tissues and organs are provided here for a range of freshwater and marine fish. The CR{sub wo-media} values in fish living in liquid discharge ponds were two orders of magnitude higher than those for similar species living in lakes receiving Pu from atmospheric fallout, suggesting the physico-chemical form of the source Pu can dominate over other factors related to transfer, such as organism size and feeding behavior. Small rodent data indicated one to two order of magnitude increases when carcass, pelt, and gastrointestinal tract were included together in the whole-organism calculation compared to that for carcass alone. Only 4% of Pu resided in the carcass of small rodents compared to 75% in the gastrointestinal tract and 21% in the pelt.

  17. Russian-U.S. joint program on the safe management of nuclear materials

    SciTech Connect (OSTI)

    Witmer, F.E.; Krumpe, P.F. [Dept. of Energy, Washington, DC (US); Carlson, D.D. [Sandia National Labs., Albuquerque, NM (US)] [and others

    1997-12-01T23:59:59.000Z

    The Russian-US joint program on the safety of nuclear materials was initiated in response to the 1993 Tomsk-7 accident. The bases for this program are the common technical issues confronting the US and Russia in the safe management of excess weapons grade nuclear materials. The US and Russian weapons dismantlement process is producing hundreds of tons of excess Pu and HEU fissile materials. The US is on a two path approach for disposition of excess Pu: (1) use Pu in existing reactors and/or (2) immobilize Pu in glass or ceramics followed by geologic disposal. Russian plans are to fuel reactors with excess Pu. US and Russia are both converting and blending HEU into LEU for use in existing reactors. Fissile nuclear materials storage, handling, processing, and transportation will be occurring in both countries for tens of years. A table provides a history of the major events comprising the Russian-US joint program on the safety of nuclear materials. A paper delineating program efforts was delivered at the SPECTRUM '96 conference. This paper provides an update on program activities since then.

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

    SciTech Connect (OSTI)

    Jones, Susan A.; Delegard, Calvin H.

    2011-03-10T23:59:59.000Z

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

  19. Americium and plutonium in water, biota, and sediment from the central Oregon coast

    SciTech Connect (OSTI)

    Nielsen, R. D.

    1982-06-01T23:59:59.000Z

    Plutonium-239, 240 and americium-241 were measured in the mussel Mytilus californianus from the region of Coos Bay, OR. The flesh of this species has a plutonium concentration of about 90 fCi/kg, and an Am-241/Pu-239, 240 ratio that is high relative to mixed fallout, ranging between two and three. Transuranic concentrations in sediment, unfiltered water, and filterable particulates were also measured; none of these materials has an Am/Pu ratio as greatly elevated as the mussels, and there is no apparent difference in the Am/Pu ratio of terrestrial runoff and coastal water. Sediment core profiles do not allow accumulation rates or depositional histories to be identified, but it does not appear that material characterized by a high Am/Pu ratio has ever been introduced to this estuary. Other bivalves (Tresus capax and Macoma nasuta) and a polychaete (Abarenicola sp.) do not have an elevated Am/Pu ratio, although the absolute activity of plutonium in the infaunal bivalves is roughly four times that in the mussels.

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

    SciTech Connect (OSTI)

    Morgan, Dane [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong Austin [Univ. of Wisconsin, Madison, WI (United States)

    2013-10-28T23:59:59.000Z

    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.

  1. Developments in Very Low Level Waste/Exempt Waste Assay at AWE - 12000

    SciTech Connect (OSTI)

    Miller, T.J. [AWE, Aldermaston, Reading, Berkshire, RG7 4PR (United Kingdom)

    2012-07-01T23:59:59.000Z

    Portable High Resolution Gamma Spectrometry (HRGS) has been developed, for Very Low Level Waste (VLLW) and Exempt Waste (EW) assay at AWE, in order to meet the latest reduced clearance levels of < 1 Bq/g (or Bq/cm{sup 2}) for uranium (U) contaminated wastes and < 0.15 Bq/g (or Bq/cm{sup 2}) for plutonium (Pu) wastes. Studies have focused on a 10 kg bag of low density soft waste monitored either as a rotating cylinder, contained within a shortened plastic drum liner, or as a contained disk monitored on each broad side. Liquid and surface contaminated metal wastes have also been studied. It was established that monitoring the disk gave the best detection levels, but uncertainties rose more sharply, compared to the cylinder, as detector offset was reduced. Exempt detection levels were readily achieved for all U compositions encountered at AWE and for most Pu compositions (via Am-241 measurement). However, performance will need to be enhanced for those Pu compositions with relatively high Pu/Am-241 activity ratios. Recommendations are made for further developments to enhance the performance of this technique so that exempt clearance can be achieved for all Pu compositions encountered. (author)

  2. Helium Behavior in Oxide Nuclear Fuels: First Principles Modeling

    SciTech Connect (OSTI)

    D. Gryaznov; S. Rashkeev; E. A. Kotomin; E. Heifets; Y. Zhukovskii

    2010-10-01T23:59:59.000Z

    UO2 and (U, Pu)O2 solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein. We present density functional theory (DFT) calculations for UO2, PuO2 and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model (in the framework of the so-called DFT + U approach). We found that PuO2 remains semiconducting with He in the octahedral position while UO2 requires a specific lattice distortion. Both materials reveal a positive energy for He incorporation, which, therefore, is an exothermic process. The He incorporation energy increases with the Pu concentration in the MOX fuel.

  3. Value of burnup credit beyond actinides

    SciTech Connect (OSTI)

    Lancaster, D.; Fuentes, E.; Kang, Chi

    1997-12-01T23:59:59.000Z

    DOE has submitted a topical report to the NRC justifying burnup credit based only on actinide isotopes (U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241). When this topical report is approved, it will allow a great deal of the commercial spent nuclear fuel to be transported in significantly higher capacity casks. A cost savings estimate for shipping fuel in 32 assembly (burnup credit) casks as opposed to 24 assembly (non-burnup credit) casks was previously presented. Since that time, more detailed calculations have been performed using the methodology presented in the Actinide-Only Burnup Credit Topical Report. Loading curves for derated casks have been generated using actinide-only burnup credit and are presented in this paper. The estimates of cost savings due to burnup credit for shipping fuel utilizing 32, 30, 28, and 24 assembly casks where only the 24 assembly cask does not burnup credit have been created and are discussed. 4 refs., 2 figs.

  4. Plutonium Detection with Straw Neutron Detectors

    SciTech Connect (OSTI)

    Mukhopadhyay, Sanjoy; Maurer, Richard; Guss, Paul

    2014-03-27T23:59:59.000Z

    A kilogram of weapons grade plutonium gives off about 56,000 neutrons per second of which 55,000 neutrons come from spontaneous fission of 240Pu (~6% by weight of the total plutonium). Actually, all even numbered isotopes (238Pu, 240Pu, and 242Pu) produce copious spontaneous fission neutrons. These neutrons induce fission in the surrounding fissile 239Pu with an approximate multiplication of a factor of ~1.9. This multiplication depends on the shape of the fissile materials and the surrounding material. These neutrons (typically of energy 2 MeV and air scattering mean free path >100 meters) can be detected 100 meters away from the source by vehicle-portable neutron detectors. [1] In our current studies on neutron detection techniques, without using 3He gas proportional counters, we designed and developed a portable high-efficiency neutron multiplicity counter using 10B-coated thin tubes called straws. The detector was designed to perform like commercially available fission meters (manufactured by Ortec Corp.) except instead of using 3He gas as a neutron conversion material, we used a thin coating of 10B.

