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Sample records for metal u-pu-zr fuelled

  1. Comparison of U-Pu-Mo, U-Pu-Nb, U-Pu-Ti and U-Pu-Zr alloys; Comparaison des alliages U-Pu-Mo, U-Pu-Nb, U-Pu-Ti, U-Pu-Zr

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    The data concerning the U-Pu, U-Pu-Mo and U-Pu-Nb are recalled. The results obtained with U-Pu-Ti and U-Pu-Zr alloys containing 15-20 per cent Pu and 10 wt. per cent ternary element are reported. The transformation temperatures, the expansion coefficients, the nature of phases, the thermal cycling behaviour have been determined. A list of the principal properties of these different alloys is presented and the possibilities of their use as fast reactor's fuel element are considered. The U-Pu-Ti alloys seem to be quite promising: easiness of fabrication, large thermal stability, excellent behaviour in air, small quantity of zeta phase, temperature of solidus superior to 1100 deg. C. (authors) [French] On rappelle brievement les connaissances acquises sur les alliages U-Pu, U-Pu-Mo et U-Pu-Nb. On presente les resultats obtenus avec les alliages U-Pu-Ti et U-Pu-Zr pour des teneurs de 15 a 20 pour cent de plutonium et 10 pour cent en poids d'element ternaire. On a determine les temperatures de transformation, les coefficients de dilatation, la nature des phases, la conductibilite thermique a 20 deg. C, la tenue au cyclage thermique et diverses autres proprietes. Un tableau resume les principales proprietes des divers alliages. On considere les possibilites d'emploi de ces alliages comme combustibles de reacteur rapide. Les alliages U-Pu-Ti paraissent particulierement interessants: facilite d'elaboration, stabilite thermique etendue, tenue dans l'air excellente, faible quantite de la phase U-Pu zeta, temperature de fusion commencante superieure a 1100 deg. C. (auteurs)

  2. Comparison of U-Pu-Mo, U-Pu-Nb, U-Pu-Ti and U-Pu-Zr alloys

    International Nuclear Information System (INIS)

    Boucher, R.; Barthelemy, P.

    1964-01-01

    The data concerning the U-Pu, U-Pu-Mo and U-Pu-Nb are recalled. The results obtained with U-Pu-Ti and U-Pu-Zr alloys containing 15-20 per cent Pu and 10 wt. per cent ternary element are reported. The transformation temperatures, the expansion coefficients, the nature of phases, the thermal cycling behaviour have been determined. A list of the principal properties of these different alloys is presented and the possibilities of their use as fast reactor's fuel element are considered. The U-Pu-Ti alloys seem to be quite promising: easiness of fabrication, large thermal stability, excellent behaviour in air, small quantity of zeta phase, temperature of solidus superior to 1100 deg. C. (authors) [fr

  3. Modeling of constituent redistribution in U-Pu-Zr metallic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States)]. E-mail: yskim@anl.gov; Hayes, S.L. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Yacout, A.M. [Argonne National Laboratory, Nuclear Engineering, RERTR, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2006-12-01

    A computer model was developed to analyze constituent redistribution in U-Pu-Zr metallic nuclear fuels. Diffusion and thermochemical properties were parametrically determined to fit the postirradiation data from a fuel test performed in the Experimental Breeder Reactor II (EBR-II). The computer model was used to estimate redistribution profiles of fuels proposed for the conceptual designs of small modular fast reactors. The model results showed that the level of redistribution of the fuel constituents of the designs was similar to the measured data from EBR-II.

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

    International Nuclear Information System (INIS)

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

    1994-10-01

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

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

    International Nuclear Information System (INIS)

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

    1992-10-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. High temperature investigation of the solid/liquid transition in the PuO2-UO2-ZrO2 system

    Science.gov (United States)

    Quaini, A.; Guéneau, C.; Gossé, S.; Sundman, B.; Manara, D.; Smith, A. L.; Bottomley, D.; Lajarge, P.; Ernstberger, M.; Hodaj, F.

    2015-12-01

    The solid/liquid transitions in the quaternary U-Pu-Zr-O system are of great interest for the analysis of core meltdown accidents in Pressurised Water Reactors (PWR) fuelled with uranium-dioxide and MOX. During a severe accident the Zr-based cladding can become completely oxidised due to the interaction with the oxide fuel and the water coolant. In this framework, the present analysis is focused on the pseudo-ternary system UO2-PuO2-ZrO2. The melting/solidification behaviour of five pseudo-ternary and one pseudo-binary ((PuO2)0.50(ZrO2)0.50) compositions have been investigated experimentally by a laser heating method under pre-set atmospheres. The effects of an oxidising or reducing atmosphere on the observed melting/freezing temperatures, as well as the amount of UO2 in the sample, have been clearly identified for the different compositions. The oxygen-to-metal ratio is a key parameter affecting the melting/freezing temperature because of incongruent vaporisation effects. In parallel, a detailed thermodynamic model for the UO2-PuO2-ZrO2 system has been developed using the CALPHAD method, and thermodynamic calculations have been performed to interpret the present laser heating results, as well as the high temperature behaviour of the cubic (Pu,U,Zr)O2±x-c mixed oxide phase. A good agreement was obtained between the calculated and experimental data points. This work enables an improved understanding of the major factors relevant to severe accident in nuclear reactors.

  8. Establishment of technological basis for fabrication of U-Pu-Zr ternary alloy fuel pins for irradiation tests in Japan

    International Nuclear Information System (INIS)

    Kikuchi, Hironobu; Iwai, Takashi; Nakajima, Kunihisa; Arai, Yasuo; Nakamura, Kinya; Ogata, Takanari

    2011-01-01

    A high-purity Ar gas atmosphere glove box accommodating injection casting and sodium-bonding apparatuses was newly installed in the Plutonium Fuel Research Facility of Oarai Research and Development Center, Japan Atomic Energy Agency, in which several nitride and carbide fuel pins were fabricated for irradiation tests. The experiences led to the establishment of the technological basis of the fabrication of U-Pu-Zr alloy fuel pins for the first time in Japan. After the injection casting of the U-Pu-Zr alloy, the metallic fuel pins were fabricated by welding upper and lower end plugs with cladding tubes of ferritic-martensitic steel. Subsequent to the sodium bonding for filling the annular gap region between the U-Pu-Zr alloy and the cladding tube with the melted sodium, the fuel pins for irradiation tests are inspected. This paper shows the apparatuses and the technological basis for the fabrication of U-Pu-Zr alloy fuel pins for the irradiation test planned at the experimental fast test reactor Joyo. (author)

  9. Thermal Expansion Property of U-Zr Alloys and U-Zr-Ce Alloys as a Surrogate Metallic Fuel for SFR

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Lee, Jong Tak; Oh, Seok Jin; Ko, Young Mo; Kim, Ki Hwan; Woo, Youn Myung; Lee, Chan Bock

    2010-01-01

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. An extensive database on the performance of advanced metal fuels was generated as a result of the operation of these reactors and the IFR program. In this study, the U-Zr binary alloys and U-Zr-Ce ternary alloys as surrogate metallic fuel were fabricated in lower pressure Ar environment by gravity casting. The melt temperature was approximately 1,500 .deg. C. Thermal expansion of the fuel during normal operation is related with fuel performance in a reactor. Therefore, it is necessary to investigate the thermal expansion of the fuel in order to warrant a good prediction the fuel performance

  10. High temperature investigation of the solid/liquid transition in the PuO{sub 2}–UO{sub 2}–ZrO{sub 2} system

    Energy Technology Data Exchange (ETDEWEB)

    Quaini, A. [CEA, DANS/DPC/SCCME/LM2T, Centre de Saclay, 91191 Gif-sur-Yvette Cedex (France); Guéneau, C., E-mail: christine.gueneau@cea.fr [CEA, DANS/DPC/SCCME/LM2T, Centre de Saclay, 91191 Gif-sur-Yvette Cedex (France); Gossé, S. [CEA, DANS/DPC/SCCME/LM2T, Centre de Saclay, 91191 Gif-sur-Yvette Cedex (France); Sundman, B. [INSTN, CEA Saclay (France); Manara, D.; Smith, A.L.; Bottomley, D.; Lajarge, P.; Ernstberger, M. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Hodaj, F. [Univ. Grenoble Alpes, SIMAP, F-38000 Grenoble (France); CNRS, Grenoble INP, SIMAP, F-38000 Grenoble (France)

    2015-12-15

    The solid/liquid transitions in the quaternary U-Pu-Zr-O system are of great interest for the analysis of core meltdown accidents in Pressurised Water Reactors (PWR) fuelled with uranium-dioxide and MOX. During a severe accident the Zr-based cladding can become completely oxidised due to the interaction with the oxide fuel and the water coolant. In this framework, the present analysis is focused on the pseudo-ternary system UO{sub 2}–PuO{sub 2}–ZrO{sub 2}. The melting/solidification behaviour of five pseudo-ternary and one pseudo-binary ((PuO{sub 2}){sub 0.50}(ZrO{sub 2}){sub 0.50}) compositions have been investigated experimentally by a laser heating method under pre-set atmospheres. The effects of an oxidising or reducing atmosphere on the observed melting/freezing temperatures, as well as the amount of UO{sub 2} in the sample, have been clearly identified for the different compositions. The oxygen-to-metal ratio is a key parameter affecting the melting/freezing temperature because of incongruent vaporisation effects. In parallel, a detailed thermodynamic model for the UO{sub 2}–PuO{sub 2}–ZrO{sub 2} system has been developed using the CALPHAD method, and thermodynamic calculations have been performed to interpret the present laser heating results, as well as the high temperature behaviour of the cubic (Pu,U,Zr)O{sub 2±x}-c mixed oxide phase. A good agreement was obtained between the calculated and experimental data points. This work enables an improved understanding of the major factors relevant to severe accident in nuclear reactors.

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

    International Nuclear Information System (INIS)

    Meneghetti, D.; Kucera, D.A.

    1988-01-01

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

  12. First TREAT transient overpower tests on U-Pu-Zr fuel: M5 and M6

    International Nuclear Information System (INIS)

    Robinson, W.R.; Bauer, T.H.; Wright, A.E.; Rhodes, E.A.; Stanford, G.S.; Klickman, A.E.

    1987-01-01

    Transient Reactor Test Facility (TREAT) tests M5 and M6 were the first transient overpower (TOP) test of the margin to cladding breach and prefailure elongation of metallic U-Pu-Zr ternary fuel, the reference fuel of the integral fast reactor concept. Similar tests on U-5 wt% Fs fueled Experimental Breeder Reactor (EBR)-II driver pins were previously performed and reported. Results from these earlier tests indicated a margin to failure of ∼ 4 times nominal power and significant axial elongation prior to failure, a feature that was very pronounced at low burnups. While these two fuels types are similar in many respects, the ternary alloy exhibits a much more complex physical structure and is typically irradiated at much higher temperatures. Thus, a prime motivation for performing M5 and M6 was to compare the safety-related fuel performance characteristics of U-Fs and U-Pu-Zr. Tests M5 and M6 indicate that, under the TOP conditions used in the tests, ternary fuel displayed about the same margin to failure as U-Fs fuel. At low burnups, ternary fuel showed less prefailure axial elongation than observed in U-Fs pins, but elongations of 3 to 5% might turn out to be typical. Finally, fuel from the breached ternary pin in M6 showed, qualitatively, the same benignly dispersive behavior as U-Fs

  13. First TREAT [Transient Reactor Test Facility] transient overpower tests on U-Pu-Zr fuel: M5 and M6

    International Nuclear Information System (INIS)

    Robinson, W.R.; Bauer, T.H.; Wright, A.E.; Rhodes, E.A.; Stanford, G.S.; Klickman, A.E.

    1987-01-01

    Transient Reactor Test Facility (TREAT) tests M5 and M6 were the first transient overpower (TOP) tests of the margin to cladding breach and prefailure elongation of metallic U-Pu-Zr ternary fuel, the reference fuel of the Integral Fast Reactor concept. Similar tests on U-Fs fueled EBR-II driver pins were previously performed and reported [1,2]. Results from these earlier tests indicated a margin to failure of about 4 times nominal power and significant axial elongation prior to failure, a feature that was very pronounced at low burnups. While these two fuel types are similar in many respects, the ternary alloy exhibits a much more complex physical structure and is typically irradiated at much higher temperatures. Thus, a prime motivation for performing M5 and M6 was to compare the safety related fuel performance characteristics of U-Fs and U-Pu-Zr. This report described conditions, results, and conclusions of testing of these fuel types

  14. Performance of Zr as FCCI barrier layer for metallic fuel of fast reactor

    International Nuclear Information System (INIS)

    Kaity, Santu; Bhagat, R.K.; Kutty, T.R.G.; Kumar, Arun; Laik, A.; Kamath, H.S.

    2011-01-01

    Uranium-plutonium (U-Pu) and uranium-plutonium-zirconium (U-Pu-Zr) alloys have been considered as promising advanced fuels for fast reactor in India because of its high breeding potential, high thermal conductivity, high fissile and fertile atom densities, low doubling time and ease of fabrication compared to other ceramic fuels. The chemical compatibility between the fuel and clad material also known as fuel-clad chemical interaction (FCCI) has been recognized as one of the major concerns about the performance of the metallic fuel. Primarily, two design concepts have been proposed for the metallic fuel development programme for FBRs. One of them is based on sodium bonded ternary U-Pu-Zr alloy with T91 grade steel clad, and the other consists of binary U-Pu alloy mechanically bonded to T91 clad with a Zr liner between the fuel and clad. U will be the axial blanket material for U-Pu binary fuel. In the present investigation, the performance of Zr as FCCI barrier layer was studied through diffusion couple experiments of U/Zr/T91. A thin Zr foil (thickness ∼ 200 μm) sandwiched between U and T91 discs was kept inside a fixture made of Inconel 600 alloy. The fixture was encapsulated in quartz tube under Helium atmosphere and then heated at 650, 700 and 750 deg C for upto 1500 h. The extent of reaction and composition of phases formed were analyzed by scanning electron microscope (SEM) equipped with an energy dispersive spectrometer (EDS) and electron probe microanalyser (EPMA) equipped with wavelength dispersive spectrometer (WDS)

  15. Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

    International Nuclear Information System (INIS)

    Devan, K.; Bachchan, Abhitab; Riyas, A.; Sathiyasheela, T.; Mohanakrishnan, P.; Chetal, S.C.

    2011-01-01

    Highlights: → In this study we made physics designs of experimental metal fast reactor cores. → Aim is for full-scale demonstration of fuel assemblies in a commercial power reactor. → Minimum power with adequate safety is considered. → In addition, fuel sustainability is also considered in the design. → Sodium bonded U-Pu-6%Zr and mechanically bonded U-Pu alloys are used. - Abstract: Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  17. Optimization of the binary breeder reactor. VIII annular core fueled with 233U - 238U and Pu-238U

    International Nuclear Information System (INIS)

    Nascimento, J.A. do; Ishiguro, Y.

    1988-04-01

    First cycle burnup characteristics of a 1200 MWe binary breeder reactor with annular core fueled with metallic 233 U- 238 U-Zr, Pu- 238 U-Zr and Th in the blankets have been analysed. The Doppler effect is small as expected in a metal fueled fast reactor. The sodium void reactivity is, in general, smaller than in metal fueled homogeneous fast reactors of 1 m core height. The estimates of the required and available control rod worths show a large shutdown margin throughout the operational cycle. There are flexibilities in the blanket fueling and well balanced breeding in the two cycles, uranium and thorium, with doubling times of about 20 years are possible. (author) [pt

  18. Method of recovery of U, Pu and their compounds

    International Nuclear Information System (INIS)

    Cech, B.; Kaderabek, E.; Hanslik, T.

    1977-01-01

    A method is proposed of removing metal sheaths consisting of either Zr or Nb or Ti or an alloy of at least two of the metals from ceramic nuclear fuel based on U or Pu compounds. The system is exposed to hydrogen, preferably at temperatures of 250 to 460 degC and pressures of 20 to 50 at. The resulting zirconium hydride or hydrides of other metals is separated, eg., mechanically. (M.K.)

  19. Investigations of chemical reactions between U-Zr alloy and FBR cladding materials

    International Nuclear Information System (INIS)

    Ishii, Tetsuya; Ukai, Shigeharu

    2005-07-01

    U-Pu-Zr alloys are candidate materials for commercial FBR fuel. However, informations about chemical reactions with cladding materials developed by JNC for commercial FBR have not been well obtained. In this work, the reaction zones formed in four diffusion couples U-10wt.%Zr/PNC-FMS, U-10wt.%Zr/9Cr-ODS, U-10wt.%Zr/12Cr-ODS, and U-10wt.%Zr/Fe at about 1013K have been examined and following results were obtained. 1) At about 1013K, in the U-10wt.%Zr/Fe couple, the liquid phase zones were obtained. In the other couples U-10wt.%Zr/PNC-FMS, U-10wt.%Zr/9Cr-ODS and U-10wt.%Zr/12Cr-ODS, no liquid phase zones were obtained. The obtained chemical reaction zones in the later 3 couples were similar to the reported ones obtained in U-Zr/Fe couples without liquid phase formation. In comparison with the reaction zones obtained in the U-10wt.%Zr/Fe couple, the reaction zones inside cladding materials obtained in the PNC-FMS, 9Cr-ODS, and 12Cr-ODS couples were thin. 2) From the investigations of relationship between the obtained depths of the chemical reaction zones inside cladding materials and composition of the cladding materials, it was considered that the depth of chemical reaction zone would depend on the Cr content of the cladding materials and the depth would decrease with increasing Cr content, resulting in prevention of liquid phase formation. 3) From the investigations of the mechanisms of chemical reactions between U-Pu-Zr/cladding materials, it was considered that the same effect of Cr obtained in the U-Zr/cladding materials would be expected in U-Pu-Zr/cladding materials. Those seemed to indicate that the threshold temperatures of liquid phase formation for U-Pu-Zr/PNC-FMS, U-Pu-Zr/9Cr-ODS, and U-Pu-Zr/12Cr-ODS might be higher than that for U-Pu-Zr/Fe. (author)

  20. Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II [Experimental Breeder Reactor

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  1. Advancing Understanding of the +4 Metal Extractant Thenoyltrifluoroacetonate (TTA-); Synthesis and Structure of MIVTTA4 (MIV = Zr, Hf, Ce, Th, U, Np, Pu) and MIII(TTA)4- (MIII = Ce, Nd, Sm, Yb).

    Science.gov (United States)

    Cary, Samantha K; Livshits, Maksim; Cross, Justin N; Ferrier, Maryline G; Mocko, Veronika; Stein, Benjamin W; Kozimor, Stosh A; Scott, Brian L; Rack, Jeffrey J

    2018-04-02

    Thenoyltrifluoroacetone (HTTA)-based extractions represent popular methods for separating microscopic amounts of transuranic actinides (i.e., Np and Pu) from macroscopic actinide matrixes (e.g. bulk uranium). It is well-established that this procedure enables +4 actinides to be selectively removed from +3, + 5, and +6 f-elements. However, even highly skilled and well-trained researchers find this process complicated and (at times) unpredictable. It is difficult to improve the HTTA extraction-or find alternatives-because little is understood about why this separation works. Even the identities of the extracted species are unknown. In addressing this knowledge gap, we report here advances in fundamental understanding of the HTTA-based extraction. This effort included comparatively evaluating HTTA complexation with +4 and +3 metals (M IV = Zr, Hf, Ce, Th, U, Np, and Pu vs M III = Ce, Nd, Sm, and Yb). We observed +4 metals formed neutral complexes of the general formula M IV (TTA) 4 . Meanwhile, +3 metals formed anionic M III (TTA) 4 - species. Characterization of these M(TTA) 4 x- ( x = 0, 1) compounds by UV-vis-NIR, IR, 1 H and 19 F NMR, single-crystal X-ray diffraction, and X-ray absorption spectroscopy (both near-edge and extended fine structure) was critical for determining that Np IV (TTA) 4 and Pu IV (TTA) 4 were the primary species extracted by HTTA. Furthermore, this information lays the foundation to begin developing and understanding of why the HTTA extraction works so well. The data suggest that the solubility differences between M IV (TTA) 4 and M III (TTA) 4 - are likely a major contributor to the selectivity of HTTA extractions for +4 cations over +3 metals. Moreover, these results will enable future studies focused on explaining HTTA extractions preference for +4 cations, which increases from Np IV to Pu IV , Hf IV , and Zr IV .