  5. Reactor physics studies for assessment of tramp uranium methods

    SciTech Connect (OSTI)

    Grimm, P.; Vasiliev, A.; Wieselquist, W.; Ferroukhi, H. [Paul Scherrer Institut, CH 5232 Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, CH 5325 Leibstadt (Switzerland)

    2012-07-01T23:59:59.000Z

    This paper presents calculation studies towards validation of a methodology for estimations of the tramp uranium mass from water chemistry measurements. Particular emphasis is given to verify, from a reactor physics point of view, the justification basis for the so-called 'Pu-based model' versus the 'U-based model' as a key assumption for the methodology. The computational studies are carried out for a typical BWR fuel assembly with CASMO-5M and MCNPX. By approximating the evolution of fissile nuclides and the fraction of {sup 235}U fissions to total fissions in different zones of a fuel rod, including tramp uranium on the clad surface, it is found that Pu gives the dominant contribution to fissions for tramp uranium after an irradiation on the outer clad surface of at least one cycle in a BWR. Thus, the use of the so-called Pu model for the determination of the tramp uranium mass (this means in particular using the yields for {sup 239}Pu fission) appears justified in the cases considered. On that basis, replacing the older U model by a Pu model is recommended. (authors)

  6. Effects of non-latching blast valves on the source term and consequences of the design-basis accidents in the Device Assembly Facility (DAF)

    SciTech Connect (OSTI)

    Nguyen, D.H.

    1993-08-01T23:59:59.000Z

    The analysis of the Design-Basis Accidents (DBA) involving high explosives (HE) and Plutonium (Pu) in the assembly cell of the Device Assembly Facility (DAF), which was completed earlier, assumed latching blast valves in the ventilation system of the assembly cell. Latching valves effectively sealed a release path through the ventilation duct system. However, the blast valves in the assembly cell, as constructed are actually non-latching valves, and would reopen when the gas pressure drops to 0.5 psi above one atmosphere. Because the reopening of the blast valves provides an additional release path to the environment, and affects the material transport from the assembly cell to other DAF buildings, the DOE/NV DAF management has decided to support an additional analysis of the DAF`s DBA to account for the effects of non-latching valves. Three cases were considered in the DAF`s DBA, depending on the amount of HE and Pu involved, as follows: Case 1 -- 423 {number_sign}HE, 16 kg Pu; Case 2 -- 150 {number_sign}HE 10 kg Pu; Case 3 -- 55 {number_sign}HE 5 kg Pu. The results of the analysis with non-latching valves are summarized.

  7. Dissolution of plutonium oxide in nitric acid at high hydrofluoric acid concentrations

    SciTech Connect (OSTI)

    Kazanjian, A.R.; Stevens, J.R.

    1984-06-15T23:59:59.000Z

    The dissolution of plutonium dioxide in nitirc acid (HNO/sub 3/) at high hydrofluoric acid (HF) concentrations has been investigated. Dissolution rate curves were obtained using 12M HNO/sub 3/ and HF at concentrations varying from 0.05 to 1.0 molar. The dissolution rate increased with HF concentration up to 0.2M and then decreased at higher concentrations. There was very little plutonium dissolved at 0.7 and 1.0M HF because of the formation of insoluble PuF/sub 4/. Various oxidizing agents were added to 12M HNO/sub 3/-1M HF dissolvent to oxidize Pu(IV) to Pu(VI) and prevent the formation of PuF/sub 4/. Ceric (Ce(IV)) and silver (Ag(II)) ions were the most effective in dissolving PuO/sub 2/. Although these two oxidants greatly increased the dissolution rate, the rates were not as rapid as those obtained with 12M HNO/sub 3/-0.2M HF.

  8. Summary of a joint US-Japan study of potential approaches to reduce the attractiveness of various nuclear materials for use in a nuclear explosive device by a terrorist group

    SciTech Connect (OSTI)

    Bathke, C.G. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM (United States); Inoue, N.; Kuno, Y.; Mihara, T.; Sagara, H. [Japan Atomic Energy Agency, 4-49 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1184 (Japan); Ebbinghaus, B.B. [Lawrence Livermore National Laboratory, P.O. Box L-168, Livermore, CA 94551 (United States); Murphy, J.; Dalton, D. [National Nuclear Security Administration, Department of Energy, 1000 Independence Ave, SW, Washington, DC 20585 (United States); Nagayama, Y. [Ministry of Education, Culture, Sports, Science and Technology, 3-2-2 Kasumigaseki, Chiyoda-ku, Tokyo 100-8959 (Japan)

    2013-07-01T23:59:59.000Z

    This paper summarizes the results of a joint US-Japan study to establish a mutual understanding, through scientific-based study, of potential approaches to reduce the attractiveness of various nuclear materials for use in a terrorist nuclear explosive device (NED). 4 approaches that can reduce materials attractiveness with a very high degree of effectiveness are: -) diluting HEU with natural or depleted U to an enrichment of less than 10% U-235; -) storing Pu in nuclear fuel that is not man portable and with a dose rate greater or equal to 10 Gy/h at 1 m; -) storing Pu or HEU in heavy items, i.e. not transportable, provided the removal of the Pu or HEU from the item requires a purification/processing capability; and -) converting Pu and HEU to very dilute forms (such as wastes) that, without any security barriers, would require very long acquisition times to acquire a Category I quantity of Pu or of HEU. 2 approaches that can reduce materials attractiveness with a high degree of effectiveness are: -) converting HEU-fueled research reactors into LEU-fueled research reactors or dilute HEU with natural or depleted U to an enrichment of less than 20% U-235; -) converting U/Al reactor fuel into U/Si reactor fuel. Other approaches have been assessed as moderately or totally inefficient to reduce the attractiveness of nuclear materials.

  9. Plutonium Mobility Studies: 216-Z-9 Trench Sample Analysis Results

    SciTech Connect (OSTI)

    Cantrell, Kirk J.; Geiszler, Keith N.; Arey, Bruce W.

    2008-09-05T23:59:59.000Z

    A variety of analyses were conducted on selected sediment samples collected from two wells (299 W15-46 and 299-W15-48) drilled near the 216-Z-9 Trench to elucidate the form and potential for Pu and Am to be mobilized under present conditions and those that could be expected in future remediation scenarios. Analyses included moisture content, determination of the less than sand size fraction (silt plus clay), carbon analysis, SEM/EDS analysis, microwave-assisted acid digestions for total element analysis, and extraction tests using Hanford groundwater as the leachate. Results of the extraction tests were used as input to conduct equilibrium geochemical modeling of the solutions with Geochemist’s Workbench®. Geochemical modeling results for Pu were evaluated in terms of recent conclusions regarding the solubility and redox reactions of Pu by Neck et al. (2007a, 2007b). It was found that the highest concentrations of Pu and Am were associated with sediments of low silt/clay content and occur above silt/clay rich layers within the sediment profile. It was also found that the Pu and Am were relatively enriched in the silt/clay portion of these samples. The fact that the highest concentrations of Pu and Am occurred in sediments with low silt/clay contents suggests that waste solutions had perched on top of the low permeability silt/clay rich layers and interactions with the high silt/clay layers was minimal. SEM/EDS analysis indicated that the Pu and Am in these sediments does not occur as discrete micron size particles, and therefore must occur as mononuclear or polynuclear/ nanoclusters size particles adsorbed throughout the sediment samples. Leaching of these samples with Hanford groundwater indicates that release of Pu and Am from the sediments is correlated most significantly with the acidity of the water and not the initial concentrations of Pu and Am in the sediments. Only extracts that were acidic after contact with the sediments (pH 4.3 to 5.4) contained detectable concentrations of extractable Pu and Am. Water extracts from samples containing high concentrations of TBP suggest that if the TBP degradation products DBP and MBP are available in these sediments, they do not significantly increase the extractability of Pu or Am. Geochemical modeling results suggest that the concentrations of Am in water in contact with these sediments is not controlled by the solubility of Am(OH)3(c), but rather by desorption of Am that has been previously adsorbed to the sediments during the period of active wastewater disposal. Sediment extracts that had measureable concentrations of Am only occurred in samples that were fairly acidic (pH 4.3 to 4.6), indicating that Am will remain effectively sequestered to sediments when pH conditions approach those of normal Hanford groundwater (mildly alkaline, ~ pH 8). The geochemical modeling results indicate that Pu in acidic extracts is significantly undersaturated with respect to PuO2(am). However, recent reviews of Pu solubility and redox reactions suggest that the data used for these calculations is incomplete (Neck et al. 2007a, 2007b). The results of Neck et al. (2007a, 2007b) suggest that Pu concentrations in solutions in contact with the 216-Z-9 Trench sediment samples might be controlled by a mixed valent solid phase [(PuV)2x(PuIV)1-2xO2+x(am)] with various dissolved Pu(V) complexes and Pu(IV)O2(am) colloids or nanoclusters being the dominant species in solution for typical Hanford groundwater conditions. Adsorption is likely to have a major impact on the mobility of these species (Neck et al. 2007a, 2007b; Clark et al. 2006; Kaplan et al. 2006; Powell et al. 2005). Further research is planned to verify these hypotheses.

  10. Plutonium-238 alpha-decay damage study of the ceramic waste form.

    SciTech Connect (OSTI)

    Frank, S M [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Barber, T L [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Cummings, D G [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; DiSanto, T [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Esh, D W [U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; Giglio, J J [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Goff, K M [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Johnson, S G [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Kennedy, J R [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Jue, J-F [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Noy, M [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; O'Holleran, T P [U.S. Department of Energy, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415; Sinkler, W [UOP LLC, 25 E Algonquin Road, Des Plaines, IL 60017

    2006-03-27T23:59:59.000Z

    An accelerated alpha-decay damage study of a glass-bonded sodalite ceramic waste form has recently been completed. The purpose of this study was to investigate the physical and chemical durability of the waste form after significant exposure to alpha decay. This accelerated alpha-decay study was performed by doping the ceramic waste form with {sup 238}Pu which has a much greater specific activity than {sup 239}Pu that is normally present in the waste form. The alpha-decay dose at the end of the four year study was approximately 1 x 10{sup 18} alpha-decays/gram of material. An equivalent time period for a similar dose of {sup 239}Pu would require approximately 1100 years. After four years of exposure to {sup 238}Pu alpha decay, the investigation observed little change to the physical or chemical durability of the ceramic waste form (CWF). Specifically, the {sup 238}Pu-loaded CWF maintained it's physical integrity, namely that the density remained constant and no cracking or phase de-bonding was observed. The materials chemical durability and phase stability also did not change significantly over the duration of the study. The only significant measured change was an increase of the unit-cell lattice parameters of the plutonium oxide and sodalite phases of the material and an increase in the release of salt components and plutonium of the waste form during leaching tests, but, as mentioned, these did not lead to any overall loss of waste form durability. The principal findings from this study are: (1) {sup 238}Pu-loaded CWF is similar in microstructure and phase composition to referenced waste form. (2) Pu was observed primarily as oxide comprised of aggregates of nano crystals with aggregates ranging in size from submicron to twenty microns in diameter. (3) Pu phases were primarily found in the intergranular glassy regions. (4) PuO phase shows expected unit cell volume expansion due to alpha decay damage of approximately 0.7%, and the sodalite phase unit cell volume has expanded slightly by 0.3% again, presumably due to alpha-decay damage. (5) No bulk sample swelling was observed. (6) No amorphization of sodalite or actinide bearing phases was observed after four years of alpha-decay damage. (7) No microcracks or phase de-bonding were observed in waste form samples aged for four years. (8) In some areas of the {sup 238}Pu doped ceramic waste form material bubbles and voids were found. Bubbles and voids with similar size and density were also found in ceramic waste form samples without actinide. These bubbles and voids are interpreted as pre-existing defects. However, some contribution to these bubbles and voids from helium gas can not be ruled out. (9) Chemical durability of {sup 238}Pu CWF has not changed significantly after four years of alpha-decay exposure except for an increase in the release of salt components and Pu. Still, the plutonium release from CWF is very low at less than 0.005 g/m{sup 2}.

  11. Implementation of Iterative Map turbo Decoder on TMS320C40 DSP

    E-Print Network [OSTI]

    Patil, Sunil S

    1997-01-01T23:59:59.000Z

    ) = P(u ) p(y lu ) If L'(us) is a priori information then (P(u, = ql) ) (P(u~ = ? I)) Using some mathematical manipulation we can write (1+ ?p( ? L'(us)jJ (2 9) P(ut) = At, exp[ut, L'(ut, )/2] (2. 10) This follows from ( /~ /P+ 3 1/P+/P = P... BITS USING UNIFORM INTERLEAVER LDA eRECD2ADR, ARO LDA @INPUT ADR, AR1 LDI 16, R9 IN LOQP7: LDI O, IRO LDI 15, RC RPTB IN LQQP6 LDI +AR1++, RO STI RO, ++ARO(IRO) IN LQQP6: ADDI 16, IRO ADDI 1, ARO SUBI 1, R9 BNZ IN LQQP7 CMEM ADR, ARi 1...

  12. Unidentified growth factors in the diet of chicks reared under commercial broiler production conditions

    E-Print Network [OSTI]

    Camp, Arthur A.

    1958-01-01T23:59:59.000Z

    LW OyiOuLmOKC SKM; CdOuOePuO 9 mLhdC rO iPWCHASCOM CP BPKCSLK HKLMOKCLeLOM huPoCd eSBCPuW ePu CdO BdLBsP adLW dtiPCdOWLW LW eHuCdOu WHiiPuCOM rt CdO eSBC CdSC CdO SMMLCLPK Pe CdO CoP WHiiAOmOKCW CP CdO rSWSA MLOC 32 BPKBHuuOKCAt iuPMHBOM S WLh... and Van Poucke (.1947) 9 and Gerry et al. (1948) that dried skim milk contained a factor or factors required by the chick whose composition was at that time unknown, Newell et al- (1947) postulated that dried penicillin mycelium was a source...