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

    Energy Technology Data Exchange (ETDEWEB)

    Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Tsai Hanchung [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439-4803 (United States)

    2012-08-15

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

  3. U-Zr-RE Fuel Alloy with Minor Actinides

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hoon; Kim, Jong Hwan; Ko, Young Mo; Kim, Ki Hwan; Park, Jeong Yong; Lee, Chan Bock [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Metallic fuels, such as the U-Pu-Zr alloys, have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing the amount of highly radioactive spent nuclear fuels since the 1980s. Metallic fuels fit well with such a concept owing to their high thermal conductivity, high thermal expansion, compatibility with a pyro-metallurgical reprocessing scheme, and their demonstrated fabrication at engineering scale in a remote hot cell environment. To increase the productivity and efficiency of the fuel fabrication process waste streams must be minimized and fuel losses quantified and reduced to lower levels. In this study, U-Zr alloy system fuel slugs were fabricated by an injection casting method. After casting a considerable number of fuel slugs in the casting furnaces, the fuel loss in the melting chamber, the crucible, and the molds have been evaluated quantitatively.

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

    International Nuclear Information System (INIS)

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

    1992-01-01

    Current fabrication techniques for the integral fast reactor (IFR) fuel utilize injection casting into quartz molds after reprocessing in the IFR fuel cycle facility. The quartz molds are destroyed during the fuel demolding process, and the quartz residue must therefore be treated as contaminated waste. Alternatively, if the fuel can be cast into molds that remain as part of the fuel slugs (i.e., if the fuel can be left inside the molds for irradiation), then the quartz mold contribution to the waste stream can be eliminated. This possibility is being addresssed in an ongoing effort to evaluate the irradiation performance of fuel cast into zirconium sheaths rather than quartz molds. Zirconium was chosen as the sheath material because it is the component of the U-Pu-Zr fuel alloy that raises the alloy solidus temperatures and provides resistance to fuel-cladding chemical interaction (FCCI)

  5. Influence of the allotropic beta to alpha transformation and of oxygen on the distribution of plutonium in Zr-Pu and Ti-Pu dilute alloys

    CERN Document Server

    Languille, A; Remy, C

    1972-01-01

    Ti-Pu, Zr-Pu and Ti-U alloys, annealed above the transformation point, can show major segregation after cooling. The segregations increase with oxygen content. Annealing at high temperatures followed by severe quenching can cause them to disappear. (7 refs).

  6. Status of solubility data for selected elements (U, Mp, Pu, Am, Te, Ni, and Zr)

    International Nuclear Information System (INIS)

    Moll, H.; Brachmann, A.; Wruck, D.; Palmer, C.

    1997-01-01

    This report is an evaluation of solubility data for U, Np, Pu, Am, Tc, Ni and Zr compounds at ambient and elevated temperatures. We review the status of such data in light of the most recently reported experimental results. The focus is on the solid phases that may control solubilities under expected conditions in and near a potential nuclear waste repository at Yucca Mountain, Nevada. Solubility data or reliable predictions over the temperature range 20 to 150 degrees C will be used in geochemical modeling studies of the Yucca Mountain Project [96PAL

  7. Thermodynamic activity measurements of U-Zr alloys by Knudsen effusion mass spectrometry

    International Nuclear Information System (INIS)

    Kanno, Masayoshi; Yamawaki, Michio; Koyama, Tadafumi; Morioka, Nobuo

    1988-01-01

    Vaporization of a series of U-Zr alloys, a fundamental subsystem of the promising metallic fuel U-Pu-Zr, was studied by using a tantalum Knudsen cell coupled with a mass spectrometer in the temperature range 1700-2060 K. Thermodynamic activities partial molar Gibbs free energies and integral molar Gibbs free energies of mixing were calculated from the partial vapor pressures of uranium over these alloys. The activities of uranium exhibit negative deviations from ideality, especially in the uranium-rich composition region. Both the solidus and liquidus lines for this system estimated from the activities show negative deviations from the tentative phase diagram previously reported. (orig.)

  8. Thermodynamic evaluation of liquid Cd cathode containing U and Pu

    International Nuclear Information System (INIS)

    Kurata, Masaki; Uozumi, Koichi; Kato, Tetsuya; Iizuka, Masatoshi

    2009-01-01

    In our previous study, a mixture of U and Pu was recovered in liquid Cd cathode from molten salt under various conditions of the U:Pu ratio. Two important things were observed. The first was that three kinds of precipitated phase had been detected in the saturated liquid Cd cathode, such as a U metal and two kinds of U-Pu-Cd compound. The compositions of the compounds were roughly (U+Pu):Cd=1:11 and (U+Pu):Cd=1:6. The second was that the U metal had selectively precipitated in the saturated liquid Cd cathode under the condition that the U:Pu ratio is higher than about 0.8 in the liquid Cd phase. In the present study, phase diagrams were evaluated by the CALPHAD method on the liquid Cd cathode containing U and Pu. The U-Pu-Cd compounds were modeled as MCd 11 -type and MCd 6 -type, respectively, based on the reported binary phase diagrams of U-Cd and Pu-Cd. The calculated result reasonably agreed with the experimental observations. The variations in the U and Pu activities were estimated along with the U:Pu ratio, which is related to the precipitation of various phases in the liquid Cd cathode. (author)

  9. Compacted and Sintered Microstructure Depending on Uranium Powder Size in Zr-U Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Chang Gun; Jun, Hyun-Joon; Ju, Jung Hwan; Lee, Ho Jin; Lee, Chong-Tak; Kim, Hyung Lae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-03-15

    In case of the uranium (U) and zirconium (Zr) powders which have been utilized for the production of a metallic fuel in the various nuclear applications, the homogenous distribution of U powders in the Zr-U pellet has influenced significantly on the nuclear fuel performance. The inhomogeneity in a powder process was changed by various intricate factors, e.g. powder size, shape, distribution and so on. Particularly, the U inhomogeneity in the Zr-U pellets occurs by segregation derived from the great gaps of densities between Zr and U during compaction of the mixed powders. In this study, the relationship between powder size and homogeneity was investigated by using the different-sized U powders. The microstructure in Zr-U pellets reveals more homogeneity when the weight ration of Zr and U powders are close to 1. In addition, homogeneous pellets which were produced by fine U powders have higher density because the homogeneity affects the alloying reaction during sintering and the densification behavior of pore induced by powder size.

  10. Fabrication of U-Pu-Zr metallic fuel containing minor actinides

    International Nuclear Information System (INIS)

    Kurata, Masaki; Sasahara, Akihiro; Inoue, Tadashi; Betti, M.; Babelot, J.F.; Spirlet, J.C.; Koch, L.

    1997-01-01

    Rods of UPuZr alloy containing 5% minor actinides, 2% minor actinides and 2% rare-earth elements, and 5% minor actinides and 5% rare-earth elements have been fabricated by casting in yttria molds. Parts of the ingots were cut off for quantitative analysis and the rods characterized to the required extent, which included measurement of length, weight, diameter, and bending. For selected samples, metallographic study was carried out to examine the dispersion of the various phases contained in the alloy. Finally, the rods were encapsulated in stainless steel pin with the UPuZr reference after sodium bonding for the irradiation study. (author)

  11. Nitridation of U and Pu recovered in liquid Cd cathode by molten salt electrorefining of (U,Pu)N

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Takumi; Iwai, Takashi; Arai, Yasuo [Japan Atomic Energy Agency (Japan)

    2009-06-15

    Solid solutions of actinide mono-nitrides have been proposed as a candidate fuel of the accelerator-driven system (ADS) and Gen.IV-type fast reactors because the thermal conductivity and metal density are higher than those of actinide oxides and also they have high melting temperature. Pyrochemical process has several advantages over conventional wet process in treating of spent nitride fuel. One of the key technologies of the pyrochemical reprocessing of nitride fuel is the formation of the nitrides from actinides in the liquid Cd cathode. The nitridation-distillation combined method was developed and has been adopted to convert the actinides to the nitrides. In this method, the nitridation of actinides and the distillation of Cd occurred simultaneously by heating the actinide-Cd alloys in N{sub 2} gas stream. In the present study, the nitride formation behavior of U and Pu recovered in Cd cathode by molten salt electrorefining of (U,Pu)N was experimentally investigated. In addition, the nitride pellet was prepared form the powder obtained by the nitridation of U and Pu recovered in Cd cathode. (U,Pu)N (PuN = 80 mol %) was used as the starting material in the experiment. Molten salt electrorefining of (U,Pu)N pellet was carried out in the LiCl-KCl eutectic salt with 1.2 wt% PuCl{sub 3} and 0.3 wt% UCl{sub 3} of about 110 g at the constant anodic potential of -0.60 to -0.55 V vs. Ag/AgCl for about 9 hours at 773 K. After the electrorefining, about 42 % of U and Pu in the starting (U,Pu)N pellet was dissolved at the anode and recovered into the liquid Cd cathode. The recovered U-Pu-Cd alloy was heated in an alumina crucible at 973 K for 10 hours under N{sub 2} gas (99.999 %) stream (0.015 L/min). Fine black powder was recovered after heating the U-Pu-Cd alloy. The powder was identified as the single phase solid solution of (U,Pu)N by the XRD analysis. After milling in the agate mortar for 1 hour, the powder was compacted into green pellet under a pressure of about

  12. Change of Composition in Metallic Fuel Slug of U-Zr Alloy from High-Temperature Annealing

    Energy Technology Data Exchange (ETDEWEB)

    Youn, Young Sang; Lee, Jeong Mook; Kim, Jong Yun; Kim, Jong Hwan; Song, Hoon [KAERI, Daejeon (Korea, Republic of)

    2016-09-15

    The U–Zr alloy is a candidate for fuel to be used as metallic fuel in sodium-cooled fast reactors (SFRs). Its chemical composition before and after annealing at the operational temperature of SFRs (610 .deg. C) was investigated using X-ray photoelectron spectroscopy, Raman spectroscopy, and X-ray diffraction. The original alloy surface contained uranium oxides with the U(IV) and U(VI) oxidation states, Zr{sub 2}O{sub 3}, and a low amount of uranium metal. After annealing at 610 .deg. C, the alloy was composed of uranium metal, uranium carbide, uranium oxide with the U(V) valence state, zirconium metal, and amorphous carbon. Meanwhile, X-ray diffraction data indicate that the bulk composition of the alloy remained unchanged.

  13. Change of Composition in Metallic Fuel Slug of U-Zr Alloy from High-Temperature Annealing

    International Nuclear Information System (INIS)

    Youn, Young Sang; Lee, Jeong Mook; Kim, Jong Yun; Kim, Jong Hwan; Song, Hoon

    2016-01-01

    The U–Zr alloy is a candidate for fuel to be used as metallic fuel in sodium-cooled fast reactors (SFRs). Its chemical composition before and after annealing at the operational temperature of SFRs (610 .deg. C) was investigated using X-ray photoelectron spectroscopy, Raman spectroscopy, and X-ray diffraction. The original alloy surface contained uranium oxides with the U(IV) and U(VI) oxidation states, Zr 2 O 3 , and a low amount of uranium metal. After annealing at 610 .deg. C, the alloy was composed of uranium metal, uranium carbide, uranium oxide with the U(V) valence state, zirconium metal, and amorphous carbon. Meanwhile, X-ray diffraction data indicate that the bulk composition of the alloy remained unchanged

  14. Chlorination of UO2, PuO2, and rare-earth oxides using ZrCl4

    International Nuclear Information System (INIS)

    Sakamura, Yoshiharu; Inoue, Tadashi; Iwai, Takashi; Moriyama, Hirotake

    2001-01-01

    A new chlorination method using ZrCl 4 , which has a high reactivity with oxygen, has been investigated for more efficient oxide treatment. After actinide oxides are chlorinated and dissolved in a molten salt bath, actinide metals can be selectively collected using the electrorefining process. This process is well suited for pyrochemical reprocessing of metallic fuels. In LiCl-KCI eutectic melts, rare-earth oxides (Y 2 O 3 , La 2 O 3 , CeO 2 , and Nd 2 O 3 ) and actinide oxides (UO 2 and PuO 2 ) were chlorinated by adding ZrCl 4 . As a result, rare-earth and actinide elements were dissolved into the salt as trivalent ions and ZrO 2 was precipitated. When an excess of ZrCI 4 was added, oxides in powder form were completely chlorinated in five hours. It was demonstrated that the ZrCI 4 chlorination method, free from corrosive gas such as chlorine, was very simple and useful. (author)

  15. Potential of multi-purpose liquid metallic fuelled fast reactor (MPFR) as a hydrogen production system

    International Nuclear Information System (INIS)

    Endo, H.; Ninokata, H.; Netchaev, A.; Sawada, T.

    2001-01-01

    Nuclear energy is the only effective alternative energy source to fossil fuels in the next century. Therefore future nuclear power plants should satisfy the following three requirements: i) multiple energy conversion capability with high temperature not only for electricity generation but also for hydrogen production, ii) extended siting capability so as to eliminate on-site refuelling, and iii) passive safety features. An aim of this paper is to describe the basic concept of the multi-purpose liquid metallic fuelled fast reactor system (MPFR). The MPFR introduces the U-Pu-X (X: Mn, Fe, Co) liquid metallic alloy with Ta and Ta/TaC structural materials, and satisfies all of the conditions listed above based on the following characteristics of the liquid metallic fuel: high temperature operation between 650 deg C (sodium-cooled system) and 1 200 deg C (lead-cooled system), a core lifetime of 15-30 years without radiation damage of fuel materials, and enhanced passive safety by the thermal expansion of liquid fuel and the avoidance of re-criticality due to local core fuel dispersion at fuel failure events. (authors)

  16. Experimental determination of liquidus of Fe-Zr by spot technique

    International Nuclear Information System (INIS)

    Ramakrishna, P.; Samanta, B.; Balakrishnan, S.

    2016-01-01

    Metallic fuel alloy for fast reactor mainly consist of U-Pu-Zr housed in T91 clad. Study of thermophysical properties of fuel element and cladding material is vital for the fuel designer to optimize the design feature and predict the fuel behavior under reactor operating conditions.To understand the fuel-clad interaction the phase diagram study of Fe-Zr system is very important since future reactors use U-Pu-Zr alloy as fuel and stainless steel as clad. The eutectic temperature in Fe-Zr alloy system has been established experimentally by various methods. Information on the liquidus temperatures of Fe-Zr is scanty in the literature excepting a very few experimental measurements. Hence measurement of liquidus temperatures is very essential to establish the phase diagram. Present work concentrates more on the generation of liquidus data of Fe-Zr binary alloy system by Spot-technique. This is one among the advanced techniques for measuring the solid-liquid phase transition temperatures. (author)

  17. Influence of Temperature to Thermal Properties of U-Zr Alloy With The Zr Content Variation

    International Nuclear Information System (INIS)

    Aslina-Br-Ginting; Masrukan; M-Husna-Al-Hasa

    2007-01-01

    Have been done thermal of characteristic covering heat stability, heat capacities, enthalpy and also phase changes from uranium, zirkonium and U-Zr alloy with the Zr content variation of Zr 2 %, 6 %, 10% and 14% weight. Change of the temperature and composition anticipated will cause the characteristic of thermal to uranium metal, zirkonium and also U-Zr alloy. Therefore at this research was conducted using analysis influence of temperature to thermal of characteristic of uranium, zirkonium and U-Zr alloy with the Zr content variation by using DTA and DSC. Result of analysis indicate that the uranium metal at temperature 662 o C stable in phase α. Above at temperature, uranium metal experience of the phase change indicated by formed the thermochemical reaction as much 3 endothermic peak. At temperature 667.16 o C, happened by the phase change of α become the phase β with the enthalpy 2,3034 cal/g, at temperature 773.05 o C happened by the phase change β becoming phase γ 2,8725 cal/g and also at temperature 1125.26 the o C uranium metal experience the phenomenon become to melt with the enthalpy 2,1316 cal/g. (author)

  18. Respiratory tract tumors in Syrian hamsters following inhalation of Pu--ZrO2 particles

    International Nuclear Information System (INIS)

    Thomas, R.G.; Smith, D.M.

    1979-01-01

    Inhalation of radionuclide-bearing particles remains one of the most intensely pursued problems concerning the nuclear industry. This route of entry is generally accepted as the most probable, in case of human exposure, with ingestion being the other prominent source of concern. Many laboratory investigations, such as those reported here, continue to evaluate the possible consequences that may present health problems to the public domain. Syrian hamsters of both sexes received either inhaled (INH) PuO 2 /ZrO 2 particles, intravenous (IV) PuO 2 /ZrO 2 microspheres, a combination of INH PuO 2 /ZrO 2 particles and injected PuO 2 /ZrO 2 microspheres, or no radionuclides (controls). The INH particles and IV microspheres were tagged with γ-emitting 57 Co to facilitate whole body counting and establishment of retention curves. Total lung burdens ranged from 8 nCi to 143 nCi. Significant numbers of primary lung tumors (5 to 50% per group) were induced in those animals that received INH exposures. Additional α radiation administered via Pu-laden IV microspheres had little or no effect on tumor production or nonneoplastic, degenerative changes in the respiratory tract

  19. Neutron inelastic-scattering cross sections of 232Th, 233U, 235U, 238U, 239Pu and 240Pu

    International Nuclear Information System (INIS)

    Smith, A.B.; Guenther, P.T.

    1982-01-01

    Differential-neutron-emission cross sections of 232 Th, 233 U, 235 U, 238 U, 239 Pu and 240 Pu are measured between approx. = 1.0 and 3.5 MeV with the angle and magnitude detail needed to provide angle-integrated emission cross sections to approx. 232 Th, 233 U, 235 U and 238 U inelastic-scattering values, poor agreement is observed for 240 Pu, and a serious discrepancy exists in the case of 239 Pu

  20. Review of the thermodynamics of the U--C, Pu--C, and U--Pu--C systems

    International Nuclear Information System (INIS)

    Tetenbaum, M.; Sheth, A.; Olson, W.

    1975-06-01

    Thermodynamic properties such as enthalpy, heat capacity, entropy, heat and free energy of formation, and vaporization behavior are presented for the U--C, Pu--C, and U--Pu--C systems. These properties are of interest to scientists and engineers involved in the expanding field of advanced fuel LMFBR systems. The information on these systems has been derived largely from the discussions of the IAEA Panel on the assessment of thermodynamic properties of the U--C, Pu--C, and U--Pu--C systems. (U.S.)

  1. Post-irradiation examination of CANDU fuel bundles fuelled with (Th, Pu)O2

    International Nuclear Information System (INIS)

    Karam, M.; Dimayuga, F.C.; Montin, J.

    2010-01-01

    AECL has extensive experience with thoria-based fuel irradiations as part of an ongoing R&D program on thorium within the Advanced Fuel Cycles Program. The BDL-422 experiment was one component of the thorium program that involved the fabrication and irradiation testing of six Bruce-type bundles fuelled with (Th, Pu)O 2 pellets. The fuel was manufactured in the Recycle Fuel Fabrication Laboratories (RFFL) at Chalk River allowing AECL to gain valuable experience in fabrication and handling of thoria fuel. The fuel pellets contained 86.05 wt.% Th and 1.53 wt.% Pu in (Th, Pu)O 2 . The objectives of the BDL-422 experiment were to demonstrate the ability of 37-element geometry (Th, Pu)O 2 fuel bundles to operate to high burnups up to 1000 MWh/kgHE (42 MWd/kgHE), and to examine the (Th, Pu)O 2 fuel performance. This paper describes the post-irradiation examination (PIE) results of BDL-422 fuel bundles irradiated to burnups up to 856 MWh/kgHE (36 MWd/kgHE), with power ratings ranging from 52 to 67 kW/m. PIE results for the high burnup bundles (>1000 MWh/kgHE) are being analyzed and will be reported at a later date. The (Th, Pu)O 2 fuel performance characteristics were superior to UO 2 fuel irradiated under similar conditions. Minimal grain growth was observed and was accompanied by benign fission gas release and sheath strain. Other fuel performance parameters, such as sheath oxidation and hydrogen distribution, are also discussed. (author)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  4. Evaluation of fission cross sections and covariances for 233U, 235U, 238U, 239Pu, 240Pu, and 241Pu

    International Nuclear Information System (INIS)

    Kawano, Toshihiko; Matsunobu, Hiroyuki; Murata, Toru

    2000-02-01

    A simultaneous evaluation code SOK (Simultaneous evaluation on KALMAN) has been developed, which is a least-squares fitting program to absolute and relative measurements. The SOK code was employed to evaluate the fission cross sections of 233 U, 235 U, 238 U, 239 Pu, 240 Pu, and 241 Pu for the evaluated nuclear data library JENDL-3.3. Procedures of the simultaneous evaluation and the experimental database of the fission cross sections are described. The fission cross sections obtained were compared with evaluated values given in JENDL-3.2 and ENDF/B-VI. (author)

  5. Americium and plutonium in phosphates of trigonal structure (NZP type) Am1/3[Zr2(PO4)3] and Pu1/4[Zr2(PO4)3

    International Nuclear Information System (INIS)

    Bykov, D.M.; Orlova, A.I.; Tomilin, S.V.; Lizin, A.A.; Lukinykh, A.N.

    2006-01-01

    Am 1/3 [Zr 2 (PO 4 ) 3 ] and Pu 1/4 [Zr 2 (PO 4 ) 3 ] phosphates are synthesized and are investigated by X-ray diffraction method. Compounds have triclinic lattices and lattice parameters are determined. Possibility of actinide inclusion into hollows of framework of NZP type is shown for the first time. It is proposed that inclusion of Pu and Am highly-charged cations into framework hollows decreases crystal structure symmetry up to primitive trigonal one. Rate of Pu leaching from ceramics on Pu 1/4 [Zr 2 (PO 4 ) 3 ] basis are measured [ru

  6. Determination of 233U, 235U, 238U and 239Pu fission yields induced by fission and 14.7 MeV neutrons

    International Nuclear Information System (INIS)

    Laurec, Jean; Adam, Albert; Bruyne, Thierry de.

    1981-12-01

    The 233 U, 235 U, 238 U, 239 Pu fission yields have been determined by a radiochemical method. A target and a fission chamber made of same fissible material are irradied together. The total fission number is measured from the fission chamber. The fission product activities are directly measured on the target using calibrated Ge-Li detectors. The fissible material masses are determined by alpha and mass spectrometries. The irradiations were made on the critical assemblies PROSPERO and CALIBAN and on the 14 MeV neutron generator of C.E. VALDUC. 3 to 5% fission yield errors are got for the most measured nuclides: 95 Zr, 97 Zr, 99 Mo, 103 Ru, 131 I, 132 Te, 140 Ba, 141 Ce, 143 Ce, 144 Ce, 147 Nd [fr

  7. Development of program for evaluating the temperature of Zr-U metallic fuel rod

    International Nuclear Information System (INIS)

    Chun, J. S.; Lee, B. H.; Ku, Y. H.; Oh, J. Y.; Im, J. S.; Sohn, D. S.

    2003-01-01

    A code for evaluating the temperature of Zr-U metallic rod has been developed. Finite element (FE) method is adopted for the developed code sharing the user subroutines which has been prepared for the ABAQUS commercial FE code. The developed program for the Zr-U metallic fuel rod corresponds to a nonlinear transient heat transfer problem, and uses a sparse matrix solver for FE equations during iterations at every time step. The verifications of the developed program were conducted using the ABAQUS code. Steady state and transient problems were analyzed for 1/8 rod model due to the symmetry of the fuel rod and full model. From the evaluation of temperature for the 1/8 rod model at steady state, maximal error of 0.18 % was present relative to the ABAQUS result. Analysis for the transient problem using the fuel rod model resulted in the same as the variation of centerline temperature from the ABAQUS code during a hypothetical power transient. The distribution of heat flux for the entire cross section and surface was almost identical for the two codes

  8. Contribution to the study of U-Ti and U-Pu-Ti carbides; Contribution a l'etude des carbures U-C-Ti et (U, Pu) - C-Ti

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-07-01

    After having discussed the reasons to use (U,Pu) carbides as fast reactor fuel, we examine the influence of the addition of titanium to these carbides. A preliminary study has been done on the system of U-C-Ti and some properties have been measured such as: density, thermal expansion, electrical resistivity, atmospheric corrosion and compatibility with stainless steel. The systems U-Pu-C-Ti (Pu/U + Pu equal to 15 per cent) and U-C-Ti have been found to be very similar. There exists a two phases region (U,Pu)C + TiC, an eutectic between (U,Pu)C and TiC for approximately 15 at %. The solubilities of U + Pu in TiC and of Ti in (U,Pu)C is less than 1 % at. The addition of titanium does not markedly change thermal expansion coefficients of (U,Pu)C. However the resistance to atmospheric corrosion and compatibility with stainless steel is improved. Thermal conductivity, calculated from electrical resistivity, has increased. On the other side, the density of fissile material is lowered. The combination of (U,Pu)C + TiC seems to be the most promising alloy for application as nuclear fuel. (author) [French] Apres avoir rappele les problemes poses par un combustible pour les reacteurs a neutrons rapides et l'interet des carbures U-Pu-C comme combustible, on examine l'influence de l'addition de titane dans ces carbures. Une etude preliminaire sur le systeme U-C-Ti a ete effectuee et quelques proprietes sont indiquees: densite, coefficients de dilatation, resistivite electrique, tenue a la corrosion atmospherique, compatibilite avec l'acier inoxydable. Le systeme U-Pu-C-Ti (Pu/U + Pu egal a 15 pour cent) presente de grandes analogies avec le systeme U-C-Ti. Il existe un domaine biphase (U,Pu)C + TiC, un eutectique entre (U,Pu)C et TiC pour environ 15 at % Ti; les solubilites de U + Pu clans TiC et de Ti dans (U,Pu)C sont inferieures a 1 at %. Par rapport a la phase (U,Pu)C, l'addition de titane est sans effets importants sur les coefficients de dilatation. Par contre la tenue a

  9. Studies of U-Pu-C-Ti compounds

    International Nuclear Information System (INIS)

    Milet, C.

    1967-01-01

    The U-Pu-C-Ti compounds (5 to 20 atoms per cent Ti) have been studied in order to improve some properties of U-Pu-C carbides and to extend the existence field of the (U, Pu) C phase. The Pu/(U+Pu) ratio has been fixed to 15 per cent. All the alloys were elaborated and cast in an arc furnace. A two-phases field (U, Pu) C + Ti C exists which permits to avoid di- and sesqui-carbides and the (U, Pu) phase. An eutectic between (U, Pu) C and Ti C was found around 15 atoms per cent Ti. Practically the whole of the titanium is in Ti C form, titanium solubility in (U, Pu) C being inferior to 1 atom per cent. The most promising alloy are those containing two phases: (U, Pu) C+ Ti C. In comparison with the (U, Pu) C phase, titanium addition does not change very much the thermal expansion coefficients nor the thermal cycling behaviour between 160 and 1000 Celsius degrees which is excellent. On the other hand atmospheric corrosion behaviour is improved; compatibility with stainless steel is better; thermal conductivity, calculated from electrical resistivity K is enhanced: for U(0.85)-Pu(0.15)-C alloy we have K 0.179 W/cm.C at 1000 C and K = 0.187 W/cm.C at 1500 C, for U-Pu-C-Ti (10 atoms % Ti) alloy we have K = 0.193 W/cm.C at 1000 C and K = 0.205 W/cm.C at 1500 C. (author) [fr

  10. Pu(IV) reduction with hydroxyurea and its application in U/Pu separation

    International Nuclear Information System (INIS)

    Zhu Zhaowu; He Jianyu; Zhang Zefu; Song Tianbao; Zhang Yu; Zheng Weifang

    2004-01-01

    The reduction of Pu(IV) with hydroxyurea (HU) in the mixed phase of 30% TBP/OK-HNO 3 system was studied. The study results show that HU can reduce Pu(IV) to Pu(III) and the reduction rate can be expressed as -dc(Pu(IV))/dt=kc(HU)c -3.2 (HNO 3 )c mix 2 (Pu(IV))c mix -1 (Pu(III)), where k is the rate constant, and k=(896 ± 59) mol 2.3 ·L -2.3 ·min -1 at 15 degree C. With HU serves as a reductant, 16 stages count-current cascade experiment was performed using centrifugal tube to simulate U/Pu separation in the 1B contactor of Purex process. The separation factor of Pu from U and the separation factor of U from Pu reach to 5.4 x 10 4 and 1.8 x 10 5 , respectively. The amount of Pu is about 11 μg in per kg U product. (authors)

  11. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    International Nuclear Information System (INIS)

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-01-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at ∼2.4, ∼7 and ∼11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of ∼7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10 15 n/cm 2 /s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between ∼410 deg. C and ∼645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

  12. Mass spectrometric study of vaporization of (U,Pu)O2 fuel simulating high burnup

    International Nuclear Information System (INIS)

    Maeda, Atsushi; Ohmichi, Toshihiko; Fukushima, Susumu; Handa, Muneo

    1985-08-01

    The vaporization behavior of (U,Pu)O 2 fuel simulatig high burnup was studied in the temperature range of 1,573 -- 2,173 K by high temperature mass spectrometry. The phases in the simulated fuel were examined by X-ray microprobe analysis. The relationship between chemical form and vaporization behavior of simulated fission product elements was discussed. Pd, Sr, Ba, Ce and actinide-bearing vapor species were observed, and it was clarified that Pd vapor originated from metallic inclusion and Sr and Ce vapors, from mixed oxide fuel matrix. The vaporization behavior of the actinide elements was somewhat similar to that of hypostoichiometric mixed oxide fuel. The behavior of Ba-bearing vapor species changed markedly over about 2,000 K. From the determination of BaO vapor pressures over simulated fuel and BaZrO 3 , it was revealed thermodynamically that the transformation of the chemical form of Ba about 2,000 K, i.e., dissolution of BaZrO 3 phase into fuel matrix, might be the reason of the observed vapor pressure change. (author)

  13. A study of accelerated radiation damage effects in PuO{sub 2} and gadolinia-stabilized cubic zirconia, Zr{sub 0.79}Gd{sub 0.14}Pu{sub 0.07}O{sub 1.93}, doped with {sup 238}Pu

    Energy Technology Data Exchange (ETDEWEB)

    Burakov, B.E., E-mail: burakov@peterlink.ru [V.G. Khlopin Radium Institute, 28, 2-nd Murinskiy Ave., St. Petersburg, 194021 (Russian Federation); Yagovkina, M.A. [A.F. Ioffe Physical-Technical Institute, 26, Politekhnicheskaya, St. Petersburg, 194021 (Russian Federation)

    2015-12-15

    Polycrystalline samples of cubic zirconia, Zr{sub 0.79}Gd{sub 0.14}Pu{sub 0.07}O{sub 1.93}, doped with approximately 9.9 wt.% {sup 238}Pu, and PuO{sub 2} containing 11.0 wt. % {sup 238}Pu (and main isotope is {sup 239}Pu) have been repeatedly studied during many years by X-ray diffraction analysis. At a temperature of 25 °C the unit-cell parameter of PuO{sub 2} increases depending on accumulated dose, and is accompanied by decrease of coherent scattering region (CSR). Self-irradiation of Zr{sub 0.79}Gd{sub 0.14}Pu{sub 0.07}O{sub 1.93} is accompanied with repeated change of unit-cell parameter and CSR.

  14. Simulation of Zr content in TiZrCuNi brazing filler metal for Ti6Al4V alloy

    International Nuclear Information System (INIS)

    Yue, Xishan; Xie, Zonghong; Jing, Yongjuan

    2017-01-01

    To optimize the Zr content in Ti-based filler metal, the covalent electron on the nearest atoms bond in unit cell (n_A"u"-"v) with Ti-based BCC structure was calculated, in which the brazing temperature was considered due to its influence on the lattice parameter. Based on EET theory (The Empirical Electron Theory for solid and molecules), n_A"u"-"v represents the strength of the unit cell with defined element composition and structure, which reflects the effect from solid solution strengthening on the strength of the unit cell. For Ti-Zr-15Cu-10Ni wt% filler metal, it kept constant as 0.3476 with Zr as 37.5 ∝ 45 wt% and decreased to 0.333 with Zr decreasing from 37.5 to 25 wt%. Finally, it increased up to 0.3406 with Zr as 2 ∝ 10 wt%. Thus, Ti-based filler metal with Zr content being 2 ∝ 10 wt% is suggested based on the simulation results. Moreover, the calculated covalent electron of n_A"u"-"v showed good agreement with the hardness of the joint by filler 37.5Zr and 10Zr. The composition of Ti-10Zr-15Cu-10Ni wt% was verified in this study with higher tensile strength of the brazing joint and uniform microstructure of the interface. (orig.)

  15. Simulation of Zr content in TiZrCuNi brazing filler metal for Ti6Al4V alloy

    Energy Technology Data Exchange (ETDEWEB)

    Yue, Xishan [Northwestern Polytechnical University, School of Astronautics, Xi' an (China); AVIC Beijing Aeronautical Manufacturing Technology Research Institute, Aeronautical Key Laboratory for Welding and Joining Technology, Beijing (China); Xie, Zonghong [Northwestern Polytechnical University, School of Astronautics, Xi' an (China); Jing, Yongjuan [AVIC Beijing Aeronautical Manufacturing Technology Research Institute, Aeronautical Key Laboratory for Welding and Joining Technology, Beijing (China)

    2017-07-15

    To optimize the Zr content in Ti-based filler metal, the covalent electron on the nearest atoms bond in unit cell (n{sub A}{sup u-v}) with Ti-based BCC structure was calculated, in which the brazing temperature was considered due to its influence on the lattice parameter. Based on EET theory (The Empirical Electron Theory for solid and molecules), n{sub A}{sup u-v} represents the strength of the unit cell with defined element composition and structure, which reflects the effect from solid solution strengthening on the strength of the unit cell. For Ti-Zr-15Cu-10Ni wt% filler metal, it kept constant as 0.3476 with Zr as 37.5 ∝ 45 wt% and decreased to 0.333 with Zr decreasing from 37.5 to 25 wt%. Finally, it increased up to 0.3406 with Zr as 2 ∝ 10 wt%. Thus, Ti-based filler metal with Zr content being 2 ∝ 10 wt% is suggested based on the simulation results. Moreover, the calculated covalent electron of n{sub A}{sup u-v} showed good agreement with the hardness of the joint by filler 37.5Zr and 10Zr. The composition of Ti-10Zr-15Cu-10Ni wt% was verified in this study with higher tensile strength of the brazing joint and uniform microstructure of the interface. (orig.)

  16. Contribution to the study of U-Ti and U-Pu-Ti carbides

    International Nuclear Information System (INIS)

    Milet, C.A.

    1968-01-01

    After having discussed the reasons to use (U,Pu) carbides as fast reactor fuel, we examine the influence of the addition of titanium to these carbides. A preliminary study has been done on the system of U-C-Ti and some properties have been measured such as: density, thermal expansion, electrical resistivity, atmospheric corrosion and compatibility with stainless steel. The systems U-Pu-C-Ti (Pu/U + Pu equal to 15 per cent) and U-C-Ti have been found to be very similar. There exists a two phases region (U,Pu)C + TiC, an eutectic between (U,Pu)C and TiC for approximately 15 at %. The solubilities of U + Pu in TiC and of Ti in (U,Pu)C is less than 1 % at. The addition of titanium does not markedly change thermal expansion coefficients of (U,Pu)C. However the resistance to atmospheric corrosion and compatibility with stainless steel is improved. Thermal conductivity, calculated from electrical resistivity, has increased. On the other side, the density of fissile material is lowered. The combination of (U,Pu)C + TiC seems to be the most promising alloy for application as nuclear fuel. (author) [fr

  17. Progress in Understanding of Fuel-Cladding Chemical interaction in Metal Fuel

    International Nuclear Information System (INIS)

    Inagaki, Okenta; Nakamura, Kinya; Ogata, Takanari

    2013-01-01

    Conclusion: Representative phases formed in FCCI were identified: • The reaction between lanthanide elements and cladding; • The reaction between U-PU-Zr and cladding (Fe). Characteristics of the wastage layer were clarified: • Time and temperature dependency of the growth ratio of the wastage layer formed by lanthanide elements; • Threshold temperature of the liquid phase formation in the reaction between U-Pu-Zr and Fe. These results are used: - as a basis for the FCCI modeling; - as a reference data in post-irradiation examination of irradiated metallic fuels

  18. Structure and physical properties of ternary uranium transition-metal antimonides U3MSb5 (M = Zr, Hf, Nb)

    International Nuclear Information System (INIS)

    Tkachuk, Andriy V.; Muirhead, Craig P.T.; Mar, Arthur

    2006-01-01

    The ternary uranium transition-metal antimonides U 3 MSb 5 (M = Zr, Hf, Nb) were prepared by arc-melting reactions followed by annealing at 800 deg. C, or by use of a Sn flux. These compounds extend the previously known series U 3 MSb 5 (M = Ti, V, Cr, Mn) and RE 3 MSb 5 (RE = La, Ce, Pr, Nd, Sm; M = Ti, Zr, Hf, Nb). The crystal structures of U 3 MSb 5 were determined by single-crystal X-ray diffraction data (Pearson symbol hP18, hexagonal, space group P6 3 /mcm, Z = 2; U 3 ZrSb 5 , a = 9.2223(3) A, c = 6.1690(2) A; U 3 HfSb 5 , a = 9.2084(4) A, c = 6.1629(3) A; U 3 NbSb 5 , a = 9.1378(4) A, c 6.0909(6) A). U 3 TaSb 5 has also been identified in microcrystalline form (a = 9.233(3) A, c = 6.142(3) A). Four-probe electrical resistivity measurements on single crystals and dc magnetic susceptibility measurements on powders indicated prominent transitions that are attributed to ferromagnetic ordering. The Curie temperatures, T C , located from ac magnetic susceptibility curves, are 135 K for U 3 ZrSb 5 , 141 K for U 3 HfSb 5 , and 107 K for U 3 NbSb 5

  19. Study of U - Pu - Fe alloys (Masurca critical experiment); Etudes d'alliages U-Pu-Fe (experience critique MASURCA)

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    Three compositions have been studied: 73.5 U - 25 Pu - 1.5 Fe (weight %) 74 U - 25 Pu - 1 Fe 74.5 U - 25 Pu - 0.5 Fe Elaboration and Casting are easy. After two weeks in air 74.5 U - 25 Pu - 0.5 Fe alloys are reduced in powder. As-cast alloys containing 1 and 1,5% Fe are kept undamaged during several months. A rapid oxidisation of the alloys is however observed when the samples undergo the phase transformation (at 595 deg. C and 590 deg. C respectively). Ignition tests in the presence of air show that the oxidisation starts at about 250 deg. C and that the reaction does not spread. Ignition is not observed during heating from 20 to 660 deg. C. The transformation temperature, the melting temperature and the thermal expansion coefficients have been determined by dilatometry. Below the transformation temperature, the principal phases are U-Pu zeta and (U, Pu){sub 6}Fe. Thermal conductibility, Young modulus, density and heat of fusion have been measured. Compatibility tests show that between U-Pu-Fe and stainless steel a phase of (U, Pu){sub 6}Fe type is formed. The 74 U - 25 Pu - 1% Fe alloy seems to behave better than 73.5 U - 25 Pu - 1.5% Fe alloy because the (U, Pu){sub 6}Fe layer is two or three times smaller. Finally, the thermal stability has been studied with the 74 U - 25 Pu - 1% Fe alloy. A dilatometric anomaly (very weak expansion) occurs when the sample is heated above transformation temperature and cooled. But there is no anomaly by thermal cycling from 50 deg. C to 400 deg. C and there is no deterioration of alloys by heat treatments at 100 deg. C, 200 deg. C, 300 deg. C during 5 months under vacuum. (authors) [French] Trois compositions ont ete etudiees: 73,5 U - 25 Pu - 1,5 Fe (% ponderaux) 74 U - 25 Pu - 1 Fe 74,5 U - 25 Pu - 0,5 Fe Les elaborations et mises en forme sont-faciles. Les alliages a 0,5% de fer tombent en poudre apres 15 jours d'exposition a l'air. Les alliages a 1 et 1,5% de fer, bruts de coulee, se conservent plusieurs mois. Cependant

  20. Self-irradiation effect on thermal conductivity of Zr{sub 0.70}Pu{sub 0.25}Cm{sub 0.05}N solid solution

    Energy Technology Data Exchange (ETDEWEB)

    Nishi, Tsuyoshi, E-mail: tsuyoshi.nishi.75@vc.ibaraki.ac.jp [Department of Materials Science and Engineering, College of Engineering, Ibaraki University, Hitachi, Ibaraki, 316-8511 (Japan); Hayashi, Hirokazu; Sato, Takumi; Takano, Masahide [Nuclear Science and Engineering Center, Japan Atomic Energy Agency, Tokai-mura, Ibaraki, 319-1195 (Japan)

    2017-04-01

    This study evaluated the dependence of the thermal conductivity of Zr{sub 0.70}Pu{sub 0.25}Cm{sub 0.05}N on storage time and temperature. The authors prepared sintered samples of Zr{sub 0.70}Pu{sub 0.25}Cm{sub 0.05}N solid solution and measured thermal diffusivity at storage times of 0, 24, 72, 144, 240, 408, 552, 816, 1,680, and 2064 h, from which it was determined that the thermal conductivity decreased exponentially with increasing storage time. This result suggests that the decrease of the thermal conductivity could be attributed to the accumulation of lattice defects from self-irradiation. To confirm the thermal recovery behavior of Zr{sub 0.70}Pu{sub 0.25}Cm{sub 0.05}N under annealing, thermal diffusivity was also measured just after annealing. The thermal conductivity of Zr{sub 0.70}Pu{sub 0.25}Cm{sub 0.05}N was determined to be larger than that of Zr{sub 0.58}Pu{sub 0.21}Cm{sub 0.21}N but smaller than that of Zr{sub 0.80}Pu{sub 0.10}Cm{sub 0.10}N. - Highlights: •The authors prepared Zr{sub 0.70}Pu{sub 0.25}Cm{sub 0.05}N and measured thermal diffusivity. •Dependence of thermal conductivity on storage time was clarified. •After annealing at 1423 K in vacuum, thermal conductivity was recovered.