  13. Ultra-high-resolution alpha spectrometry for nuclear forensics and safeguards applications

    SciTech Connect (OSTI)

    Bacrania, Minesh K [Los Alamos National Laboratory; Croce, Mark [Los Alamos National Laboratory; Bond, Evelyn [Los Alamos National Laboratory; Dry, Donald [Los Alamos National Laboratory; Moody, W. Allen [Los Alamos National Laboratory; Lamont, Stephen [Los Alamos National Laboratory; Rabin, Michael [Los Alamos National Laboratory; Rim, Jung [Los Alamos National Laboratory; Smith, Audrey [Los Alamos National Laboratory; Beall, James [NIST-BOULDER; Bennett, Douglas [NIST-BOULDER; Kotsubo, Vincent [NIST-BOULDER; Horansky, Robert [NIST-BOULDER; Hilton, Gene [NIST-BOULDER; Schmidt, Daniel [NIST-BOULDER; Ullom, Joel [NIST-BOULDER; Cantor, Robin [STAR CRYOELECTRONICS

    2010-01-01T23:59:59.000Z

    We will present our work on the development of ultra-high-resolution detectors for alpha particle spectrometry. These detectors, based on superconducting transition-edge sensors, offer energy resolution that is five to ten times better than conventional silicon detectors. Using these microcalorimeter detectors, the isotopic composition of mixed-actinide samples can be determined rapidly without the need for actinide separation chemistry to isolate each element, or mass spectrometry to separate isotopic signatures that can not be resolved using traditional alpha spectrometry (e.g. Pu-239/Pu-240, or Pu-238/Am-241). This paper will cover the detector and measurement system, actinide source preparation, and the quantitative isotopic analysis of a number of forensics- and safeguards-relevant radioactive sources.

  14. FURTHER DEVELOPMENT OF MODIFIED MONOSODIUM TITANATE, AN IMPROVED SORBENT FOR PRETREATMENT OF HIGH LEVEL NUCLEAR WASTE AT THE SAVANNAH RIVER SITE

    SciTech Connect (OSTI)

    Taylor-Pashow, K.; Hobbs, D.; Fondeur, F.; Fink, S.

    2011-01-12T23:59:59.000Z

    High-level nuclear waste produced from fuel reprocessing operations at the Savannah River Site (SRS) requires pretreatment to remove Cs-137, Sr-90, and alpha-emitting radionuclides (i.e., actinides) prior to disposal onsite as low level waste. Separation processes planned at SRS include caustic side solvent extraction, for Cs-137 removal, and sorption of Sr-90 and alpha-emitting radionuclides onto monosodium titanate (MST). The predominant alpha-emitting radionuclides in the highly alkaline waste solutions include plutonium isotopes Pu-238, Pu-239, and Pu-240. This paper describes recent results from the development of an improved titanate material that exhibits increased removal kinetics and effective capacity for Sr-90 and alpha-emitting radionuclides compared to the baseline MST material.

  15. Recommended Method To Account For Daughter Ingrowth For The Portsmouth On-Site Waste Disposal Facility Performance Assessment Modeling

    SciTech Connect (OSTI)

    Phifer, Mark A.; Smith, Frank G. III

    2013-06-21T23:59:59.000Z

    A 3-D STOMP model has been developed for the Portsmouth On-Site Waste Disposal Facility (OSWDF) at Site D as outlined in Appendix K of FBP 2013. This model projects the flow and transport of the following radionuclides to various points of assessments: Tc-99, U-234, U-235, U-236, U-238, Am-241, Np-237, Pu-238, Pu-239, Pu-240, Th-228, and Th-230. The model includes the radioactive decay of these parents, but does not include the associated daughter ingrowth because the STOMP model does not have the capability to model daughter ingrowth. The Savannah River National Laboratory (SRNL) provides herein a recommended method to account for daughter ingrowth in association with the Portsmouth OSWDF Performance Assessment (PA) modeling.

  16. Improved Radiation Dosimetry/Risk Estimates to Facilitate Environmental Management of Plutonium-Contaminated Sites

    SciTech Connect (OSTI)

    Scott, Bobby R.; Tokarskaya, Zoya B.; Zhuntova, Galina V.; Osovets, Sergey V.; Syrchikov, Victor A., Belyaeva, Zinaida D.

    2007-12-14T23:59:59.000Z

    This report summarizes 4 years of research achievements in this Office of Science (BER), U.S. Department of Energy (DOE) project. The research described was conducted by scientists and supporting staff at Lovelace Respiratory Research Institute (LRRI)/Lovelace Biomedical and Environmental Research Institute (LBERI) and the Southern Urals Biophysics Institute (SUBI). All project objectives and goals were achieved. A major focus was on obtaining improved cancer risk estimates for exposure via inhalation to plutonium (Pu) isotopes in the workplace (DOE radiation workers) and environment (public exposures to Pu-contaminated soil). A major finding was that low doses and dose rates of gamma rays can significantly suppress cancer induction by alpha radiation from inhaled Pu isotopes. The suppression relates to stimulation of the body's natural defenses, including immunity against cancer cells and selective apoptosis which removes precancerous and other aberrant cells.

  17. Influence of soil biopopulation on migration of waste radionuclides

    SciTech Connect (OSTI)

    Fowler, E.B.; Polzer, W.L.; Essington, E.H.

    1983-01-01T23:59:59.000Z

    This paper reports the interpretation of some results obtained when a Maxey Flats burial pit radioactive waste solution was reacted with a Tilsit soil. The influence of a biopopulation on the degree of sorption and on the stability of that system was investigated. The data have been interpreted as follows: the removal of /sup 137/Cs from solution by the soil is essentially complete within a one-hour period and is not influenced by an active biopopulation. The soil studied contains complexers which solublize /sup 238/Pu. The soluble complex does not sorb to soil and thus is potentially mobile. In the presence of an active biopopulation 86% of the complexes is degraded; the released /sup 238/Pu was rendered immobile. The remaining 14% of the soluble /sup 238/Pu was not released to the soil during 53 days incubation. That fraction is heat stable and non- or slowly-biodegradable and thus retains its potential to migrate.

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

    SciTech Connect (OSTI)

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10T23:59:59.000Z

    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.

  19. Subthreshold Photofission of Even-Even Nuclei

    SciTech Connect (OSTI)

    Kadmensky, S.G.; Rodionova, L.V. [Voronezh State University, Universitetskaya pl. 1, Voronezh, 394693 (Russian Federation)

    2005-09-01T23:59:59.000Z

    Within quantum-mechanical fission theory, the angular distributions of fragments originating from the subthreshold photofission of the even-even nuclei {sup 232}Th, {sup 234}U, {sup 236}U, {sup 238}U, {sup 238}Pu, {sup 240}Pu, and {sup 242}Pu are analyzed for photon energies below 7 MeV. Special features of various fission channels are assessed under the assumption that the fission barrier has a two-humped shape. It is shown that the maximum value of the relative orbital angular momentum L{sub m} of fission fragments can be found upon taking into account deviations from the predictions of A. Bohr's formula for the angular distributions of fission fragments. The result is L{sub m} {approx_equal} 30. The existence of an 'isomeric shelf' for the angular distributions of fragments from {sup 236}U and {sup 238}U photofission in the low-energy region is confirmed.

  20. A Dependence of the Electronuclear System Parameters on the Component Concentration in Fuel MOX

    E-Print Network [OSTI]

    Barashenkov, V S

    2001-01-01T23:59:59.000Z

    A dependence of the parameters of a electronuclear system with U-Pu fuel MOX on the relative share of plutonium and its isotope ^{240}Pu in fuel is investigated by means of mathematical modelling. As an example, we consider an experimental system with a heat power 10-20 kW designed in Dubna on the basis of the 660 MeV proton phasotron. The 2 % admixture of ^{240}Pu decreases the value of the neutron multiplication coefficient from 0.95 down to 0.90, neutron yield and heat power are diminished almost twice. Such a decrease can be compensated by the increase of Plutonium share in MOX from 25 up to 27 %.