  1. Plutonium diffusion in advanced fuels (U,Pu)(C,O) and (U,Pu)(C,N)

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Matzke, H.

    1983-01-01

    The self-diffusion of 238 Pu was measured in an oxicarbide (U,Pu)(C,O) and a carbonitride (U,Pu) (C,N). The activation enthalpies were 447 and 347 kJ mol -1 , respectively. The carbonitrides were confirmed to fall into three classes: carbide-like compositions with less than 30% nitrogen in the metalloid lattice, nitride-like composition with more than 70% nitrogen and with reduced atomic mobilities, and carbonitrides with about 50% nitrogen showing an intermediate behavior. The oxicarbide showed diffusion coefficients slightly larger than those of pure carbides

  2. Development of an U and Pu recovery process by molten salt electrorefining. Behavior of U and Pu at simultaneous recoveries into liquid cadmium cathodes

    International Nuclear Information System (INIS)

    Uozumi, Koichi; Kato, Tetsuya; Iizuka, Masatoshi; Inoue, Tadashi; Iwai, Takashi; Shirai, Osamu; Arai, Yasuo

    2003-01-01

    In order to study behaviors of U and Pu at simultaneous recoveries into liquid cadmium cathodes (LCCs) in the electrorefining of pyrometallurgical reprocessing, several experiments were conducted to recover U and Pu into LCCs at different U/Pu ratios in the salt phase. The major results were as follows: (1) The weight ratios of U and Pu in 120 g LCCs reached 10 wt.% (the tentative target), with current efficiencies higher than 80 %. (2) Under the conditions of U/Pu ratios in the initial salt phase less than 1/4.3, the amounts of recovered U and Pu were proportional to the passed electric charges, with the separation factors of U to Pu (= (U/Pu ration in the recovered product)/(U/Pu ratio in the salt)) between 1.2 and 2.0. (3) On the other hand, under the condition of U/Pu ratio in the initial salt phase at 1/1.73, only U was recovered into the LCC after the saturation of LCC with U and Pu. Accordingly, there will be a threshold in the U/Pu ratio of the salt phase for the simultaneous recovery of U and Pu. (4) Am showed a similar behavior to Pu. The separation factors of Am to Pu (=(AM/Pu ratio in the recovered product)/(Am/Pu ratio in the salt)) was 0.78, which means that Am is co-recovered with Pu into LCC. (author)

  3. ZZ WPPR-FR-MOX/BNCMK, Benchmark on Pu Burner Fast Reactor

    International Nuclear Information System (INIS)

    Garnier, J.C.; Ikegami, T.

    1993-01-01

    Description of program or function: In order to intercompare the characteristics of the different reactors considered for Pu recycling, in terms of neutron economy, minor actinide production, uranium content versus Pu burning, the NSC Working Party on Physics of Plutonium Recycling (WPPR) is setting up several benchmark studies. They cover in particular the case of the evolution of the Pu quality and Pu fissile content for Pu recycling in PWRs; the void coefficient in PWRs partly fuelled with MOX versus Pu content; the physics characteristics of non-standard fast reactors with breeding ratios around 0.5. The following benchmarks are considered here: - Fast reactors: Pu Burner MOX fuel, Pu Burner metal fuel; - PWRs: MOX recycling (bad quality Pu), Multiple MOX recycling

  4. Study of a ;hot; particle with a matrix of U-bearing metallic Zr: Clue to supercriticality during the Chernobyl nuclear accident

    Science.gov (United States)

    Pöml, P.; Burakov, B.

    2017-05-01

    This paper is dedicated to the 30th anniversary of the severe nuclear accident that occurred at the Chernobyl NPP on 26 April 1986. A detailed study on a Chernobyl "hot" particle collected from contaminated soil was performed. Optical and electron microscopy, as well as quantitative x-ray microbeam analysis methods were used to determine the properties of the sample. The results show that the particle (≈ 240 x 165 μm) consists of a metallic Zr matrix containing 2-3 wt. % U and bearing veins of an U,Nb admixture. The metallic Zr matrix contains two phases with different amounts of O with the atomic proportions (U,Zr,Nb)0.73O0.27 and (U,Zr,Nb)0.61O0.39. The results confirm the interaction between UO2 fuel and zircaloy cladding in the reactor core. To explain the process of formation of the particle, its properties are compared to laboratory experiments. Because of the metallic nature of the particle it is concluded that it must have formed during a very high temperature (> 2400∘C) process that lasted for only a very short time (few microseconds or less); otherwise the particle should have been oxidised. Such a rapid very high temperature process indicates that at least part of the reactor core could have been supercritical prior to an explosion as it was previously suggested in the literature.

  5. Irradiation performance of metallic fuels

    International Nuclear Information System (INIS)

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

    1989-01-01

    Argonne National Laboratory has been working for the past five years to develop and demonstrate the Integral Fast Reactor (IFR) concept. The concept involves a closed system for fast-reactor power generation and on-site fuel reprocessing, both designed specifically around the use of metallic fuel. The Experimental Breeder Reactor-II (EBR-II) has used metallic fuel for all of its 25-year life. In 1985, tests were begun to examine the irradiation performance of advanced-design metallic fuel systems based on U-Zr or U-Pu-Zr fuels. These tests have demonstrated the viable performance of these fuel systems to high burnup. The initial testing program will be described in this paper. 2 figs

  6. KORELASI KOMPOSISI UNSUR TERHADAP SIFAT TERMAL SERBUK BAHAN BAKAR U-ZrHX

    Directory of Open Access Journals (Sweden)

    Masrukan Masrukan

    2016-10-01

    yaitu kapasitas panas diperoleh nilai kapasitas panas tertinggi pada serbuk U-35ZrHx, sedangkan dari pengujian transisi perubahan fasa diperoleh bahwa pada U-45ZrHx mengalami dua tahapan reaksi disertai perubahan fasa. Terdapat pengaruh komposisi terhadap sifat termalnya, dimana semakin tinggi kandungan Zr maka nilai kapaistas panas hidrida uranium zirkonium semakin rendah. Kata kunci: komposisi, sifat termal, bahan bakar, U-ZrHx. ABSTRACT COMPOSITION CORRECTION ON THE THERMAL PROPERTIES OF U-ZrHX FUEL POWDERS. Analysis has been conducted to determine the composition correlation on the thermal properties of the powder fuel U-ZrHx. U-ZrHx powder made from the process hidriding U-Zr ingot, where the ingot is the result of U-Zr and Zr U metal melting. In this experiment made three variations of powders, namely U-35ZrHx, U-45ZrHx, and U-55ZrHx. Need for determination of the thermal properties of Zr was to determine the effect of the nature of the transformation of Zr levels of heat from the fuel. At first, U and Zr metal is melted in electric arc furnaces to produce ingot U-Zr. U-Zr ingot then made powder with hidridring-milling techniques to produce U-Zr powder. U-Zr powder composition analyzed using techniques sepectroscopy atomic absorption (AAS and UV-Vis spectroscopy. The results of composition analysis showed that the analysis to determine the content of U and Zr nearly all the test samples analyzed have quite a big difference between the content of U and Zr as determined by the results of the analysis of U and Zr exception analysis result in powder U-45Zr which differ only 0.609%. From the analysis of impurities obtained that nearly all the impurities that exist still meet the requirements for fuel unless the elements Fe, where elements of the existing Fe amounted to 382.912 g/g while the requirement of £ 250 mg /µg. Testing conducted heat capacity in the temperature range 35 ° C to 437 ° C showed that the capacity were greatest powder 35ZrHx U-with a value heat

  7. A COMPARISON OF (PuU)C AND (PuU)O$sub 2$ FUELED FAST BREEDER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Kazi, A. Halim; Rosen, Stanley

    1963-11-15

    BS>A conceptual evaluation is presented of a (PuU)C 500Mw(e) reactor which uses a moderate power generation of 40 kw/ft and a maximum fuel temperature of 2900 deg F, well below the 4500 deg F melting point of Pu/sub 0.2/U/sub 0.8/C. The principal design characteristics and fuel cycle costs are tabulated together with those of a (PuU)O/sub 2/ design of equal power. The comparison shows that the carbide core has significant economic and safety advantages over the oxide core. (D.L.C.)

  8. Evaluation of fission cross sections and covariances for {sup 233}U, {sup 235}U, {sup 238}U, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu

    Energy Technology Data Exchange (ETDEWEB)

    Kawano, Toshihiko [Kyushu Univ., Fukuoka (Japan); Matsunobu, Hiroyuki [Data Engineering, Inc. (Japan); Murata, Toru [AITEL Corporation, Tokyo (JP)] [and others

    2000-02-01

    A simultaneous evaluation code SOK (Simultaneous evaluation on KALMAN) has been developed, which is a least-squares fitting program to absolute and relative measurements. The SOK code was employed to evaluate the fission cross sections of {sup 233}U, {sup 235}U, {sup 238}U, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu for the evaluated nuclear data library JENDL-3.3. Procedures of the simultaneous evaluation and the experimental database of the fission cross sections are described. The fission cross sections obtained were compared with evaluated values given in JENDL-3.2 and ENDF/B-VI. (author)

  9. Study of U - Pu - Fe alloys (Masurca critical experiment)

    International Nuclear Information System (INIS)

    Barthelemy, P.; Boucher, R.

    1965-01-01

    Three compositions have been studied: 73.5 U - 25 Pu - 1.5 Fe (weight %) 74 U - 25 Pu - 1 Fe 74.5 U - 25 Pu - 0.5 Fe Elaboration and Casting are easy. After two weeks in air 74.5 U - 25 Pu - 0.5 Fe alloys are reduced in powder. As-cast alloys containing 1 and 1,5% Fe are kept undamaged during several months. A rapid oxidisation of the alloys is however observed when the samples undergo the phase transformation (at 595 deg. C and 590 deg. C respectively). Ignition tests in the presence of air show that the oxidisation starts at about 250 deg. C and that the reaction does not spread. Ignition is not observed during heating from 20 to 660 deg. C. The transformation temperature, the melting temperature and the thermal expansion coefficients have been determined by dilatometry. Below the transformation temperature, the principal phases are U-Pu zeta and (U, Pu) 6 Fe. Thermal conductibility, Young modulus, density and heat of fusion have been measured. Compatibility tests show that between U-Pu-Fe and stainless steel a phase of (U, Pu) 6 Fe type is formed. The 74 U - 25 Pu - 1% Fe alloy seems to behave better than 73.5 U - 25 Pu - 1.5% Fe alloy because the (U, Pu) 6 Fe layer is two or three times smaller. Finally, the thermal stability has been studied with the 74 U - 25 Pu - 1% Fe alloy. A dilatometric anomaly (very weak expansion) occurs when the sample is heated above transformation temperature and cooled. But there is no anomaly by thermal cycling from 50 deg. C to 400 deg. C and there is no deterioration of alloys by heat treatments at 100 deg. C, 200 deg. C, 300 deg. C during 5 months under vacuum. (authors) [fr

  10. Development and testing of metallic fuels with high minor actinide content

    International Nuclear Information System (INIS)

    Meyer, M.K.; Hayes, S.L.; Kennedy, J.R.; Keiser, D.D.; Hilton, B.A.; Frank, S.M.; Kim, Y.-S.; Chang, G.; Ambrosek, R.G.

    2003-01-01

    Metallic alloys are promising candidates for use as fuels for transmutation and in advanced closed nuclear cycles. Metallic alloys have high heavy metal atom density, relatively high thermal conductivity, favorable gas release behavior, and lend themselves to remote recycle processes. Both non-fertile and uranium-bearing metal fuels containing minor actinide are under consideration for use as transmutation fuels by the U.S. Advanced Fuel Cycle (AFC) program, however, little irradiation performance data exists for fuel forms containing significant fractions of minor actinides. The first irradiation tests of non-fertile high-actinide-content fuels are scheduled to begin in early 2003 in the Advanced Test Reactor (ATR). The irradiation test matrix was designed to provide basic information on the irradiation behavior of binary Pu-Zr alloy fuel and the effect of the minor actinides americium and neptunium on alloy fuel behavior, together and separately. Five variants of transuranic containing zirconium-based alloy fuels are included in the AFC-1 irradiation test matrix. These are (in wt.%) Pu-40Zr, Pu-60Zr, Pu-12Am-40Zr, Pu-10Np-40Zr and Pu-10Np-10Am-40Zr. PuN-ZrN based fuels containing Am and Np are also included. All five of the fuel alloys have been fabricated in the form of cylindrical fuel slugs by arc-casting. Short melt times, on the order or 5-20 seconds, prevent the volatilization of significant quantities of americium metal, despite the high melt temperatures characteristic of the arc-melting process. Alloy microstructure have been characterized by x-ray diffraction and scanning electron microscopy. Thermal analysis has also been performed. The AFC-1 irradiation experiment configuration consists of twenty-four sodium bonded fuel specimens sealed in helium filled secondary capsules. The first capsule has a design burnup to 7 at.% 239 Pu; goal peak burnup of the second capsule is ∼18 at%. Capsule assemblies are placed within an aluminum flow-through basket

  11. Study of new U(VI) and Pu(VI) coprecipitation methods for the preparation of (U,Pu)O2

    International Nuclear Information System (INIS)

    Sanoit, J. de.

    1990-01-01

    Two U(VI) and Pu(VI) coprecipitation methods have been studied, for the definition of new processes to prepare (U,Pu)O 2 mixed oxides suitable for making MoX fuels or fast breeder reactor fuels. The first system is based on the coprecipitation of a new U(VI), Pu(VI) compound; ammonium uranoplutonate, where as a second system is related to the precipitation of uranyl plutonyl monocarbonate. Experimental conditions to optimize the precipitation and the filtration steps of these two systems have been determined. After calcination under reducing conditions, the mixed oxides obtained are characterized according to different techniques: granulometry, thermogravimetry, solubility in boiling HNO 3 solutions. The properties of such oxides are excellent. The possible processes for preparing (U, Pu)O 2 using these new routes are compared with those actually exploited [fr

  12. U-Zr alloy: XPS and TEM study of surface passivation

    Science.gov (United States)

    Paukov, M.; Tkach, I.; Huber, F.; Gouder, T.; Cieslar, M.; Drozdenko, D.; Minarik, P.; Havela, L.

    2018-05-01

    Surface reactivity of Uranium metal is an important factor limiting its practical applications. Bcc alloys of U with various transition metals are much less reactive than pure Uranium. So as to specify the mechanism of surface protection, we have been studying the U-20 at.% Zr alloy by photoelectron spectroscopy and transmission electron microscopy. The surface was studied in as-obtained state, in various stages of surface cleaning, and during an isochronal annealing cycle. The analysis based on U-4f, Zr-3p, and O-1 s spectra shows that a Zr-rich phase segregates at the surface at temperatures exceeding 550 K, which provides a self-assembled coating. The comparison of oxygen exposure of the stoichiometric and coated surfaces shows that the coating is efficiently preventing the oxidation of uranium even at elevated temperatures. The coating can be associated with the UZr2+x phase. TEM study indicated that the coating is about 20 nm thick. For the clean state, the U-4f core-level lines of the bcc alloy are practically identical to those of α-U, revealing similar delocalization of the 5f electronic states.

  13. Reusable molds for casting U-Zr alloys

    International Nuclear Information System (INIS)

    Chen, P.S.; Stevens, W.C.; Trybus, C.L.

    1992-09-01

    Refractory oxides, carbides, nitrides and sulfides were examined as mold coating materials for use in casting nuclear fuel. The molds require excellent high temperature chemical and mechanical stability combined with reasonable room temperature ductility to allow for fuel removal. Coatings were applied onto quartz and refractory metal coupons using various techniques. Sessile drop tests employing molten U-10%Zr (by weight) at 1550 degrees C were used to characterize coating performance. Results indicate that NbC, TiN, and Y 2 O 3 were non-wetting with U-10%Zr. However, only the Y 2 O 3 coating completely prevented adhesion of the fuel. The paper describes coating methods and details of the sessile drop experiments

  14. Validation of MCNP6.1 for Criticality Safety of Pu-Metal, -Solution, and -Oxide Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kiedrowski, Brian C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favorite, Jeffrey A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kahler, III, Albert C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kersting, Alyssa R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parsons, Donald K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Walker, Jessie L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-05-13

    Guidance is offered to the Los Alamos National Laboratory Nuclear Criticality Safety division towards developing an Upper Subcritical Limit (USL) for MCNP6.1 calculations with ENDF/B-VII.1 nuclear data for three classes of problems: Pu-metal, -solution, and -oxide systems. A benchmark suite containing 1,086 benchmarks is prepared, and a sensitivity/uncertainty (S/U) method with a generalized linear least squares (GLLS) data adjustment is used to reject outliers, bringing the total to 959 usable benchmarks. For each class of problem, S/U methods are used to select relevant experimental benchmarks, and the calculational margin is computed using extreme value theory. A portion of the margin of sub criticality is defined considering both a detection limit for errors in codes and data and uncertainty/variability in the nuclear data library. The latter employs S/U methods with a GLLS data adjustment to find representative nuclear data covariances constrained by integral experiments, which are then used to compute uncertainties in keff from nuclear data. The USLs for the classes of problems are as follows: Pu metal, 0.980; Pu solutions, 0.973; dry Pu oxides, 0.978; dilute Pu oxide-water mixes, 0.970; and intermediate-spectrum Pu oxide-water mixes, 0.953.

  15. Measurement of oxygen chemical diffusion in PuO2-x and analysis of oxygen diffusion in PuO2-x and (Pu,U)O2-x

    International Nuclear Information System (INIS)

    Kato, Masato; Uchida, Teppei; Sunaoshi, Takeo

    2013-01-01

    Oxygen chemical diffusion in PuO 2-x was investigated in the temperature range of 1473-1873 K by thermogravimetry as functions of oxygen-to-metal (O/M) ratios and temperatures. The oxygen chemical diffusion coefficients, D were determined assuming that the reduction curves were dominated by a diffusion process. The O/M ratio and Pu content dependence on the chemical diffusion coefficients were evaluated. The chemical diffusion coefficient had its minimum value at around O/M=1.98 and decreased with increasing Pu content in (U,Pu)O 2-x . The self-diffusion coefficients were evaluated. A model for describing the relationship among O/M ratio, oxygen chemical diffusion, and self-diffusion was proposed based on defect chemistry. (copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  16. A 233U/236U/242Pu/244Pu spike for isotopic and isotope dilution analysis by mass spectrometry with internal calibration

    International Nuclear Information System (INIS)

    Stepanov, A.; Belyaev, B.; Buljanitsa, L.

    1989-11-01

    The Khlopin Radium Institute prepared on behalf of the IAEA a synthetic mixture of 233 U, 236 U, 242 Pu and 244 Pu isotopes. The isotopic composition and elemental concentration of uranium and plutonium were certified on the basis of analyses done by four laboratories of the IAEA Network, using mass spectrometry with internal standardization. The certified values for 233 U/ 236 U ratio and the 236 U chemical concentration have a coefficient of variation of 0.05%. The latter is fixed by the uncertainty in the 235 U/ 238 U ratio of NBS500 used as internal standard. The coefficients of variation of the 244 Pu/ 242 Pu ratio and the 242 Pu chemical concentration are respectively 0.10% and 0.16% and limited by the uncertainty in the 240 Pu/ 239 Pu ratio of NBS947. This four isotope mixture was used as an internal standard as well as a spike, to analyze 30 batches of LWR spent fuel solutions. The repeatability of the mass spectrometric measurements have a coefficient of variation of 0.025% for the uranium concentration, and of 0.039% for the plutonium concentration. The spiking and treatment errors had a coefficient of variation of 0.048%. (author). Refs, figs and tabs

  17. Direct dissolution of g-level U metal and U-6 % Zr alloy bits by TBP-nitric acid adduct and in situ extraction at ambient pressures

    International Nuclear Information System (INIS)

    Shekhar Kumar; Bijendra Kumar; Gelatar, J.K.; Pranay Kumar Sinha; Alok Kumar Mishra; Kamachi Mudali, U.