  1. Ground state properties and high pressure behavior of plutonium dioxide: Systematic density functional calculations

    E-Print Network [OSTI]

    Zhang, Ping; Zhao, Xian-Geng

    2010-01-01T23:59:59.000Z

    Plutonium dioxide is of high technological importance in nuclear fuel cycle and is particularly crucial in long-term storage of Pu-based radioactive waste. Using first-principles density-functional theory, in this paper we systematically study the structural, electronic, mechanical, thermodynamic properties, and pressure induced structural transition of PuO$_{2}$. To properly describe the strong correlation in the Pu $5f$ electrons, the local density approximation$+U$ and the generalized gradient approximation$+U$ theoretical formalisms have been employed. We optimize the $U$ parameter in calculating the total energy, lattice parameters, and bulk modulus at the nonmagnetic, ferromagnetic, and antiferromagnetic configurations for both ground state fluorite structure and high pressure cotunnite structure. The best agreement with experiments is obtained by tuning the effective Hubbard parameter $U$ at around 4 eV within the LDA$+U$ approach. After carefully testing the validity of the ground state, we further in...

  2. A Note on the Reaction of Hydrogen and Plutonium

    SciTech Connect (OSTI)

    Noone, Bailey C [Los Alamos National Laboratory

    2012-08-15T23:59:59.000Z

    Plutonium hydride has many practical and experimental purposes. The reaction of plutonium and hydrogen has interesting characteristics, which will be explored in the following analysis. Plutonium is a radioactive actinide metal that emits alpha particles. When plutonium metal is exposed to air, the plutonium oxides and hydrides, and the volume increases. PuH{sub 2} and Pu{sub 2}O{sub 3} are the products. Hydrogen is a catalyst for plutonium's corrosion in air. The reaction can take place at room temperature because it is fairly insensitive to temperature. Plutonium hydride, or PuH{sub 2}, is black and metallic. After PuH{sub 2} is formed, it quickly flakes off and burns. The reaction of hydrogen and plutonium is described as pyrophoric because the product will spontaneously ignite when oxygen is present. This tendency must be considered in the storage of metal plutonium. The reaction is characterized as reversible and nonstoichiometric. The reaction goes as such: Pu + H{sub 2} {yields} PuH{sub 2}. When PuH{sub 2} is formed, the hydrogen/plutonium ratio is between 2 and 2.75 (approximately). As more hydrogen is added to the system, the ratio increases. When the ratio exceeds 2.75, PuH{sub 3} begins to form along with PuH{sub 2}. Once the ratio surpasses 2.9, only PuH{sub 3} remains. The volume of the plutonium sample increases because of the added hydrogen and the change in crystal structure which the sample undergoes. As more hydrogen is added to a system of metal plutonium, the crystal structure evolves. Plutonium has a crystal structure classified as monoclinic. A monoclinic crystal structure appears to be a rectangular prism. When plutonium reacts with hydrogen, the product PuH{sub 2}, becomes a fluorite structure. It can also be described as a face centered cubic structure. PuH{sub 3} forms a hexagonal crystal structure. As plutonium evolves from metal plutonium to plutonium hydride to plutonium trihydride, the crystal structure evolves from monoclinic to fluorite to hexagonal. This change in crystal structure as a result of adding hydrogen is a shared characteristic with other actinide elements. Americium is isostructural with plutonium because they both form cubic dihyrides and hexagonal trihydrides. Reacting hydrogen with plutonium has the practical application of separating plutonium from other materials that don't react as well with hydrogen. When plutonium is placed in a chamber where there is very little oxygen, it can react with hydrogen without igniting. The hydrogen plutonium reaction can then be reversed, thus regaining the separated plutonium. Another application of this reaction is that it can be used to predict how plutonium reacts with other substances. Deuterium and tritium are two isotopes of hydrogen that are of interest. They are known to react likewise to hydrogen because they have similar properties. The reaction of plutonium and isotopes of hydrogen can prove to be very informative.

  3. Investigation of the Performance of D2O-Cooled High-Conversion Reactors for Fuel Cycle Calculations

    SciTech Connect (OSTI)

    Hikaru Hiruta; Gilles Youinou

    2013-09-01T23:59:59.000Z

    This report presents FY13 activities for the analysis of D2O cooled tight-pitch High-Conversion PWRs (HCPWRs) with U-Pu and Th-U fueled cores aiming at break-even or near breeder conditions while retaining the negative void reactivity. The analyses are carried out from several aspects which could not be covered in FY12 activities. SCALE 6.1 code system is utilized, and a series of simple 3D fuel pin-cell models are developed in order to perform Monte Carlo based criticality and burnup calculations. The performance of U-Pu fueled cores with axial and internal blankets is analyzed in terms of their impact on the relative fissile Pu mass balance, initial Pu enrichment, and void coefficient. In FY12, Pu conversion performances of D2O-cooled HCPWRs fueled with MOX were evaluated with small sized axial/internal DU blankets (approximately 4cm of axial length) in order to ensure the negative void reactivity, which evidently limits the conversion performance of HCPWRs. In this fiscal year report, the axial sizes of DU blankets are extended up to 30 cm in order to evaluate the amount of DU necessary to reach break-even and/or breeding conditions. Several attempts are made in order to attain the milestone of the HCPWR designs (i.e., break-even condition and negative void reactivity) by modeling of HCPWRs under different conditions such as boiling of D2O coolant, MOX with different 235U enrichment, and different target burnups. A similar set of analyses are performed for Th-U fueled cores. Several promising characteristics of 233U over other fissile like 239Pu and 235U, most notably its higher fission neutrons per absorption in thermal and epithermal ranges combined with lower ___ in the fast range than 239Pu allows Th-U cores to be taller than MOX ones. Such an advantage results in 4% higher relative fissile mass balance than that of U-Pu fueled cores while retaining the negative void reactivity until the target burnup of 51 GWd/t. Several other distinctions between U-Pu and Th-U fueled cores are identified by evaluating the sensitivity coefficients of keff, mass balance, and void coefficient. The effect of advanced iron alloy cladding (i.e., FeCrAl) on the performance of Pu conversion in MOX fueled cores is studied instead of using standard stainless-steel cladding. Variations in clad thickness and coolant-to-fuel volume ratio are also exercised. The use of FeCrAl instead of SS as a cladding alloy reduces the required Pu enrichment and improves the Pu conversion rate primarily due to the absence of nickel in the cladding alloy that results in the reduction of the neutron absorption. Also the difference in void coefficients between SS and FeCrAl alloys is nearly 500 pcm over the entire burnup range. The report also shows sensitivity and uncertainty analyses in order to characterize D2O cooled HCPWRs from different aspects. The uncertainties of integral parameters (keff and void coefficient) for selected reactor cores are evaluated at different burnup points in order to find similarities and trends respect to D2O-HCPWR.

  4. PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

    SciTech Connect (OSTI)

    S. T. Khericha

    2007-04-01T23:59:59.000Z

    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.

  5. Development of a Phosphate Ceramic as a Host for Halide-Contaminated Plutonium Pyrochemical Reprocessing Wastes

    SciTech Connect (OSTI)

    Metcalfe, Brian; Fong, Shirley; Gerrard, Lee; Donald, Ian [MSRD, AWE plc, AWE Aldermaston, Reading, RG7 4PR (United Kingdom); Strachan, Denis; Scheele, Randall [PNNL, Richland, WA, 99352 (United States)

    2007-07-01T23:59:59.000Z

    The presence of halide anions in four types of wastes arising from the pyrochemical reprocessing of plutonium required an immobilization process to be developed in which not only the actinide cations but also the halide anions were immobilized in a durable, leach resistant form. AWE has developed such a process using Ca{sub 3}(PO{sub 4}){sub 2} as the host material. Successful trials of the process using actinide-doped Type I waste (essentially a chloride-based waste) were carried out at PNNL where the immobilization of the waste in a form resistant to aqueous leaching was confirmed. Normalized mass losses determined using a modified MCC-1 test at 40 deg. C/28 days were 12 x 10{sup -6} g.m{sup -2} and 2.7 x 10{sup -3} g.m{sup -2} for Pu and Cl, respectively. Accelerated radiation-induced damage effects are being determined with specimens containing {sup 238}Pu. No changes in the crystalline lattice have been detected with XRD after the {sup 239}Pu equivalent of 400 years ageing. Confirmation of the process for Type II waste (an oxyhydroxide-based waste) is currently underway at PNNL. Differences in the ionic state of plutonium in the four types of waste have required different surrogates to be used. Samarium chloride was used successfully as a surrogate for both Pu(III) and Am(III) chlorides. Early investigations into the use of HfO{sub 2} as the surrogate for Pu(IV) oxide in Type II waste showed some apparent differences in the phase assemblages of the surrogate and actinide-based products. However XRD examination of the products at higher resolution has demonstrated there is no significant difference and that for this work HfO{sub 2} is a suitable surrogate for PuO{sub 2}. (authors)

  6. THE SUITABILITY OF SODIUM PEROXIDE FUSION FOR PRODUCTION-SCALE PLUTONIUM PROCESSING OPERATIONS

    SciTech Connect (OSTI)

    Pierce, R.; Edwards, T.

    2010-10-26T23:59:59.000Z

    Sodium peroxide (Na{sub 2}O{sub 2}) fusion is a method that offers significant benefits to the processing of high-fired plutonium oxide (PuO{sub 2}) materials. Those benefits include reduction in dissolution cycle time, decrease in residual solids, and reduction of the potential for generation of a flammable gas mixture during dissolution. Implementation of Na{sub 2}O{sub 2} fusion may also increase the PuO{sub 2} throughput in the HB-Line dissolving lines. To fuse a material, Na{sub 2}O{sub 2} is mixed with the feed material in a crucible and heated to 600-700 C. For low-fired and high-fired PuO{sub 2}, Na{sub 2}O{sub 2} reacts with PuO{sub 2} to form a compound that readily dissolves in ambient-temperature nitric acid without the use of potassium fluoride. The Savannah River National Laboratory (SRNL) demonstrated the feasibility of Na{sub 2}O{sub 2} fusion and subsequent dissolution for the processing of high-fired PuO{sub 2} materials in HB-Line. Testing evaluated critical dissolution characteristics and defined preliminary process parameters. Based on experimental measurements, a dissolution cycle can be complete in less than one hour, compared to the current processing time of 6-10 hours for solution heating and dissolution. Final Pu concentrations of 30-35 g/L were produced without the formation of precipitates in the final solution.

  7. ARRAYS OF BOTTLES OF PLUTONIUM NITRATE SOLUTION

    SciTech Connect (OSTI)

    Margaret A. Marshall

    2012-09-01T23:59:59.000Z

    In October and November of 1981 thirteen approaches-to-critical were performed on a remote split table machine (RSTM) in the Critical Mass Laboratory of Pacific Northwest Laboratory (PNL) in Richland, Washington using planar arrays of polyethylene bottles filled with plutonium (Pu) nitrate solution. Arrays of up to sixteen bottles were used to measure the critical number of bottles and critical array spacing with a tight fitting Plexiglas® reflector on all sides of the arrays except the top. Some experiments used Plexiglas shells fitted around each bottles to determine the effect of moderation on criticality. Each bottle contained approximately 2.4 L of Pu(NO3)4 solution with a Pu content of 105 g Pu/L and a free acid molarity H+ of 5.1. The plutonium was of low 240Pu (2.9 wt.%) content. These experiments were sponsored by Rockwell Hanford Operations because of the lack of experimental data on the criticality of arrays of bottles of Pu solution such as might be found in storage and handling at the Purex Facility at Hanford. The results of these experiments were used “to provide benchmark data to validate calculational codes used in criticality safety assessments of [the] plant configurations” (Ref. 1). Data for this evaluation were collected from the published report (Ref. 1), the approach to critical logbook, the experimenter’s logbook, and communication with the primary experimenter, B. Michael Durst. Of the 13 experiments preformed 10 were evaluated. One of the experiments was not evaluated because it had been thrown out by the experimenter, one was not evaluated because it was a repeat of another experiment and the third was not evaluated because it reported the critical number of bottles as being greater than 25. Seven of the thirteen evaluated experiments were determined to be acceptable benchmark experiments. A similar experiment using uranyl nitrate was benchmarked as U233-SOL-THERM-014.

  8. The use of non-destructive passive neutron measurement methods in dismantling and radioactive waste characterization

    SciTech Connect (OSTI)

    Jallu, F.; Allinei, P. G. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); Bernard, P.; Loridon, J. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); Soyer, P.; Pouyat, D. [CEA, DEN, Marcoule, DPAD, F-30207 Bagnols-sur-Ceze Cedex (France); Torreblanca, L. [CEA, DEN, Cadarache, LMDE, F-13108 Saint-Paul-lez-Durance (France); Reneleau, A. [AREVA NC, Pierrelatte, DDAC/ESD, BP16, F-26701 Pierrelatte Cedex (France)

    2011-07-01T23:59:59.000Z

    The cleaning up and dismantling of nuclear facilities lead to a great volume of technological radioactive wastes which need to be characterized in order to be sent to the adequate final disposal or interim storage. The control and characterization can be performed with non-destructive nuclear measurements such as gamma-ray spectrometry. Passive neutron counting is an alternative when the alpha-gamma emitters cannot be detected due to the presence of a high gamma emission resulting from fission or activation products, or when the waste matrix is too absorbing for the gamma rays of interest (too dense and/or made of high atomic number elements). It can also be a complement to gamma-ray spectrometry when two measurement results must be confronted to improve the confidence in the activity assessment. Passive neutron assays involve the detection of spontaneous fission neutrons emitted by even nuclides ({sup 238}Pu, {sup 240}Pu, {sup 242}Pu, {sup 242}Cm, {sup 244}Cm...) and neutrons resulting from ({alpha}, n) reactions with light nuclides (O, F, Be...). The latter is conditioned by the presence of high {alpha}-activity radionuclides ({sup 234}U, {sup 238}Pu, {sup 240}Pu, {sup 241}Am...) and low-Z elements, which depends on the chemical form (metallic, oxide or fluorine) of the plutonium or uranium contaminant. This paper presents the recent application of passive neutron methods to the cleaning up of a nuclear facility located at CEA Cadarache (France), which concerns the Pu mass assessment of 2714 historic, 100 litre radioactive waste drums produced between 1980 and 1997. Another application is the dismantling and decommissioning of an uranium enrichment facility for military purposes, which involves the {sup 235}U and total uranium quantifications in about a thousand, large compressors employed in the gaseous diffusion enrichment process. (authors)