    2016-01-01

    A study on direct dissolution of g-level metallic U and U-6 % Zr alloy bits by TBP-nitric acid adduct followed by in situ extraction at ambient pressures under batch and dynamic conditions was performed. The product organic solution was adjusted to 30 % TBP in dodecane and from it, U(VI) could be stripped quantitatively with 0.01 N nitric acid. Experimental results of the study are presented in this paper. (author)

  18. Contribution to the study of the (U,Pu)C,N system; Contribution a l'etude du systeme (U,Pu)C,N

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-07-01

    The reactions of UC, PuC, (U,Pu)C, UC{sub 2} and U(C{sub 1-x}O{sub x}) with nitrogen at moderate temperatures (room temperature to 400 C) are described. The influence of the uptake of nitrogen by the powders necessary to sinter the carbides upon the nature of the final product has been investigated; it has been shown that the sintered carbides are hyper-stoichiometric. The reactions of carbon with UN, PuN and (U,Pu)N has also been studied. Under vacuum, carbon reacts on the nitrides at temperatures as low as 1100 C; nitrogen is replaced by carbon and the final product is a carbonitride. The reaction is: MN + x C {yields} MN{sub 1-x}C{sub x} + x/2N{sub 2}. The reaction is limited and the carbonitrides have a fixed composition in presence of M{sub 2}C{sub 3} or MC{sub 2}; hence it is impossible to produce pure MC using the reaction. The ternary diagram U-C-N, Pu-C-N and (U,Pu)C-N have been drawn. They show clearly that it is possible to obtain single phase carbonitrides in a wide domain of compositions. (author) [French] On decrit les reactions avec l'azote de UC, PuC,(U,Pu)C,UC{sub 2} et U(C{sub 1-x}O{sub x}), par action directe de l'azote a temperature moderee (de l'ambiante a 450 C). On a etudie l'influence de la contamination par l'azote des poudres de carbures necessaires au frittage sur la nature des produits frittes; on a montre que les carbures frittes obtenus sont hyperstoechiometriques. On a etudie parallelement les reactions du carbone avec UN, PuN et (U,Pu)N. Sous vide le carbone reagit sur les nitrures des 1100 C: le carbone se substitue a l'azote; l'azote libere est elimine et le produit final est un carbonitrure. La reaction s'ecrit: MN + x C {yields} MN{sub 1-x}C{sub x} + x/2N{sub 2}. La reaction est limitee et les carbonitrures obtenus ont une composition limite fixe en presence des carbures superieurs M{sub 2}C{sub 3} et MC{sub 2}; il est donc impossible d'obtenir MC pur par cette reaction. Les diagrammes d'equilibre U-C-N, Pu-C-N et (U,Pu) C-N, ont

  19. Mixed PWR core loadings with inert matrix Pu-fuel assemblies

    International Nuclear Information System (INIS)

    Stanculescu, A.; Kasemeyer, U.; Paratte, J.-M.; Chawla, R.

    1999-01-01

    The most efficient way to enhance plutonium consumption in light water reactors is to eliminate the production of plutonium all together. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. At PSI, studies have focused on employing ZrO 2 as inert matrix. Adding a burnable poison to such a fuel proves to be necessary. As a result of scoping studies, Er 2 O 3 was identified as the most suitable burnable poison material. The results of whole-core three-dimensional neutronics analyses indicated, for a present-day 1000 MW e pressurised water reactor, the feasibility of an asymptotic equilibrium four-batch cycle fuelled solely with the proposed PuO 2 -Er 2 O 3 -ZrO 2 inert matrix fuel (IMF). The present paper presents the results of more recent investigations related to 'real-life' situations, which call for transition configurations in which mixed IMF and UO 2 assembly loadings must be considered. To determine the influence of the introduction of IMF assemblies on the characteristics of a UO 2 -fuelled core, three-dimensional full-core calculations have been performed for a present-day 1000 MW e PWR containing up to 12 optimised IMF assemblies. (author)

  20. Consequence analysis of core meltdown accidents in liquid metal fast reactor

    International Nuclear Information System (INIS)

    Suk, S.D.; Hahn, D.

    2001-01-01

    Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of work to demonstrate the inherent and ultimate safety of the conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 Mw pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method was developed using a modified Bethe-Tait method to simulate the kinetics and hydraulic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the method for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies was also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters. (author)

  1. Contribution to the study of the (U,Pu)C,N system; Contribution a l'etude du systeme (U,Pu)C,N

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-07-01

    The reactions of UC, PuC, (U,Pu)C, UC{sub 2} and U(C{sub 1-x}O{sub x}) with nitrogen at moderate temperatures (room temperature to 400 C) are described. The influence of the uptake of nitrogen by the powders necessary to sinter the carbides upon the nature of the final product has been investigated; it has been shown that the sintered carbides are hyper-stoichiometric. The reactions of carbon with UN, PuN and (U,Pu)N has also been studied. Under vacuum, carbon reacts on the nitrides at temperatures as low as 1100 C; nitrogen is replaced by carbon and the final product is a carbonitride. The reaction is: MN + x C {yields} MN{sub 1-x}C{sub x} + x/2N{sub 2}. The reaction is limited and the carbonitrides have a fixed composition in presence of M{sub 2}C{sub 3} or MC{sub 2}; hence it is impossible to produce pure MC using the reaction. The ternary diagram U-C-N, Pu-C-N and (U,Pu)C-N have been drawn. They show clearly that it is possible to obtain single phase carbonitrides in a wide domain of compositions. (author) [French] On decrit les reactions avec l'azote de UC, PuC,(U,Pu)C,UC{sub 2} et U(C{sub 1-x}O{sub x}), par action directe de l'azote a temperature moderee (de l'ambiante a 450 C). On a etudie l'influence de la contamination par l'azote des poudres de carbures necessaires au frittage sur la nature des produits frittes; on a montre que les carbures frittes obtenus sont hyperstoechiometriques. On a etudie parallelement les reactions du carbone avec UN, PuN et (U,Pu)N. Sous vide le carbone reagit sur les nitrures des 1100 C: le carbone se substitue a l'azote; l'azote libere est elimine et le produit final est un carbonitrure. La reaction s'ecrit: MN + x C {yields} MN{sub 1-x}C{sub x} + x/2N{sub 2}. La reaction est limitee et les carbonitrures obtenus ont une composition limite fixe en presence des carbures superieurs M{sub 2}C{sub 3} et MC{sub 2}; il est donc impossible d'obtenir MC pur par cette reaction. Les diagrammes

  2. Casting of metallic fuel containing minor actinide additions

    International Nuclear Information System (INIS)

    Trybus, C.L.; Henslee, S.P.; Sanecki, J.E.

    1992-01-01

    A significant attribute of the Integral Fast Reactor (IFR) concept is the transmutation of long-lived minor actinide fission products. These isotopes require isolation for thousands of years, and if they could be removed from the waste, disposal problems would be reduced. The IFR utilizes pyroprocessing of metallic fuel to separate auranium, plutonium, and the minor actinides from nonfissionable constituents. These materials are reintroduced into the fuel and reirradiated. Spent IFR fuel is expected to contain low levels of americium, neptunium, and curium because the hard neutron spectrum should transmute these isotopes as they are produced. This opens the possibility of using an IFR to trnasmute minor actinide waste from conventional light water reactors (LWRs). A standard IFR fuel is based on the alloy U-20% Pu-10% Zr (in weight percent). A metallic fuel system eases the requirements for reprocessing methods and enables the minor actinide metals to be incorporated into the fuel with simple modifications to the basic fuel casting process. In this paper, the authors report the initial casting experience with minor actinide element addition to an IFR U-Pu-Zr metallic fuel

  3. Investigation on U - O - Na, Pu - O - Na and U,Pu - O - Na phase diagrams

    International Nuclear Information System (INIS)

    Pillon, S.

    1989-03-01

    The thermochemical interaction between the nuclear fuel (uranium and plutonium mixed oxides) and the sodium has been investigated and particularly the three phase diagrams: U - O - Na; Pu - O - Na; U,Pu - O - Na. High temperature neutron diffraction, microcalorimetry and powder X-ray diffraction were used for the characterization of the compounds synthetized. This study allowed to complete the knowledge about each of these diagrams and to measure some physical and thermal properties on the compounds. The limits on the modelization of the fuel-sodium interaction are discussed from the results of the UO 2 - Na reaction [fr

  4. Development of fast reactor metal fuels containing minor actinides

    International Nuclear Information System (INIS)

    Ohta, Hirokazu; Ogata, Takanari; Kurata, Masaki; Koyama, Tadafumi; Papaioannou, Dimitrios; Glatz, Jean-Paul; Rondinella, Vincenzo V.

    2011-01-01

    Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phenix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU. (author)

  5. Benchmark physics tests in the metallic-fuelled assembly ZPPR-15

    International Nuclear Information System (INIS)

    McFarlane, H.F.; Brumbach, S.B.; Carpenter, S.G.; Collins, P.J.

    1987-01-01

    In the last two years a shift in emphasis to inherent safety and economic competitiveness has led to a resurgence in US interest in metallic-alloy fuels for LMRs. Argonne National Laboratory initiated an extensive testing program for metallic-fuelled LMR technology that has included benchmark physics as one component. The tests done in the ZPPR-15 Program produced the first physics results in over 20 years for a metal-composition LMR core

  6. Measurement of oxygen chemical diffusion in PuO{sub 2-x} and analysis of oxygen diffusion in PuO{sub 2-x} and (Pu,U)O{sub 2-x}

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Masato; Uchida, Teppei [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Ibaraki 319-1194 (Japan); Sunaoshi, Takeo [Inspection Development Company Ltd., 4-33 Muramatsu, Tokai-mura, Ibaraki 319-1194 (Japan)

    2013-02-15

    Oxygen chemical diffusion in PuO{sub 2-x} was investigated in the temperature range of 1473-1873 K by thermogravimetry as functions of oxygen-to-metal (O/M) ratios and temperatures. The oxygen chemical diffusion coefficients, D were determined assuming that the reduction curves were dominated by a diffusion process. The O/M ratio and Pu content dependence on the chemical diffusion coefficients were evaluated. The chemical diffusion coefficient had its minimum value at around O/M=1.98 and decreased with increasing Pu content in (U,Pu)O{sub 2-x}. The self-diffusion coefficients were evaluated. A model for describing the relationship among O/M ratio, oxygen chemical diffusion, and self-diffusion was proposed based on defect chemistry. (copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  7. Determination of Pu and U by ICP-MS

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Y.; Yoshida, S. [National Inst. of Radiological Sciences, Chiba (Japan)

    2002-06-01

    In order to understand the distribution and behaviour of long-lived radionuclides in the terrestrial environment, we have developed analytical methods for {sup 238}U, {sup 232}Th and {sup 239},{sup 240}Pu and determined them in several environmental samples. ICP-MS was applied to the analyses. Concentrations of {sup 238}U and {sup 232}Th were determined in variety of soils, plants and environmental waters. The mean concentration of U and Th in 77 soils, collected several parts of Japan, were 2.3 and 7.0 ppm (dry wt). Red and yellow soils, which are mainly found in western Japan, have high concentrations of Th, suggesting the contribution from granitic rocks. Higher concentration of U in agricultural soils than forest soils suggested the possible accumulation of this element from phosphate fertilizer. The soil-to-plant transfer factors (concentration ratios between plants and soils) of U and Th were estimated from the analytical results of these nuclides in soils and in several crops. The factors for U were about 0.0005 in leaf vegetables, 0.0002 in root vegetables, 0.0001 in potatoes and 0.00001 in rice (polished). For Th, the factors were about 0.0005 in leaf vegetables, 0.0008 in root vegetables, 0.0005 in potatoes and 0.00003 in rice (polished). For the analysis of plutonium, chromatographic resin, Dowex 1x8, was examined for their applicability in the separation of Pu from the matrix elements. Sufficient decontamination factors (10{sup 4}-10{sup 5}) for many matrix elements including U, which interferes in the detection of mass 239, were obtained in both resins. Analytical results of {sup 239+240}Pu in standard reference materials (IAEA-135, -SOIL-6, -368 and -134) indicated that the accuracy of this method was satisfactory. Data on the {sup 240}Pu/{sup 239}Pu atom ratios in these reference materials, which are rare in the literature, were also obtained. Soil samples from the Marshall Islands showed a large variation of the {sup 240}Pu/{sup 239}Pu atom ratios

  8. Chapter 3: Exponential experiments on graphite-moderated lattices fuelled with near-natural uranium metal rods

    International Nuclear Information System (INIS)

    McCulloch, D.B.; Clarke, W.G.; Ashworth, F.P.O.; Hoskins, T.A.

    1963-01-01

    Exponential experiments have been carried out on graphite lattices fuelled by 1.2 in. diameter uranium metal rods at three near-natural U 235 compositions, 0.6 Co, 1.3 Co and 1.6 Co. The results, together with those already existing from earlier exponential or critical measurements on these and similar natural uranium rods, have been correlated with the theory of Syrett (1961) and also with the modified form of this theory given in Vol.1, Ch. 7. (author)

  9. Direct reduction of 238PuO2 and 239PuO2 to metal

    International Nuclear Information System (INIS)

    Mullins, L.J.; Foxx, C.L.

    1982-02-01

    The process for reducing 700 g 239 PuO 2 to metal is a standard procedure at Los Alamos National Laboratory. This process is based on research for reducing 200 g 238 PuO 2 to metal. This report describes in detail the experiments and development of the 200-g process. The procedure uses calcium metal as the reducing agent in a molten CaCl 2 solvent system. The process to convert impure plutonia to high-purity metal by oxide reduction followed by electrorefining is also described

  10. Ceramic grade (U,Pu)O2 powder fabrication

    International Nuclear Information System (INIS)

    Cristallini, O.A.; Villegas de Maroto, Marina; De Pino, J.I.; Osuna, H.A.

    1980-01-01

    Ceramic grade UO 2 powder was obtained by the homogeneous precipitation method. This procedure was afterwards applied to the fabrication of ceramic grade (U,Pu)O 2 powders, and mixed oxide powders with Pu content ranging from 0.7 to 16% were obtained. The obtainment of mixed ceramic oxides as well as the recuperation of fabrication scraps were developed in three steps: 1)study of the process of homogeneous precipitation of ammonium diuranate (ADU); 2) co-precipitation of ADU/PuO 2 .H 2 O for Pu concentrations of 0.6 and 6.8; 3) the thermal conditioning to mixed oxide (U,Pu)O 2 powders. The experimental procedure involves the following steps: preparation of the PuO 2 (NO 3 ) 4 solution; co-precipitation of the PuO 2 (NO 3 ) 2 solution with an UO 2 (NO 3 ) 2 solution; filtration and drying of the precipitate, thermal treatment and finally, mixing, pressing and sintering of the (U,Pu)O 2 and Nukem UO 2 powder with a 0. of zinc stearate. Different controls were made by means of physical, chemical and ceramographic tests. This method can be used for the fabrication of fast reactor fuels or, previous mechanical dispersion in UO 2 powder, for the fabrication of thermal reactors fuels. (M.E.L.) [es

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

    Science.gov (United States)

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

    2017-04-01

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

  12. Interdiffusion and reaction between U and Zr

    Science.gov (United States)

    Park, Y.; Newell, R.; Mehta, A.; Keiser, D. D.; Sohn, Y. H.

    2018-04-01

    The microstructural development and diffusion kinetics were examined for the binary U vs. Zr system using solid-to-solid diffusion couples, U vs. Zr, annealed at 580 °C for 960 h, 650 °C for 480 h, 680 °C for 240 h, and 710 °C for 96 h. Scanning and transmission electron microscopies with X-ray energy dispersive spectroscopy were employed for detailed microstructural and compositional analyses. Interdiffusion and reaction in U vs. Zr diffusion couples primarily produced: δ-UZr2 solid solution (hP3) and α‧-U at 580 °C; and (γU,βZr) solid solution (cI2) and α‧-U at 650°, 680° and 710 °C. The α‧-phase was confirmed as a reduced variant of the α-U orthorhombic structure with lattice parameters, a × b × c = 2.65 × 5.40 × 4.75 (Å) with a negligible solubility for Zr at room temperature. Concentration profiles were examined to determine interdiffusion coefficients, integrated interdiffusion coefficients, and intrinsic diffusion coefficients using Boltzmann-Matano, Wagner, and Heumann analyses, respectively. Composition-dependence of interdiffusion coefficients were documented for α-U, δ-UZr2 (at 580 °C) and (γU,βZr) solid solution (at 650°, 680° and 710 °C). U was determined to intrinsically diffuse faster than Zr, approximately by an order of magnitude, in the δ-UZr2 at 580 °C, and (γU,βZr) phases at 650°, 680° and 710 °C. Based on Darken's approach, thermodynamic data available in literature were coupled to estimate the tracer diffusion coefficients and atomic mobilities of U and Zr.

  13. U-233 fuelled low critical mass solution reactor experiment PURNIMA II

    International Nuclear Information System (INIS)

    Srinivasan, M.; Chandramoleshwar, K.; Pasupathy, C.S.; Rasheed, K.K.; Subba Rao, K.

    1987-01-01

    A homogeneous U-233 uranyl nitrate solution fuelled BeO reflected, low critical mass reactor has been built at the Bhabha Atomic Research Centre, India. Christened PURNIMA II, the reactor was used for the study of the variation of critical mass as a function of fuel solution concentration to determine the minimum critical mass achievable for this geometry. Other experiments performed include the determination of temperature coefficient of reactivity, study of time behaviour of photoneutrons produced due to interaction between decaying U-233 fission product gammas and the beryllium reflector and reactor noise measurements. Besides being the only operational U-233 fuelled reactor at present, PURNIMA II also has the distinction of having attained the lowest critical mass of 397 g of fissile fuel for any operating reactor at the current time. The paper briefly describes the facility and gives an account of the experiments performed and results achieved. (author)

  14. Chemical Reduction of SIM MOX in Molten Lithium Chloride Using Lithium Metal Reductant

    Science.gov (United States)

    Kato, Tetsuya; Usami, Tsuyoshi; Kurata, Masaki; Inoue, Tadashi; Sims, Howard E.; Jenkins, Jan A.

    2007-09-01

    A simulated spent oxide fuel in a sintered pellet form, which contained the twelve elements U, Pu, Am, Np, Cm, Ce, Nd, Sm, Ba, Zr,Mo, and Pd, was reduced with Li metal in a molten LiCl bath at 923 K. More than 90% of U and Pu were reduced to metal to form a porous alloy without significant change in the Pu/U ratio. Small fractions of Pu were also combined with Pd to form stable alloys. In the gap of the porous U-Pu alloy, the aggregation of the rare-earth (RE) oxide was observed. Some amount of the RE elements and the actinoides leached from the pellet. The leaching ratio of Am to the initially loaded amount was only several percent, which was far from about 80% obtained in the previous ones on simple MOX including U, Pu, and Am. The difference suggests that a large part of Am existed in the RE oxide rather than in the U-Pu alloy. The detection of the RE elements and actinoides in the molten LiCl bath seemed to indicate that they dissolved into the molten LiCl bath containing the oxide ion, which is the by-product of the reduction, as solubility of RE elements was measured in the molten LiCl-Li2O previously.

  15. Structural study of Zr-based metallic glasses

    International Nuclear Information System (INIS)

    Matsubara, E.; Ichitsubo, T.; Saida, J.; Kohara, S.; Ohsumi, H.

    2007-01-01

    Structures of Zr 70 Ni 20 Al 10 , Zr 70 Cu 20 Al 10 , Zr 70 Cu 30 and Zr 70 Ni 30 amorphous alloys were analyzed by high-energy X-ray diffraction. The relatively stable Zr 2 Cu amorphous alloy shows a local atom arrangement different from the Zr 2 Cu crystalline phase. By contrast, the less stable Zr 70 Ni 30 amorphous alloy has a structure similar to Zr 2 Ni. In the Zr 70 Cu 20 Al 10 metallic glass, Zr-Al nearest neighbor pairs are introduced in the amorphous structure. In the Zr 70 Ni 20 Al 10 metallic glass, the strong correlation between Zr-Ni pairs is drastically modified by the formation of Zr-Al pairs. The presence of Zr-Al pairs in the ternary alloys suppresses the crystallization and stabilizes the glassy state

  16. U, Pu, and Am nuclear signatures of the Thule hydrogen bomb debris

    DEFF Research Database (Denmark)

    Eriksson, Mats; Lindahl, Patric; Roos, Per

    2008-01-01

    ). In the five hot particles examined, the measured uranium atomic ratio was U-235/U-238 = 1.02 +/- 0.16 and the Pu-isotopic ratios were as follows: Pu-240/Pu-239 0.0551 +/- 0.0008 (atom ratio), Pu-238/Pu239+240 = 0.0161 +/- 0.0005 (activity ratio), Pu-241/Pu239+240 = 0.87 +/- 0.12 (activity ratio), and Am-241...... than one Pu source involved in the accident, confirming earlier studies. The Pu-238/Pu239+240 activity ratio and the Pu-240/Pu-239 atomic ratio were divided into at least two Pu-isotopic ratio groups. For both Pu-isotopic ratios, one ratio group had identical ratios as the five hot particles described...... above and for the other groups the Pu isotopic ratios were lower (Pu-238/Pu239+240 activity ratio similar to 0.01 and the Pu-240/Pu-239 atomic ratio 0.03). On the studied particles we observed that the U/Pu ratio decreased as a function of the time these particles were present in the sediment. We...