  9. Characterization of optically stimulated luminescent dosimeters, OSLDs, for clinical dosimetric measurements

    SciTech Connect (OSTI)

    Jursinic, Paul A. [West Michigan Cancer Center, 200 North Park Street, Kalamazoo, Michigan 49007 (United States)

    2007-12-15T23:59:59.000Z

    Optically stimulated luminescent dosimeters, OSLDs, are plastic disks infused with aluminum oxide doped with carbon (Al{sub 2}O{sub 3}:C). These disks are encased in a light-tight plastic holder. Crystals of Al{sub 2}O{sub 3}:C when exposed to ionizing radiation store energy that is released as luminescence (420 nm) when the OSLD is illuminated with stimulation light (540 nm). The intensity of the luminescence depends on the dose absorbed by the OSLD and the intensity of the stimulation light. OSLDs used in this work were InLight/OSL Dot dosimeters, which were read with a MicroStar reader (Landauer, Inc., Glenwood, IL). The following are dosimetric properties of the OSLD that were determined: After a single irradiation, repeated readings cause the signal to decrease by 0.05% per reading; the signal could be discharged by greater than 98% by illuminating them for more than 45 s with a 150 W tungsten-halogen light; after irradiation there was a transient signal that decayed with a 0.8 min halftime; after the transient signal decay the signal was stable for days; repeated irradiations and readings of an individual OSLD gave a signal with a coefficient of variation of 0.6%; the dose sensitivity of OSLDs from a batch of detectors has a coefficient of variation of 0.9%, response was linear with absorbed dose over a test range of 1-300 cGy; above 300 cGy a small supra-linear behavior occurs; there was no dose-per-pulse dependence over a 388-fold range; there was no dependence on radiation energy or mode for 6 and 15 MV x rays and 6-20 MeV electrons; for Ir-192 gamma rays OSLD had 6% higher sensitivity; the dose sensitivity was unchanged up to an accumulated dose of 20 Gy and thereafter decreased by 4% per 10 Gy of additional accumulated dose; dose sensitivity was not dependent on the angle of incidence of radiation; the OSLD in its light-tight case has an intrinsic buildup of 0.04 g/cm{sup 2}; dose sensitivity of the OSLD was not dependent on temperature at the time of irradiation in the range of 10-40 deg. C. The clinical use of OSLDs for in vivo dosimetric measurements is shown to be feasible.

  10. Changes in optically stimulated luminescent dosimeter (OSLD) dosimetric characteristics with accumulated dose

    SciTech Connect (OSTI)

    Jursinic, Paul A. [West Michigan Cancer Center, 200 North Park St., Kalamazoo, Michigan 49007 (United States)

    2010-01-15T23:59:59.000Z

    Purpose: A new type of in vivo dosimeter, an optically stimulated luminescent dosimeter (OSLD), has now become commercially available for clinical use. The OSLD is a plastic disk infused with aluminum oxide doped with carbon (Al{sub 2}O{sub 3}:C). Crystals of Al{sub 2}O{sub 3}:C, when exposed to ionizing radiation, store energy that is released as luminescence (420 nm) when the OSLD is illuminated with stimulation light (540 nm). The intensity of the luminescence depends on the dose absorbed by the OSLD and the intensity of the stimulation light. The effects of accumulated dose on OSLD response were investigated. Methods: The OSLDs used in this work were nanodot dosimeters, which were read with a MicroStar reader (Landauer, Inc., Glenwood, IL). Dose to the OSLDs was delivered by 6 MV x rays and gamma rays from Co-60 and Ir-192. The signal on the OSLDs after irradiation is removed by optical annealing with a 150 W tungsten-halogen lamp or a 14 W compact fluorescent lamp was investigated. Results: It was found that OSLD response to dose was supralinear and this response was altered with the amount of accumulated dose to the OSLD. The OSLD response can be modeled by a quadratic and an exponential equation. For accumulated doses up to 60 Gy, the OSLD sensitivity (counts/dose) decreases and the extent of supralinear increases. Above 60 Gy of accumulated dose the sensitivity increases and the extent of supralinearity decreases or reaches a plateau, depending on how the OSLDs were optically annealed. With preirradiation of OSLDs with greater than 1 kGy, it is found that the sensitivity reaches a plateau 2.5 folds greater than that of an OSLD with no accumulated dose and the supralinearity disappears. A regeneration of the luminescence signal in the dark after full optical annealing occurs with a half time of about two days. The extent of this regeneration signal depends on the amount of accumulated dose. Conclusions: For in vivo dosimetric measurements, a precision of {+-}0.5% can be achieved if the sensitivity and extent of supralinearity is established for each OSLD and use. Methods are presented for accomplishing this task.

  11. Development and implementation of a remote audit tool for high dose rate (HDR) Ir-192 brachytherapy using optically stimulated luminescence dosimetry

    SciTech Connect (OSTI)

    Casey, Kevin E.; Kry, Stephen F.; Howell, Rebecca M.; Followill, David [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, Houston, Texas 77030 and The University of Texas Graduate School of Biomedical Sciences at Houston, Houston, Texas 77030 (United States)] [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, Houston, Texas 77030 and The University of Texas Graduate School of Biomedical Sciences at Houston, Houston, Texas 77030 (United States); Alvarez, Paola; Lawyer, Ann [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, Houston, Texas 77030 (United States)] [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, Houston, Texas 77030 (United States)

    2013-11-15T23:59:59.000Z

    Purpose: The aim of this work was to create a mailable phantom with measurement accuracy suitable for Radiological Physics Center (RPC) audits of high dose-rate (HDR) brachytherapy sources at institutions participating in National Cancer Institute-funded cooperative clinical trials. Optically stimulated luminescence dosimeters (OSLDs) were chosen as the dosimeter to be used with the phantom.Methods: The authors designed and built an 8 × 8 × 10 cm{sup 3} prototype phantom that had two slots capable of holding Al{sub 2}O{sub 3}:C OSLDs (nanoDots; Landauer, Glenwood, IL) and a single channel capable of accepting all {sup 192}Ir HDR brachytherapy sources in current clinical use in the United States. The authors irradiated the phantom with Nucletron and Varian {sup 192}Ir HDR sources in order to determine correction factors for linearity with dose and the combined effects of irradiation energy and phantom characteristics. The phantom was then sent to eight institutions which volunteered to perform trial remote audits.Results: The linearity correction factor was k{sub L}= (?9.43 × 10{sup ?5}× dose) + 1.009, where dose is in cGy, which differed from that determined by the RPC for the same batch of dosimeters using {sup 60}Co irradiation. Separate block correction factors were determined for current versions of both Nucletron and Varian {sup 192}Ir HDR sources and these vendor-specific correction factors differed by almost 2.6%. For the Nucletron source, the correction factor was 1.026 [95% confidence interval (CI) = 1.023–1.028], and for the Varian source, it was 1.000 (95% CI = 0.995–1.005). Variations in lateral source positioning up to 0.8 mm and distal/proximal source positioning up to 10 mm had minimal effect on dose measurement accuracy. The overall dose measurement uncertainty of the system was estimated to be 2.4% and 2.5% for the Nucletron and Varian sources, respectively (95% CI). This uncertainty was sufficient to establish a ±5% acceptance criterion for source strength audits under a formal RPC audit program. Trial audits of four Nucletron sources and four Varian sources revealed an average RPC-to-institution dose ratio of 1.000 (standard deviation = 0.011).Conclusions: The authors have created an OSLD-based {sup 192}Ir HDR brachytherapy source remote audit tool which offers sufficient dose measurement accuracy to allow the RPC to establish a remote audit program with a ±5% acceptance criterion. The feasibility of the system has been demonstrated with eight trial audits to date.

  12. Radionuclides and heavy metals in rainbow trout from Tsichomo, Nana Ka, Wen Povi, and Pin De Lakes in Santa Clara Canyon

    SciTech Connect (OSTI)

    Fresquez, P.R.; Armstrong, D.R.; Naranjo, L. Jr.

    1998-04-01T23:59:59.000Z

    Radionuclide ({sup 3}H, {sup 90}Sr, {sup 137}Cs, {sup 238}Pu, {sup 239}Pu, and total uranium) and heavy metal (Ag, As, Ba, Be, Cd, Cr, Hg, Ni, Pb, Sb, Se, and TI) concentrations were determined in rainbow trout collected from Tsichomo, Nana Ka, Wen Povi, and Pin De lakes in Santa Clara Canyon in 1997. Most radionuclide and heavy metal concentrations in fish collected from these four lakes were within or just above upper limit background concentrations (Abiquiu reservoir), and as a group were statistically (p < 0.05) similar in most parameters to background.

  13. Milliwatt Generator Project: April 1988--September 1996. Progress report

    SciTech Connect (OSTI)

    Latimer, T.W.

    1997-07-01T23:59:59.000Z

    This report covers progress on the Milliwatt Generator (MWG) Project from April 1988 to September 1996. Manufacturing of heat sources for the project ended by September 1990. Beginning in October 1990, the major activities of the project have been surveillance and testing of MWGs, disposal of excess MWGs, and reclamation of the PuO{sub 2} from excess MWG heat sources. Reported activities include fuel processing and characterization, production of heat sources, compatibility studies, impact testing, examination and electrical testing of surveillance units, and recovery of PuO{sub 2} from heat sources.

  14. Separation and Purification and Beta Liquid Scintillation Analysis of Sm-151 in Savannah River Site and Hanford Site DOE High Level Waste

    SciTech Connect (OSTI)

    Dewberry, R.A.

    2001-02-13T23:59:59.000Z

    This paper describes development work to obtain a product phase of Sm-151 pure of any other radioactive species so that it can be determined in US Department of Energy high level liquid waste and low level solid waste by liquid scintillation {beta}-spectroscopy. The technique provides separation from {mu}Ci/ml levels of Cs-137, Pu alpha and Pu-241 {beta}-decay activity, and Sr-90/Y-90 activity. The separation technique is also demonstrated to be useful for the determination of Pm-147.

  15. Integral Validation of Minor Actinide Nuclear Data by using Samples Irradiated at Dounreay Prototype Fast Reactor

    SciTech Connect (OSTI)

    Tsujimoto, Kazufumi; Oigawa, Hiroyuki; Shinohara, Nobuo [Japan Atomic Energy Research Institute, Shirakata Shirane 2-4, Tokai, Ibaraki 319-1195 (Japan)

    2005-05-24T23:59:59.000Z

    The reliability of nuclear data for minor actinides was evaluated by using the results of the post-irradiation experiment for actinide samples irradiated at the Dounreay Prototype Fast Reactor. The burnup calculations with JENDL-3.3, ENDF/B-VI.8, and JEFF-3.0 were performed. From the comparison between the experimental data and the calculational results, in general, the reliability of nuclear data for the minor actinides are at an adequate level for the conceptual design study of transmutation systems. It is, however, found that improvement of the accuracy is necessary for some nuclides, such as 238Pu, 242Pu, and 241Am.

  16. IN-SITU ASSAY OF TRANSURANIC RADIONUCLIDES IN THE VADOSE ZONE USING HIGH-RESOLUTION SPECTRAL GAMMA LOGGING - A HANFORD CASE STUDY

    SciTech Connect (OSTI)

    ROHAY VJ; HENWOOD P; MCCAIN R

    2009-11-30T23:59:59.000Z

    High-resolution spectral gamma logging in steel-cased boreholes is used to detect and quantify transuranic radionuclides in the subsurface. Pu-239, Pu-241, Am-241, and Np-237 are identified based on characteristic decay gammas. Typical minimum detectable levels are on the order of 20 to 40 nCi/g. In intervals of high transuranic concentrations, gamma rays from other sources may complicate analysis and interpretation. Gamma rays detected in the borehole may originate from three sources: decay of the parent transuranic radionuclide or a daughter; alpha interactions; and interactions with neutrons resulting from either spontaneous fission or alpha particle interactions.

  17. Analysis of nuclear materials by energy dispersive x-ray fluorescence and spectral effects of alpha decay

    SciTech Connect (OSTI)

    Worley, Christopher G [Los Alamos National Laboratory

    2009-01-01T23:59:59.000Z

    Energy dispersive X-ray fluorescence (EDXRF) spectra collected from alpha emitters are complicated by artifacts inherent to the alpha decay process, particularly when using portable instruments. For example, {sup 239}Pu EDXRF spectra exhibit a prominent uranium L X-ray emission peak series due to sample alpha decay rather than source-induced X-ray fluorescence. A portable EDXRF instrument was used to collect spectra from plutonium, americium, and a Pu-contaminated steel sample. The plutonium sample was also analyzed by wavelength dispersive XRF to demonstrate spectral differences observed when using these very different instruments.

  18. Assay of low-level plutonium effluents

    SciTech Connect (OSTI)

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

    1981-01-01T23:59:59.000Z

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

  19. Spectral analysis of slender tensioned cylinder interaction

    E-Print Network [OSTI]

    Diao, Weiguo

    1998-01-01T23:59:59.000Z

    the coefficient k, and c, are constants: Case I The Duffing nonlinear system is defined by p(u, u, t) =k, u'(t) Case 2 The Van der Pol nonlinear system is defined by (60) p(u, u, t)= ? ? u (t)= c, u (t)u(t) c, c 3 Ct (6l) Case 3 The combined Duffing-Van der... models, the Duffing model, the Van Der Pol model and a combination model will be examined using the test data obtained in the study by Rijken (1997). By developing a better understanding of the interactive behavior between two tensioned tandem...

  20. Renewability and sustainability aspects of nuclear energy

    SciTech Connect (OSTI)

    ?ahin, Sümer, E-mail: ssahin@atilim.edit.tr [Department of Mechanical Engineering, Faculty of Engineering, ATILIM University, 06836 ?ncek, Gölba??, Ankara (Turkey)

    2014-09-30T23:59:59.000Z

    Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, {sup 233}U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO{sub 2}/RG?PuO{sub 2}) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG?PuO{sub 2} + 96 % ThO{sub 2}; 6 % RG?PuO{sub 2} + 94 % ThO{sub 2}; 10 % RG?PuO{sub 2} + 90 % ThO{sub 2}; 20 % RG?PuO{sub 2} + 80 % ThO{sub 2}; 30 % RG?PuO{sub 2} + 70 % ThO{sub 2}, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ? 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ? 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG?PuO{sub 2} fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MW{sub th} has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ?160 kg {sup 233}U per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of ?1.3.