  17. Irradiation performance of full-length metallic IFR fuels

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.

    1992-07-01

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

  18. Development of Melting Crucible Materials of Metallic Fuel Slug for SFR

    International Nuclear Information System (INIS)

    Kim, K. H.; Lee, C. T.; Oh, S. J.; Kim, S. K.; Lee, C. B.; Ko, Y. M.; Woo, W. M.

    2010-01-01

    The fabrication process of metallic fuel for SFR(sodium fast reactor) of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA(minor actinide) elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr-Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized

  19. Structural study of Zr-based metallic glasses

    Energy Technology Data Exchange (ETDEWEB)

    Matsubara, E. [Department of Materials Science and Engineering, Kyoto University, Kyoto 606-8501 (Japan)]. E-mail: e.matsubara@materials.mbox.media.kyoto-u.ac.jp; Ichitsubo, T. [Department of Materials Science and Engineering, Kyoto University, Kyoto 606-8501 (Japan); Saida, J. [Center of Interdisciplinary Research, Tohoku University, Sendai 980-8578 (Japan); Kohara, S. [JASRI, SPring-8, Sayo-gun, Hyogo 679-5198 (Japan); Ohsumi, H. [JASRI, SPring-8, Sayo-gun, Hyogo 679-5198 (Japan)

    2007-05-31

    Structures of Zr{sub 70}Ni{sub 20}Al{sub 10}, Zr{sub 70}Cu{sub 20}Al{sub 10}, Zr{sub 70}Cu{sub 30} and Zr{sub 70}Ni{sub 30} amorphous alloys were analyzed by high-energy X-ray diffraction. The relatively stable Zr{sub 2}Cu amorphous alloy shows a local atom arrangement different from the Zr{sub 2}Cu crystalline phase. By contrast, the less stable Zr{sub 70}Ni{sub 30} amorphous alloy has a structure similar to Zr{sub 2}Ni. In the Zr{sub 70}Cu{sub 20}Al{sub 10} metallic glass, Zr-Al nearest neighbor pairs are introduced in the amorphous structure. In the Zr{sub 70}Ni{sub 20}Al{sub 10} metallic glass, the strong correlation between Zr-Ni pairs is drastically modified by the formation of Zr-Al pairs. The presence of Zr-Al pairs in the ternary alloys suppresses the crystallization and stabilizes the glassy state.

  20. Electrodeposition of uranium and transuranic metals (Pu) on solid cathode

    International Nuclear Information System (INIS)

    Laplace, A. F.; Lacquement, J.; Willitt, J. L.; Finch, R. A.; Fletcher, G. A.; Williamson, M. A.

    2008-01-01

    The results from a study of U and Pu metal electrodeposition from molten eutectic LiCl-KCl on a solid inert cathode are presented. This study has been conducted using ∼ to 50 g of U-Pu together with rare earths (mostly Nd) and 1.5 kg of salt. The introduction of a three-electrode probe with an Ag/AgCl reference electrode has allowed voltammetric measurement during electrolysis and control of the cathode potential versus the reference. Cyclic and square-wave voltammetric measurements proved to be very useful tools for monitoring the electrolysis as well as selecting the cathode versus reference potential to maximize the separation between actinides and rare earths. The voltammetric data also highlighted the occurrence of back reactions between the cathode deposit and oxidizing equivalents formed at the anode that remained in the molten salt electrolyte. Any further electrolysis test needs to be conducted continuously and followed by immediate removal of the cathode to minimize those back reactions. (authors)

  1. Determination of Nb and Zr in U-Nb-Zr alloys by ICP-AES

    International Nuclear Information System (INIS)

    Wang Cuiping; Dong Shizhe; Li Lin; He Meiying

    2003-01-01

    The U-Nb-Zr alloy sample is dissolved by HNO 3 , H 2 O 2 and HF, and the contents of Nb and Zr in the sample are determined on the JY-70 II type ICP-AES by using the internal standard synchronous dilution method. The range of determination is 1%-10% and 0.33%-3.33%, respectively for Nb and Zr. The relative standard deviation is better than 3.2% for Nb, and 2.5% for Zr. The method is rapid and convenient for determining Nb and Zr in U-Nb-Zr alloy sample

  2. Synthesis and characterization of Pa(IV), Np(IV), and Pu(IV) borohydrides

    International Nuclear Information System (INIS)

    Banks, R.H.; Edelstein, N.M.

    1979-12-01

    The actinide borohydrides of Pa, Np, and Pu have been prepared and some of their physical and optical properties measured. X-ray powder diffraction photographs of Pa(BH 4 ) 4 have shown that it is isostructural to Th(BH 4 ) 4 and U(BH 4 ) 4 . Np(BH 4 ) 4 and Pu(BH 4 ) 4 are much more volatile than the borohydrides of Th, Pa, and U and are liquids at room temperature. Results from low-temperature single-crystal x-ray diffraction investigation of Np(BH 4 ) 4 show that its structure is very similar to Zr(BH 4 ) 4 . With the data from low-temperature infrared and Raman spectra, a normal coordinate analysis on Np(BH 4 ) 4 and Np(BD 4 ) 4 has been completed. EPR experiments on Np(BH 4 ) 4 /Zr(BH 4 ) 4 and Np(BD 4 ) 4 /Zr(BD 4 ) 4 have characterized the ground electronic state. 5 figures

  3. Cation interdiffusion in the UO2 - (U, Pu)O2 and UO2 - PuO2 systems

    International Nuclear Information System (INIS)

    Leme, D.G.

    1985-01-01

    The interdiffusion of U and Pu ions in UO sub(2 +- x) - (U sub(0,83) Pu sub(0,17))O sub(2 + - x) and UO sub(2 + - x) -PuO sub(2 - x) sintered pellets and UO sub(2 +- x) -(U sub(0,82) Pu sub(0,18))O sub(2 + - x) single crystals has been studied as a function of the oxygen potential ΔG sup(-) (O 2 ) or the stoichiometric ratio O/M. The diffusion profiles of UO 2 /(U,Pu)O 2 and UO 2 /PuO 2 couples of different O/M ratios have been measured using high resolution α-spectrometer and microprobe. Thermal annealing of the specimens was performed in controlled atmospheres using either CO-CO 2 gas mixtures for constant O/M ratios or purified argon. The interdiffusion profiles have been analysed by means of the Boltzmann-Matano and Hall methods. The interdiffusion coefficient D sus(approx.) increases with increasing Pu content in sintered pellets (up to 17 wt. %PuO 2 ) showing a strong dependence of D sup(approx.) on the O/M ratio. The micropobe results show that the interdiffusion along grain boundaries is the main diffusion mechanism in the pellets. Experiments have also been carried out in single cristals to measure just the bulk-interdiffusion and avoiding effects due to grain boundaries. A marked dependence of D sup(approx.) on O/M ratio or on oxygen potential ΔG sup(-) (O 2 ), similar to the dependence already reported for self diffusion by means of radioactive tracers, has also been observed. (Author) [pt

  4. HPLC method for determination of Th, U and Pu in irradiated (Th,Pu)O{sub 2} using mandelic acid as an eluent

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Pranaw; Paul, Sumana; Jaison, Perumpillil George; Telmore, Vijay Madhukar; Alamelu, Devanathan; Aggarwal, Suresh Kumar [Bhabha Atomic Research Centre, Mumbai (India). Fuel Chemistry Div.

    2014-07-01

    Studies for chromatographic separation of Th, U(VI) and Pu(IV) were carried out using mandelic acid as an eluent. The different chromatographic conditions like concentration of mandelic acid, pH of the mobile phase, presence of MeOH and effect of ion interaction reagent (IIR) were studied. The method was optimized for the separation of Th, U(VI) and Pu(IV). At pH<3.5 of mobile phase, Pu(IV) was more retained compared to U(VI) whereas at pH>3.5, reverse trend was observed. The optimized parameters were employed for the separation and determination of Th, U(VI) and Pu(IV) in a dissolved solution of irradiated (Th,Pu)O{sub 2}. Sample treatment was optimized to minimize loss of Pu and Th during chromatographic determination. Studies were carried out using two IIRs to understand the anomalous chromatographic behavior of Pu(IV). Retention behavior of different oxidation states of Pu viz. Pu(III), Pu(IV) and Pu(VI) was also studied in mandelic acid.

  5. Solubility of reactor fuels in the mouse lung with respect to their U/Pu and 238Pu/239Pu ratios

    International Nuclear Information System (INIS)

    Talbot, R.J.; Baker, S.T.

    1989-01-01

    The studies reported were designed to assess the comparative in vivo solubilities of a range of plutonium containing reactor fuels. To simulate these fuels, mixed U/Pu oxides were prepared and calcined at 1600 0 C. A plutonium content of 3% w/w was chosen as typical of water-cooled reactor fuel. Higher concentrations of plutonium (10, 20 and 30%) were included to simulate fast reactor fuel. As it is known that 238 PuO 2 , with high specific activity, is translocated more rapidly from lung than 239 PuO 2 , the effect of isotopic composition of plutonium in simulated reactor fuels was also investigated. For this purpose, both the water-cooled and fast-reactor fuels were prepared with plutonium containing 2% of 238 Pu by weight. The resulting oxides had about 6 times the specific activity of those prepared with 239 Pu. Groups of mice were killed at 1, 3, 6, 12 and 18 months after inhaling aerosols of the simulated reactor fuels. After 3 months, measurements of Pu retention in the lung showed no marked differences between materials. After 6 months, measurements of plutonium deposited in the liver and skeleton showed that mixed U/Pu oxides were more soluble in vivo than 239 PuO 2 . Their solubility was inversely related to their plutonium content. The addition of 238 Pu to the plutonium resulted in enhanced translocation of plutonium from the lung, in the cases of water-cooled reactor fuel by a factor of two. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, Dane [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong Austin [Univ. of Wisconsin, Madison, WI (United States)

    2013-10-28

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

  7. Contribution to the study of the (U,Pu)C,N system

    International Nuclear Information System (INIS)

    Lorenzelli, R.

    1968-01-01

    The reactions of UC, PuC, (U,Pu)C, UC 2 and U(C 1-x O x ) with nitrogen at moderate temperatures (room temperature to 400 C) are described. The influence of the uptake of nitrogen by the powders necessary to sinter the carbides upon the nature of the final product has been investigated; it has been shown that the sintered carbides are hyper-stoichiometric. The reactions of carbon with UN, PuN and (U,Pu)N has also been studied. Under vacuum, carbon reacts on the nitrides at temperatures as low as 1100 C; nitrogen is replaced by carbon and the final product is a carbonitride. The reaction is: MN + x C → MN 1-x C x + x/2N 2 . The reaction is limited and the carbonitrides have a fixed composition in presence of M 2 C 3 or MC 2 ; hence it is impossible to produce pure MC using the reaction. The ternary diagram U-C-N, Pu-C-N and (U,Pu)C-N have been drawn. They show clearly that it is possible to obtain single phase carbonitrides in a wide domain of compositions. (author) [fr

  8. Obtaining of U-2.5Zr7.5Nb and U-3Zr-9Nb alloys by sintering process

    International Nuclear Information System (INIS)

    Mazzeu, Thiago de Oliveira; Paula, Joao Bosco de; Ferraz, Wilmar Barbosa; Santos, Ana Maria Matildes dos; Brina, Jose Giovanni Mascarenhas

    2011-01-01

    The development of metallic fuels with low enrichment to be used in research and test reactors, as well in the future pressurized water reactors, focuses on the search for uranium alloys of high density. Alloying elements such as Zr, Nb and Mo are added to uranium to improve fuel performance in reactors. In this context, the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) in Belo Horizonte is developing the U-2.5Zr-7.5Nb and U- 3Zr-9Nb (weight %) alloys by the innovative process of sintering that utilizes raw materials in the form of powders. The powders were pressed at 400MPa and then sintered under a vacuum of about 5 x 10-6 Torr at temperatures ranging from 1050 deg to 1300 deg C. The densities of the alloys were measured geometrically and by hydrostatic method using water. The microstructures of the pellets were observed by scanning electron microscopy (SEM) and the elements of alloying were identified by energy dispersive X-ray spectroscopy (SEM/EDS) analysis. The obtained results showed a small increasing density with rising sintering temperature. The highest density achieved was approximately 80% of theoretical density. It was also qualitatively observed that the superficial oxidation of the pellets increased with increasing sintering temperature thus avoiding the fusion of the alloys at higher temperatures. (author)

  9. Fuel cycle modelling of open cycle thorium-fuelled nuclear energy systems

    International Nuclear Information System (INIS)

    Ashley, S.F.; Lindley, B.A.; Parks, G.T.; Nuttall, W.J.; Gregg, R.; Hesketh, K.W.; Kannan, U.; Krishnani, P.D.; Singh, B.; Thakur, A.; Cowper, M.; Talamo, A.

    2014-01-01

    Highlights: • We study three open cycle Th–U-fuelled nuclear energy systems. • Comparison of these systems is made to a reference U-fuelled EPR. • Fuel cycle modelling is performed with UK NNL code “ORION”. • U-fuelled system is economically favourable and needs least separative work per kWh. • Th–U-fuelled systems offer negligible waste and proliferation resistance advantages. - Abstract: In this study, we have sought to determine the advantages, disadvantages, and viability of open cycle thorium–uranium-fuelled (Th–U-fuelled) nuclear energy systems. This has been done by assessing three such systems, each of which requires uranium enriched to ∼20% 235 U, in comparison to a reference uranium-fuelled (U-fuelled) system over various performance indicators, spanning material flows, waste composition, economics, and proliferation resistance. The values of these indicators were determined using the UK National Nuclear Laboratory’s fuel cycle modelling code ORION. This code required the results of lattice-physics calculations to model the neutronics of each nuclear energy system, and these were obtained using various nuclear reactor physics codes and burn-up routines. In summary, all three Th–U-fuelled nuclear energy systems required more separative work capacity than the equivalent benchmark U-fuelled system, with larger levelised fuel cycle costs and larger levelised cost of electricity. Although a reduction of ∼6% in the required uranium ore per kWh was seen for one of the Th–U-fuelled systems compared to the reference U-fuelled system, the other two Th–U-fuelled systems required more uranium ore per kWh than the reference. Negligible advantages and disadvantages were observed for the amount and the properties of the spent nuclear fuel (SNF) generated by the systems considered. Two of the Th–U-fuelled systems showed some benefit in terms of proliferation resistance of the SNF generated. Overall, it appears that there is little

  10. Chronology of Pu isotopes and 236U in an Arctic ice core.

    Science.gov (United States)

    Wendel, C C; Oughton, D H; Lind, O C; Skipperud, L; Fifield, L K; Isaksson, E; Tims, S G; Salbu, B

    2013-09-01

    In the present work, state of the art isotopic fingerprinting techniques are applied to an Arctic ice core in order to quantify deposition of U and Pu, and to identify possible tropospheric transport of debris from former Soviet Union test sites Semipalatinsk (Central Asia) and Novaya Zemlya (Arctic Ocean). An ice core chronology of (236)U, (239)Pu, and (240)Pu concentrations, and atom ratios, measured by accelerator mass spectrometry in a 28.6m deep ice core from the Austfonna glacier at Nordaustlandet, Svalbard is presented. The ice core chronology corresponds to the period 1949 to 1999. The main sources of Pu and (236)U contamination in the Arctic were the atmospheric nuclear detonations in the period 1945 to 1980, as global fallout, and tropospheric fallout from the former Soviet Union test sites Novaya Zemlya and Semipalatinsk. Activity concentrations of (239+240)Pu ranged from 0.008 to 0.254 mBq cm(-2) and (236)U from 0.0039 to 0.053 μBq cm(-2). Concentrations varied in concordance with (137)Cs concentrations in the same ice core. In contrast to previous published results, the concentrations of Pu and (236)U were found to be higher at depths corresponding to the pre-moratorium period (1949 to 1959) than to the post-moratorium period (1961 and 1962). The (240)Pu/(239)Pu ratio ranged from 0.15 to 0.19, and (236)U/(239)Pu ranged from 0.18 to 1.4. The Pu atom ratios ranged within the limits of global fallout in the most intensive period of nuclear atmospheric testing (1952 to 1962). To the best knowledge of the authors the present work is the first publication on biogeochemical cycles with respect to (236)U concentrations and (236)U/(239)Pu atom ratios in the Arctic and in ice cores. Copyright © 2013 Elsevier B.V. All rights reserved.

  11. Chemical thermodynamic representation of (U, Pu, Am)O2-x

    International Nuclear Information System (INIS)

    Osaka, Masahiko; Namekawa, Takashi; Kurosaki, Ken; Yamanaka, Shinsuke

    2005-01-01

    The oxygen potential isotherms of (U, Pu, Am)O 2-x were represented by a chemical thermodynamic model proposed by Lindemer et al. It was assumed in the present model that (U, Pu, Am)O 2-x consisted of the chemical species [UO 2 ], [PuO 2 ], [Pu 4/3 O 2 ], [AmO 2 ] and [Am 5/4 O 2 ] in a pseudo-quaternary system by treating the reduction rates of Pu and Am as identical; furthermore an interaction between [Am 5/4 O 2 ] and [UO 2 ] was introduced. The agreement between analytical and experimental isotherms was good, but the analytical values slightly overestimated the experimental values especially in the case of lower Am content. Adding an interaction between [Am 5/4 O 2 ] and [PuO 2 ] to the model resulted in a better representation

  12. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-07

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

  13. Demonstration of pyrometallurgical processing for metal fuel and HLW

    International Nuclear Information System (INIS)

    Tadafumi, Koyama; Kensuke, Kinoshita; Takatoshi, Hizikata; Tadashi, Inoue; Ougier, M.; Rikard, Malmbeck; Glatz, J.P.; Lothar, Koch

    2001-01-01

    CRIEPI and JRC-ITU have started a joint study on pyrometallurgical processing to demonstrate the capability of this type of process for separating actinide elements from spent fuel and HLW. The equipment dedicated for this experiments has been developed and installed in JRC-ITU. The stainless steel box equipped with tele-manipulators is operated under pure Ar atmosphere, and prepared for later installation in a hot cell. Experiments on pyro-processing of un-irradiated U-Pu-Zr metal alloy fuel by molten salt electrorefining has been carried out. Recovery of U and Pu from this type alloy fuel was first demonstrated with using solid iron cathode and liquid Cd cathode, respectively. (author)

  14. Chronology of Pu isotopes and {sup 236}U in an Arctic ice core

    Energy Technology Data Exchange (ETDEWEB)

    Wendel, C.C., E-mail: cato.wendel@umb.no [Isotope Laboratory, Department of Plant and Environmental Sciences, Agricultural University of Norway, P.O. Box 5003, N-1432 Aas (Norway); Oughton, D.H., E-mail: deborah.oughton@umb.no [Isotope Laboratory, Department of Plant and Environmental Sciences, Agricultural University of Norway, P.O. Box 5003, N-1432 Aas (Norway); Lind, O.C., E-mail: ole-christian.lind@umb.no [Isotope Laboratory, Department of Plant and Environmental Sciences, Agricultural University of Norway, P.O. Box 5003, N-1432 Aas (Norway); Skipperud, L., E-mail: lindis.skipperud@umb.no [Isotope Laboratory, Department of Plant and Environmental Sciences, Agricultural University of Norway, P.O. Box 5003, N-1432 Aas (Norway); Fifield, L.K., E-mail: keith.fifield@anu.edu.au [Department of Nuclear Physics, Australian National University, Canberra ACT 0200 (Australia); Isaksson, E., E-mail: elisabeth.isaksson@npolar.no [Norwegian Polar Institute, Fram Centre, Hjalmar Johansens Gate 14, N9296 Tromsø (Norway); Tims, S.G., E-mail: steve.tims@anu.edu.au [Department of Nuclear Physics, Australian National University, Canberra ACT 0200 (Australia); Salbu, B., E-mail: brit.salbu@umb.no [Isotope Laboratory, Department of Plant and Environmental Sciences, Agricultural University of Norway, P.O. Box 5003, N-1432 Aas (Norway)

    2013-09-01

    In the present work, state of the art isotopic fingerprinting techniques are applied to an Arctic ice core in order to quantify deposition of U and Pu, and to identify possible tropospheric transport of debris from former Soviet Union test sites Semipalatinsk (Central Asia) and Novaya Zemlya (Arctic Ocean). An ice core chronology of {sup 236}U, {sup 239}Pu, and {sup 240}Pu concentrations, and atom ratios, measured by accelerator mass spectrometry in a 28.6 m deep ice core from the Austfonna glacier at Nordaustlandet, Svalbard is presented. The ice core chronology corresponds to the period 1949 to 1999. The main sources of Pu and {sup 236}U contamination in the Arctic were the atmospheric nuclear detonations in the period 1945 to 1980, as global fallout, and tropospheric fallout from the former Soviet Union test sites Novaya Zemlya and Semipalatinsk. Activity concentrations of {sup 239+240}Pu ranged from 0.008 to 0.254 mBq cm{sup −2} and {sup 236}U from 0.0039 to 0.053 μBq cm{sup −2}. Concentrations varied in concordance with {sup 137}Cs concentrations in the same ice core. In contrast to previous published results, the concentrations of Pu and {sup 236}U were found to be higher at depths corresponding to the pre-moratorium period (1949 to 1959) than to the post-moratorium period (1961 and 1962). The {sup 240}Pu/{sup 239}Pu ratio ranged from 0.15 to 0.19, and {sup 236}U/{sup 239}Pu ranged from 0.18 to 1.4. The Pu atom ratios ranged within the limits of global fallout in the most intensive period of nuclear atmospheric testing (1952 to 1962). To the best knowledge of the authors the present work is the first publication on biogeochemical cycles with respect to {sup 236}U concentrations and {sup 236}U/{sup 239}Pu atom ratios in the Arctic and in ice cores. - Highlights: • Concentrations and atom ratios of Pu and {sup 236}U determined in an Arctic ice core. • Concentrations of U and Pu found to be higher pre- than post-moratorium. • U and Pu concentrations

  15. Aqueous corrosion study on U-Zr alloy

    International Nuclear Information System (INIS)

    Pal, Titas; Venkatesan, V.; Kumar, Pradeep; Khan, K.B.; Kumar, Arun

    2009-01-01

    In low power or research reactor, U-Zr alloy is a potential candidate for dispersion fuel. Moreover, Zirconium has a low thermal-neutron cross section and uranium alloyed with Zr has excellent corrosion resistance and dimensional stability during thermal cycling. In the present study aqueous corrosion behavior of U-Zr alloy samples was studied in autoclave at 200 deg C temperature. Corrosion rate was determined from weight loss with time. (author)

  16. Quantitative dissolution of (U, Pu)O2 MOX (0.4% to 44% PuO2) using microwave heating technique

    International Nuclear Information System (INIS)

    Malav, R.K.; Fulzele, Ajit K.; Prakash, Amrit; Afzal, Md.; Panakkal, J.P.

    2011-01-01

    AFFF has fabricated the (U, Pu)O 2 mixed oxide fuels for PHWRs, BWRs, PFBRs and FBTRs. The quantitative dissolution of the fuel samples are required within time for accurate determination of uranium-plutonium in chemical quality control laboratory. This paper describes the use of microwave heating technique in quantitative dissolution of (U, Pu)O 2 MOX (from 0.4% to 44% PuO 2 ). (author)

  17. Safety characteristics of the US advanced liquid metal reactor core

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Gyorey, G.L.; Lipps, A.J.; Wu, T.

    1991-01-01

    The U.S. Advanced Liquid Metal Reactor (ALMR) design employs innovative, passive features to provide an unprecedented level of public safety and the ability to demonstrate this safety to the public. The key features employed in the core design to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters, and gas expansion modules. In addition, the reactor vessel and closure are designed to have the capability to withstand, with large margins, the maximum possible core disruptive accident without breach and radiological release. (author)

  18. Alternative Fabrication of Recycling Fast Reactor Metal Fuel

    International Nuclear Information System (INIS)

    Kim, Ki-Hwan; Kim, Jong Hwan; Song, Hoon; Kim, Hyung-Tae; Lee, Chan-Bock

    2015-01-01

    Metal fuels such as U-Zr/U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. In order to develop innovative fabrication method of metal fuel for preventing the evaporation of volatile elements such as Am, modified casting under inert atmosphere has been applied for metal fuel slugs for SFR. Alternative fabrication method of fuel slugs has been introduced to develop an improved fabrication process of metal fuel for preventing the evaporation of volatile elements. In this study, metal fuel slugs for SFR have been fabricated by modified casting method, and characterized to evaluate the feasibility of the alternative fabrication method. In order to prevent evaporation of volatile elements such as Am and improve quality of fuel slugs, alternative fabrication methods of metal fuel slugs have been studied in KAERI. U-10Zr-5Mn fuel slug containing volatile surrogate element Mn was soundly cast by modified injection casting under modest pressure. Evaporation of Mn during alternative casting could not be detected by chemical analysis. Mn element was most recovered with prevention of evaporation by alternative casting. Modified injection casting has been selected as an alternative fabrication method in KAERI, considering evaporation prevention, and proven benefits of high productivity, high yield, and good remote control

  19. On the fission probability for 235U, 239Pu and 241Pu

    International Nuclear Information System (INIS)

    Benzi, V.; Maino, G.; Menapace, E.

    1978-01-01

    An evaluation of the GAMMAsub(n)/GAMMAsub(f) ratio for the 236 U, 240 Pu and 242 Pu compound nuclei is carried out. First chance and second chance fission cross sections are estimated from the ''evaporation'' model; particularly, a largely increasing trend was found for the first chance fission cross section above the (n,n'f) process threshold. The GAMMAsub(n)/GAMMAsub(f) ratios for the analyzed nuclei show a bump-like structure, that seems to be in agreement with the theoretical predictions reported in literature

  20. Low silicon U(Al,Si)3 stabilization by Zr addition

    International Nuclear Information System (INIS)

    Pizarro, L.M.; Alonso, P.R.; Rubiolo, G.H.

    2009-01-01

    Previous knowledge states that (U,Zr)Al 3 and U(Al,Si) 3 phases with Zr and Si content higher than 6 at.% (7.7 wt%) and 4 at.% (1.4 wt%), respectively, does not partially transform to UAl 4 at 600 o C. In this work, four alloys within the quaternary system U-Al-Si-Zr were made with a fixed nominal 0.18 at.% (0.1 wt%) Si content in order to assess the synergetic effect of both Zr and Si alloying elements to the thermodynamic stability of the (U,Zr)(Al,Si) 3 phase. Heat treatments at 600 deg. C were undertaken and samples were analyzed by means of XRD, EPMA and EDS techniques. A remarkable conclusion is that addition of 0.3 at.% Si in the (U,Zr)(Al,Si) 3 phase reduces in 2.7 at.% the necessary Zr content to inhibit its transformation to U(Al,Si) 4 .

  1. Cu-Zr-Ag bulk metallic glasses based on Cu8Zr5 icosahedron

    International Nuclear Information System (INIS)

    Xia Junhai; Qiang Jianbing; Wang Yingmin; Wang Qing; Dong Chuang

    2007-01-01

    Based on the cluster line criterion, the Ag addition into the Cu 8 Zr 5 cluster composition is investigated for the search of ternary Cu-Zr-Ag bulk metallic glasses with high glass forming abilities. Two initial binary compositions Cu 0.618 Zr 0.382 and Cu 0.64 Zr 0.36 are selected. The former one corresponds to a deep eutectic point; it is also the composition of the Cu 8 Zr 5 icosahedron, which is derived from the Cu 8 Zr 3 structure. The latter one, which can be regarded as the Cu 8 Zr 5 cluster plus a glue atom Cu, is the best glass-forming composition in the Cu-Zr binary system. Two composition lines (Cu 0.618 Zr 0.382 ) 1-x Ag x and (Cu 0.64 Zr 0.36 ) 1-x Ag x are thus constructed in the Cu-Zr-Ag system by linking these two compositions with the third constitute Ag. A series of Cu-Zr-Ag bulk metallic glasses are found with 2-8 at.% Ag contents in both composition lines. The optimum composition (Cu 0.618 Zr 0.382 ) 0.92 Ag 0.08 within the searched region with the highest T g /T l = 0.633, is located along the cluster line (Cu 0.618 Zr 0.382 ) 1-x Ag x , where the deep eutectic Cu 0.618 Zr 0.382 exactly corresponds to the dense packing cluster Cu 8 Zr 5 . The alloying mechanism is discussed in the light of atomic size and electron concentration factors

  2. Interdiffusion between U(Mo,Pt) or U(Mo,Zr) and Al or Al A356 alloy

    International Nuclear Information System (INIS)

    Komar Varela, C.; Mirandou, M.; Arico, S.; Balart, S.; Gribaudo, L.

    2009-01-01

    Solid state reactions in chemical diffusion couples U-7 wt.%Mo-0.9 wt.%Pt/Al at 580 deg. C and U-7 wt.%Mo-0.9 wt.%Pt/Al A356 alloy, U-7 wt.%Mo-1 wt.%Zr/Al and U-7 wt.%Mo-1 wt.%Zr/Al A356 alloy at 550 deg. C were characterized. Results were obtained from optical and scanning electron microscopy, electron probe microanalysis and X-ray diffraction. The UAl 3, UAl 4 and Al 20 Mo 2 U phases were identified in the interaction layers of γU(Mo,Pt)/Al and γU(Mo,Zr)/Al diffusion couples. Al 43 Mo 4 U 6 ternary compound was also identified in γU(Mo,Zr)/Al due to the decomposition of γU(Mo,Zr) phase. The U(Al,Si) 3 and U 3 Si 5 phases were identified in the interaction layers of γU(Mo,Pt)/Al A356 and γU(Mo,Zr)/Al A356 diffusion couples. These phases are formed due to the migration of Si to the interaction layer. In the diffusion couple U(Mo,Zr)/Al A356, Zr 5 Al 3 phase was also identified in the interaction layer. The use of synchrotron radiation at Brazilian Synchrotron Light Laboratory (LNLS, CNPq, Campinas, Brazil) was necessary to achieve a complete crystallographic characterization.

  3. Metal-carbide multilayers for molten Pu containment

    International Nuclear Information System (INIS)

    Summers, T.S.E.; Curtis, P.G.; Juntz, R.S.; Krueger, R.L.

    1991-12-01

    Multilayers composed of nine or ten alternating layers of Ta or W and TaC were studied for the feasibility of their use in containing molten plutonium (Pu) at 1200 degrees C. Single layers of W and TaC were also investigated. A two-source electron beam evaporation process was developed to deposit these coatings onto the inside surface of hemispherical Ta cups about 38 mm in diameter. Pu testing was done by melting Pu in the coated hemispherical cups and holding them under vacuum at 1200 degrees C for two hours. Metallographic examination and microprobe analysis of cross sections showed that Pu had penetrated to the Ta substrate in all cases to some extent. Full penetration to the outer surface of the Ta substrate, however, occurred in only a few of the samples. The fact that full penetration occurred in any of the samples suggests that it would have occurred in uncoated Ta under these testing conditions which in turn suggests that the multilayer coatings do afford some protection against Pu attack. The TaC used for these specimens was wet by Pu under these testing conditions, and following testing, Pu was found uniformly distributed throughout the carbide layers which appeared to be rather porous. Pu was seen in the W and Ta layers only when exposed directly to molten Pu during testing or near defects suggesting that Pu penetrated the multilayers at defects in the coating and traveled parallel to the layers along the carbide layers. These results indicate that the use of alternating metal and ceramic layers for Pu containment should be possible through the use of nonporous ceramic that is not wet by molten Pu and defect-free films

  4. Minimum pickup velocity (U{sub pu}) of nanoparticles in gas–solid pneumatic conveying

    Energy Technology Data Exchange (ETDEWEB)

    Anantharaman, Aditya [Nanyang Technological University, School of Chemical and Biomedical Engineering (Singapore); Ommen, J. Ruud van [Delft University of Technology, Department of Chemical Engineering (Netherlands); Chew, Jia Wei, E-mail: JChew@ntu.edu.sg [Nanyang Technological University, School of Chemical and Biomedical Engineering (Singapore)

    2015-12-15

    This paper is the first systematic study of the pneumatic conveying of nanoparticles. The minimum pickup velocity, U{sub pu}, of six nanoparticle species of different materials [i.e., silicon dioxide (SiO{sub 2}), aluminum oxide (Al{sub 2}O{sub 3}), and titanium dioxide (TiO{sub 2})] and surfaces (i.e., apolar and polar) was determined by the weight loss method. Results show that (1) due to relative lack of hydrogen bonding, apolar nanoparticles had higher mass loss values at the same velocities, mass loss curves with accentuated S-shaped profiles, and lower U{sub pu} values, (2) among the three species, SiO{sub 2}, which has the lowest Hamaker coefficient, exhibited the greatest discrepancy between apolar and polar surfaces with respect to both mass loss curves and U{sub pu} values, (3) U{sub mf,polar}/U{sub mf,apolar} was between 1 and 3.5 times that of U{sub pu,polar}/U{sub pu,apolar} due to greater extents of hydrogen bonding associated with U{sub mf}, (4) U{sub pu} values were at least an order-of-magnitude lower than that expected from the well-acknowledged U{sub pu} correlation (Kalman et al., Powder Technol 160:103–113, 2005) due to agglomeration, (5) although nanoparticles should be categorized as Zone III (Kalman et al. 2005) (or Geldart group C, Powder Technol 7:285–292, 1973), the nanoparticles, and primary and complex agglomerates agreed more with the Zone I (or Geldart group B) correlation.

  5. Thermodynamic assessments and inter-relationships between systems involving Al, Am, Ga, Pu, and U

    Energy Technology Data Exchange (ETDEWEB)

    Perron, A., E-mail: perron1@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Turchi, P.E.A.; Landa, A. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Oudot, B.; Ravat, B.; Delaunay, F. [CEA-Centre de Valduc, 21120, Is sur Tille (France)

    2016-12-15

    A newly developed self-consistent CALPHAD thermodynamic database involving Al, Am, Ga, Pu, and U is presented. A first optimization of the slightly characterized Am-Al and completely unknown Am-Ga phase diagrams is proposed. To this end, phase diagram features as crystal structures, stoichiometric compounds, solubility limits, and melting temperatures have been studied along the U-Al → Pu-Al → Am-Al, and U-Ga → Pu-Ga → Am-Ga series, and the thermodynamic assessments involving Al and Ga alloying are compared. In addition, two distinct optimizations of the Pu-Al phase diagram are proposed to account for the low temperature and Pu-rich region controversy. The previously assessed thermodynamics of the other binary systems (Am-Pu, Am-U, Pu-U, and Al-Ga) is also included in the database and is briefly described in the present work. Finally, predictions on phase stability of ternary and quaternary systems of interest are reported to check the consistency of the database.

  6. Effect of the Zr elements with thermal properties changes of U-7Mo-xZr/Al dispersion fuel

    International Nuclear Information System (INIS)

    Supardjo; Agoeng Kadarjono; Boybul; Aslina Br Ginting

    2016-01-01

    Thermal properties data of nuclear fuel is required as input data to predict material properties change phenomenon during the fabrication process and irradiated in a nuclear reactor. Study the influence of Zr element in the U-7Mo-xZr/Al (x = 1%, 2% and 3%) fuel dispersion to changes in the thermal properties at various temperatures have been stiffened. Thermal analysis includes determining the melting temperature, enthalpy, and phase changes made using Differential Thermal Analysis (DTA) in the temperature range between 30 °C up to 1400 °C, while the heat capacity of U-7Mo-xZr alloy and U-7Mo-xZr/Al dispersion fuel using Differential Scanning Calorimeter (DSC) at room temperature up to 450 °C. Thermal analyst data DTA shows that Zr levels of all three compositions showed a similar phenomenon. At temperatures between 565.60 °C - 584.98 °C change becomes α + δ to α + γ phase and at 649.22 °C – 650.13 °C happen smelting Al matrix Occur followed by a reaction between Al matrix with U-7Mo-xZr on 670.38 °C - 673.38 °C form U (Al, Mo)x Zr. Furthermore a phase change α + β becomes β + γ Occurs at temperatures 762.08 °C - 776.33 °C and diffusion between the matrix by U-7Mo-xZr/Al on 853.55 °C - 875.20 °C. Every phenomenon that Occurs, enthalpy posed a relative stable. Consolidation of uranium Occur in 1052.42 °C - 1104.99 °C and decomposition reaction of U (Al, Mo)x and U (Al, Zr)_x becomes (UAl_4, UAl_3, UAl_2), U-Mo, and UZr on 1328,34 °C - 1332,06 °C , The existence of Zr in U-Mo alloy increases the heat capacity of the U-7Mo-xZr/Al, dispersion fuel and the higher heat capacity of Zr levels increased due to interactions between the atoms of Zr with Al matrix so that the heat absorbed by the fuel increase. (author)

  7. Subcritical measurements with a cylindrical tank of Pu-U nitrate

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; Valentine, T.E.; King, W.T.

    1997-01-01

    This series of measurements with a mixed Pu-U nitrate solution (280 g Pu/liter, 180 g U/liter) in a 35.54-cm-diam cylindrical tank provides a wide variety of experimental data for subcritical configurations that can be used to verify calculational methods and nuclear data. The Pu contained 7.85 wt% 240 Pu and the uranium was natural uranium. The measurements performed were: inverse count rate, prompt neutron decay constants, inverse kinetics, and frequency analysis by the 252 Cf source driven method. These data are presented in sufficient detail that the results of the experiments can be calculated directly. For purposes of extrapolating to the delayed critical height the ratio of spectral densities was linear with height and thus provided the best estimate of critical height

  8. Optimization of binary breeder reactor V - Binary breeder reactors with two and four zones and a conventional LMFBR - (Pu/U) of two zones

    International Nuclear Information System (INIS)

    Dias, A.F.; Ishiguro, Y.

    1986-04-01

    Comparative analyses of a commercial-size Pu/U-fueled liquid metal fast breeder reactor and two binary breeder reactors with different numbers of enrichment zones have been done. Principal parameters of comparison are safety and breeding characteristics and reactivity losses during an operational cycle. The comparison shows that in a binary breeder reactor, good breeding characteristics in both cycles, Pu/U and U/Th, in addition to a possibility of an efficient utilization of thorium, and superior inherent safety than current LMFBRs can be achieved. (Author) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

    2013-07-01

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

  10. Zr-(Cu,Ag)-Al bulk metallic glasses

    International Nuclear Information System (INIS)

    Jiang, Q.K.; Wang, X.D.; Nie, X.P.; Zhang, G.Q.; Ma, H.; Fecht, H.-J.; Bendnarcik, J.; Franz, H.; Liu, Y.G.; Cao, Q.P.; Jiang, J.Z.

    2008-01-01

    In this paper, we report the formation of a series Zr-(Cu,Ag)-Al bulk metallic glasses (BMGs) with diameters at least 20 mm and demonstrate the formation of about 25 g amorphous metallic ingots in a wide Zr-(Cu,Ag)-Al composition range using a conventional arc-melting machine. The origin of high glass-forming ability (GFA) of the Zr-(Cu,Ag)-Al alloy system has been investigated from the structural, thermodynamic and kinetic points of view. The high GFA of the Zr-(Cu,Ag)-Al system is attributed to denser local atomic packing and the smaller difference in Gibbs free energy between amorphous and crystalline phases. The thermal, mechanical and corrosion properties, as well as elastic constants for the newly developed Zr-(Cu,Ag)-Al BMGs, are also presented. These newly developed Ni-free Zr-(Cu,Ag)-Al BMGs exhibit excellent combined properties: strong GFA, high strength, high compressive plasticity, cheap and non-toxic raw materials and biocompatible property, as compared with other BMGs, leading to their potential industrial applications

  11. Refining U-Zr-Nb alloys by remelting

    International Nuclear Information System (INIS)

    Aguiar, B.M.; Kniess, C.T.; Riella, H.G.; Ferraz, W.B.

    2011-01-01

    The high density U-Zr-Nb and U-Nb uranium-based alloys can be employed as nuclear fuel in a PWR reactor due to their high density and nuclear properties. These alloys can stabilize the gamma phase, however, according to TTT diagrams, at the working temperature of a PWR reactor, all gamma phase transforms to α'' phase in a few hours. To avoid this kind of transformation during the nuclear reactor operation, the U-Zr-Nb alloy and U-Nn are used in α'' phase. The stability of α'' phase depends on the alloy composition and cooling rate. The alloy homogenization has to be very effective to eliminate precipitates rich in Zr and Nb to avoid changes in the alloying elements contents in the matrix. The homogenization was obtained by remelting the alloy and keeping it in the liquid state for enough time to promote floating of the precipitates (usually carbides, less dense) and leaving the matrix free of precipitates. However, this floating by density difference may result in segregation between the alloying elements (Nb and Zr, at the top) and uranium (at the bottom). The homogenized alloys were characterized in terms of metallographic techniques, optical microscopy, scanning electronic microscopy, EDS and X-ray diffraction. In this paper, it is shown that the contents of Zr and Nb at the bottom and at the top of the matrix are constant. (author)

  12. Characterization of the interaction layer in diffusion couples U-Mo-Zr/Al and U-Mo-Zr/Al-A356 at 550 C degrees; Caracterizacion de la zona de interaccion en pares de difusion a 550 grados C U-Mo-Zr/Al y U-Mo-Zr/Al-A356

    Energy Technology Data Exchange (ETDEWEB)

    Komar Varela, Carolina; Arico, Sergio; Mirandou, Marcela; Balart, Silvia; Gribaudo, Luis [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Materiales; com, carolinakomar@gmail

    2007-07-01

    Out-of-pile diffusion experiments were performed between U-7 wt.% Mo-1 wt.% Zr and Al or Al A356 (7,1 wt.% Si) at 550 C degrees. In this work morphological characterization and phase identification on both interaction layers are presented. They were carried out by the use of different techniques: optical and scanning electron microscopy, X-ray diffraction and WDS microanalysis. In the interaction layer U-7 wt.% Mo-1 wt.% Zr/Al, the phases UAl{sub 3}, UAl{sub 4}, Al{sub 20}Mo{sub 2}U and Al{sub 43}Mo{sub 4}U{sub 6} were identified. Similar results in the interaction layer of the U-7 % Mo/Al at 580 C degrees were previously obtained. In the interaction layer U-7 wt.% Mo-1 wt.% Zr/Al A356, the phases U(Al,Si){sub 3} with 25 at.% Si and Si{sub 5}U{sub 3} were identified. This last phase, with a higher Si concentration, was identified with X-ray diffraction synchrotron radiation performed at the National Synchrotron Light Laboratory, Campinas, Brazil. (author) [Spanish] Se realizaron experiencias fuera de reactor en pares de difusion quimica U-7 % Mo-1 % Zr/Al y U-7 % Mo-1 % Zr/Al A356. En este trabajo se presentan los resultados de la caracterizacion morfologica e identificacion de fases presentes en la zona de interaccion que se forma al ser sometidos a un tratamiento isotermico de 1,5 h a 550 grados C. Las tecnicas utilizadas fueron: microscopia optica y electronica de barrido, difraccion de rayos X y microanalisis cuantitativo por sonda electronica. En la zona de interaccion correspondiente al par U-7 % Mo-1 % Zr/Al se identificaron las fases UAl{sub 3}, UAl{sub 4}, Al{sub 20}Mo{sub 2}U y Al{sub 43}Mo{sub 4}U{sub 6}. Estas cuatro fases fueron identificadas en pares U-7 % Mo/Al a 580 grados C en trabajos anteriores. En la zona de interaccion correspondiente al par U-7 % Mo-1 % Zr/Al A356 se identificaron las fases U(Al,Si){sub 3} (con una concentracion de 25 %at.Si) y Si{sub 5}U{sub 3}. Este compuesto rico en Si solo pudo ser identificado mediante la utilizacion de

  13. The fission cross sections of 230Th, 232Th, 233U, 234U, 236U, 238U, 237Np, 239Pu and 242Pu relative 235U at 14.74 MeV neutron energy

    International Nuclear Information System (INIS)

    Meadows, J.W.

    1986-12-01

    The measurement of the fission cross section ratios of nine isotopes relative to 235 U at an average neutron energy of 14.74 MeV is described with particular attention to the determination of corrections and to sources of error. The results are compared to ENDF/B-V and to other measurements of the past decade. The ratio of the neutron induced fission cross section for these isotopes to the fission cross section for 235 U are: 230 Th - 0.290 +- 1.9%; 232 Th - 0.191 +- 1.9%; 233 U - 1.132 +- 0.7%; 234 U - 0.998 +- 1.0%; 236 U - 0.791 +- 1.1%; 238 U - 0.587 +- 1.1%; 237 Np - 1.060 +- 1.4%; 239 Pu - 1.152 +- 1.1%; 242 Pu - 0.967 +- 1.0%. 40 refs., 11 tabs., 9 figs

  14. Thermal conductivities of (ZrxPu(1-x)/2Am(1-x)/2)N solid solutions

    International Nuclear Information System (INIS)

    Nishi, Tsuyoshi; Takano, Masahide; Akabori, Mitsuo; Arai, Yasuo

    2011-01-01

    The thermal conductivity of Zr-based transuranium (TRU) nitride solid solutions is important for designing subcritical cores in nitride-fueled ADS. Some results have been reported concerning the thermal conductivities of (Zr,Pu)N. However, there have been no experimental data on the thermal conductivities of Zr-based nitride solid solutions containing MA. In this study, the authors prepared sintered samples of (Zr x Pu (1-x)/2 Am (1-x)/2) N (x=0.0, 0.58, 0.80) solid solutions. The thermal diffusivity and heat capacity of (Zr x Pu (1-x)/2 Am (1-x)/2) N solid solutions were measured using a laser flash method and drop calorimetry, respectively. Thermal conductivities were determined from the measured thermal diffusivities, heat capacities and bulk densities over a temperature range of 473 to 1473 K. The thermal conductivities of (Zr 0.58 Pu 0.21 Am 0.21 )N and (Zr 0.80 Pu 0.10 Am 0.10 )N solid solutions were found to be higher than that of (Pu 0.5 Am 0.5 )N due to the high thermal conductivity of ZrN as the principal component, although they were lower than that of ZrN due to the impurifying effect of the transuranium elements. Thus, the thermal conductivities of (Zr x Pu (1-x)/2 Am (1-x)/2) N solid solutions increased with increasing ZrN concentration. Moreover, in order to help to promote the design study of nitride-fueled ADS, the thermal conductivity of the (Zr x Pu (1-x)/2 Am (1-x)/2) N solid solutions were fitted to an equation using the least squares method. (author)

  15. Recent improvements in modelling fission gas release and rod deformation on metallic fuel in LMR

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, Byoung-Oon; Kim, Young Jin

    2000-01-01

    Metallic fuel design is a key feature to assure LMR core safety goals. To date, a large effort has been devoted to the development of the MACSIS code for metallic fuel rod design and the evaluation of operational limits under irradiation conditions. The updated models of fission gas release, fuel core swelling, and rod deformation are incorporated into the correspondence routines in MACSIS MOD1. The MACSIS MOD1 which is a new version of MACSIS, has been partly benchmarked on FGR, fuel swelling and rod deformation comparing with the results of U-Zr and U-Pu-Zr metal fuels irradiated in LMRs. The MACSIS MOD1 predicts, relatively well, the absolute magnitudes and trends of the gas release and rod deformations depending on burn-up, and it gives better agreement with the experimental data than the previous predictions of MACSIS and the results of the empirical model

  16. Oxidation of U-20 at% Zr alloy in air at 423-1063 K

    International Nuclear Information System (INIS)

    Matsui, Tsuneo; Yamada, Takanobu; Ikai, Yasushi; Naito, Keiji

    1993-01-01

    The oxidation behavior of U 0.80 Zr 0.20 alloy (two-phase mixture of U and UZr 2 below 878 K and single solid solution above 1008 K) was studied by thermogravimetry in the temperature range from 423 to 1063 K in air. During oxidation in the low temperature region (423-503 K), the sample kept its initial shape (a rectangular rod) and the surface of the sample was covered by a black thin adherent UO 2+x oxide layer. On the other hand, by oxidation in the middle temperature region, the sample broke to several pieces of thin plates and blocks, and fine powder at 643-723 K and entirely to fine powder at 775-878 K, all of which were analyzed to be a mixture of U 3 O 8 and ZrO 2 . By oxidation in the high temperature region (1008-1063 K) the sample broke to very fine powder, which consisted of U 3 O 8 and ZrO 2 . Based on the sample shape, the oxide phase identified after oxidation and the slope value of the bilogarithmic plots of the weight gain against time, the oxidation kinetics was analyzed with a paralinear equation in the low temperature region below 503 K and a linear equation in the middle and high temperature regions above 643 K. Oxidation rates of U 0.80 Zr 0.20 (two-phase mixture) in the low and middle temperature regions were smaller than those of uranium metal. A discontinuity in the plot of the linear oxidation rate constant versus reciprocal temperature was found to be present between 723 and 838 K, similarly to the case of uranium metal previously reported. The linear rate constants of single-phase solid solution in the high temperature region above 1008 K seemed to be a little smaller than those estimated by the extrapolation of the values in the middle temperature region. (orig.)

  17. Oxidation of U-20 at% Zr alloy in air at 423 1063 K

    Science.gov (United States)

    Matsui, Tsuneo; Yamada, Takanobu; Ikai, Yasushi; Naito, Keiji

    1993-01-01

    The oxidation behavior of U 0.80Zr 0.20 alloy (two-phase mixture of U and UZr 2 below 878 K and single solid solution above 1008 K) was studied by thermogravimetry in the temperature range from 423 to 1063 K in air. During oxidation in the low temperature region (423-503 K), the sample kept its initial shape (a rectangular rod) and the surface of the sample was covered by a black thin adherent UO2 + x oxide layer. On the other hand, by oxidation in the middle temperature region, the sample broke to several pieces of thin plates and blocks, and fine powder at 643-723 K and entirely to fine powder at 775-878 K, all of which were analyzed to be a mixture of U 3O 8 and ZrO 2. By oxidation in the high temperature region (1008-1063 K) the sample broke to very fine powder, which consisted of U 3O 8 and ZrO 2. Based on the sample shape, the oxide phase identified after oxidation and the slope value of the bilogarithmic plots of the weight gain against time, the oxidation kinetics was analyzed with a paralinear equation in the low temperature region below 503 K and a linear equation in the middle and high temperature regions above 643 K. Oxidation rates of U 0.80Zr 0.20 (two-phase mixture) in the low and middle temperature regions were smaller than those of uranium metal. A discontinuity in the plot of the linear oxidation rate constant versus reciprocal temperature was found to be present between 723 and 838 K, similarly to the case of uranium metal previously reported. The linear rate constants of single-phase solid solution in the high temperature region above 1008 K seemed to be a little smaller than those estimated by the extrapolation of the values in the middle temperature region.

  18. Refining U-Zr-Nb alloys by remelting

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, B.M.; Kniess, C.T.; Riella, H.G., E-mail: bmaguiar@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Ferraz, W.B. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The high density U-Zr-Nb and U-Nb uranium-based alloys can be employed as nuclear fuel in a PWR reactor due to their high density and nuclear properties. These alloys can stabilize the gamma phase, however, according to TTT diagrams, at the working temperature of a PWR reactor, all gamma phase transforms to {alpha}'' phase in a few hours. To avoid this kind of transformation during the nuclear reactor operation, the U-Zr-Nb alloy and U-Nn are used in {alpha}'' phase. The stability of {alpha}'' phase depends on the alloy composition and cooling rate. The alloy homogenization has to be very effective to eliminate precipitates rich in Zr and Nb to avoid changes in the alloying elements contents in the matrix. The homogenization was obtained by remelting the alloy and keeping it in the liquid state for enough time to promote floating of the precipitates (usually carbides, less dense) and leaving the matrix free of precipitates. However, this floating by density difference may result in segregation between the alloying elements (Nb and Zr, at the top) and uranium (at the bottom). The homogenized alloys were characterized in terms of metallographic techniques, optical microscopy, scanning electronic microscopy, EDS and X-ray diffraction. In this paper, it is shown that the contents of Zr and Nb at the bottom and at the top of the matrix are constant. (author)

  19. Steady-state fission gas behavior in uranium-plutonium-zirconium metal fuel elements

    International Nuclear Information System (INIS)

    Steele, W.G.; Wazzan, A.R.; Okrent, D.

    1989-01-01

    An analysis of fission gas release and induced swelling in steady state irradiated U-Pu-Zr metal fuels is developed and computer coded. The code is used to simulate, with fair success, some gas release and induced swelling data obtained under the IFR program. It is determined that fuel microstructural changes resulting from zirconium migration, anisotropic swelling, and thermal variations are major factors affecting swelling and gas release behavior. (orig.)

  20. Comparison of the ENDF/B-V and SOKRATOR evaluations of 235U, 239Pu, 240Pu and 241Pu at low neutron energies

    International Nuclear Information System (INIS)

    de Saussure, G.; Wright, R.Q.

    1981-01-01

    The US and USSR's most recent evaluationsof 235 U, 239 Pu, 240 Pu and 241 Pu are compared over the thermal region and over the first few resonances. The two evaluations rest on essentially the same experimental data base and the differences reflect different approaches to the representation of the cross sections or different weightings of the experimental results. It is found that over the thermal and resolved ranges the two evaluations are very similar. Some differences in approaches are briefly discussed

  1. Evaluation of U-Zr hydride fuel for a thorium fuel cycle in an RTR concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Taek; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1999-12-31

    In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-ZrH{sub 1.6}), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTEM 80+and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-ZrH{sub 1.6} fuel in the seed region without additional penalties in comparison with the standard RTR`s U-Zr fuel. 6 refs., 2 figs., 6 tabs. (Author)

  2. Evaluation of U-Zr hydride fuel for a thorium fuel cycle in an RTR concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Taek; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-ZrH{sub 1.6}), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTEM 80+and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-ZrH{sub 1.6} fuel in the seed region without additional penalties in comparison with the standard RTR`s U-Zr fuel. 6 refs., 2 figs., 6 tabs. (Author)

  3. Recycling U and Pu in LWR

    International Nuclear Information System (INIS)

    Zheng Hualing.

    1986-01-01

    This article, from viewpoints of technical feasibility, safety evaluation and socioeconomic benefit-risk analysis, introduces and comments on history and status of recycling U and Pu in LWR, dealing with reactor, reprocessing, conversion and fuel element fabrication et al. Author has analysed LWR fuel cycle strategies in China and made a proposal

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

    Energy Technology Data Exchange (ETDEWEB)

    Hedberg, Marcus, E-mail: marhed@chalmers.se; Ekberg, Christian

    2016-12-15

    Fabrication of plutonium containing fuels through the internal gelation method has mostly been studied in mixed metal systems such as U, Pu or Zr,Pu. In this work production of undiluted PuN has been performed by carbothermal reduction on both oxide powder and Pu microspheres produced by the internal gelation method. Nitride purities reached using the different methods have been studied together with final densities achieved during pellet fabrication as well as losses of ingrown Am during the different production steps. Formation of Pu microspheres was successfully performed using the internal gelation method, although extensive microsphere fracturing occurred during thermal treatment. Final densities of PuN pellets produced by cold pressing and sintering reached 70–80% of theoretical density. Am losses during the carbothermal reduction step was on average about 3.7%. After sintering about 11% of Am was lost in total through the entire production process if sintering in N{sub 2} + 5% H{sub 2} atmosphere while about 50% of the Am in total was lost when using Ar as sintering atmosphere. - Highlights: • Internal gelation Pu based sols was performed. • Nitride formation by carbothermal reduction on Pu microspheres and powders was performed. • Pelletization and sintering of pellets was performed. • Am losses were measured throughout the production steps.

  5. A multi-component Zr alloy with comparable strength and Higher plasticity than Zr-based bulk metallic glasses

    International Nuclear Information System (INIS)

    Liang, S.X.; Yin, L.X.; Ma, M.Z.; Jing, R.; Yu, P.F.; Zhang, Y.F.; Wang, B.A.; Liu, R.P.

    2013-01-01

    Zirconium (Zr)-based bulk metallic glass possesses the highest potential as a structural material among metallic glasses. Although Zr-based bulk metallic glass exhibits extremely high strength, its potential application has been restricted by a number of issues, such as fragility, small size, difficult fabrication into different shapes and poisonous beryllium content, among others. In this paper, a Zr-based crystal alloy with comparable strength and higher plasticity than Zr-based bulk metallic glass is presented. The proposed Zr-based alloy has a tensile strength greater than 1600 MPa. That value is comparable to the 1500 MPa to 2000 MPa strength of Zr-based bulk metallic glasses (BMGs). The ductility in terms of elongation reached 6.2%; at the same time, the 1400 MPa tensile strength was retained. This phenomenon is not possible for Zr-based BMGs. XRD results show that the proposed ultrahigh-strength Zr-based crystal alloy has two-phase structures: an hcp-structured α phase and a bcc-structured β phase. The forged specimen exhibits a typical basket-weave microstructure, which is characterised by the interlaced plate α phase separated from the β phase matrix. Fine, short bar-shaped α phases precipitated along the original β grain boundary together with ultrafine dot-shaped α phases that presented inside the original β grain when the ageing temperature was between 500 °C and 525 °C. As the ageing temperature increased, the dot-shaped α phase grew into plate shapes, decreasing the material's strength and increasing its plasticity. The ultrafine dot-shaped and short bar-shaped α phases in the original β phase matrix are the main strengthening mechanisms of the ultrahigh-strength Zr-based crystal alloy.

  6. DENSITY-FUNCTIONAL STUDY OF U-Mo AND U-Zr ALLOYS

    Energy Technology Data Exchange (ETDEWEB)

    Landa, A; Soderlind, P; Turchi, P A

    2010-11-01

    Density-functional theory previously used to describe phase equilibria in U-Zr alloys [A. Landa, P. Soederlind, P.E.A. Turchi, J. Alloys Comp. 478 (2009) 103-110] is extended to investigate the ground-state properties of U-Mo solid solutions. We discuss how the heat of formation in both alloys correlates with the charge transfer between the alloy components, and how the specific behavior of the density of states in the vicinity of the Fermi level promotes the stabilization of the U{sub 2}Mo compound. Our calculations prove that, due to the existence of a single {gamma}-phase over the typical fuel operation temperatures, {gamma}-U-Mo alloys should indeed have much lower constituent redistribution than {gamma}-U-Zr alloys for which binodal decomposition causes a high degree of constituent redistribution.

  7. Effect of thermo-mechanical processing on microstructure and mechanical properties of U - Nb - Zr alloys: Part 2 - U - 3 wt % Nb - 9 wt % Zr and U - 9 wt% Nb - 3 wt% Zr

    Science.gov (United States)

    Morais, Nathanael Wagner Sales; Lopes, Denise Adorno; Schön, Cláudio Geraldo

    2018-04-01

    The present work is the second and final part of an extended investigation on Usbnd Nb - Zr alloys. It investigates the effect of mechanical processing routes on microstructure of alloys U - 3 wt % Nb - 9 wt % Zr and U - 9 wt% Nb - 3 wt% Zr, through X-ray diffraction and scanning electron microscopy, completing the investigation, which started with alloy U - 6 wt% Nb - 6 wt% Zr in part 1. Mechanical properties are determined using microhardness and bending tests and correlated with the developed microstructures. The results show that processing sequence, in particular the inclusion of a 1000 °C heat treatment step, affects significantly the microstructure and mechanical properties of these alloys alloy in different ways. Microstructural characterization shows that both alloys present significant volume fraction of precipitates of a body-centered cubic (BCC) γ-Nb-Zr rich phase in addition the uranium-rich matrix. Bending tests show that sample ductility does not correlate necessarily with hardness and that the key factor appears to be the amount of the γ-Nb-Zr precipitates, which controls the matrix microstructure. Samples with a monoclinic α″ cellular microstructure and/or with the tetragonally-distorted BCC phase (γ0), although not strictly ductile, showed the largest allowed strains-before-break and complete elastic recovery of the broken pieces, pointing out to the macroscopic observation of superelasticity.

  8. PEMBUATAN SERBUK U-6Zr DENGAN PENGKAYAAN URANIUM 19,75 % UNTUK BAHAN BAKAR REAKTOR RISET

    Directory of Open Access Journals (Sweden)

    Masrukan Masrukan

    2016-03-01

    eksotermik dengan mengeluarkan panas sebesar ∆H = 9,3449 cal/g. Dari hasil analisis dapat diketahui bahwa paduan U-6Zr tersebut dapat digunakan sebagai bahan bakar pengganti untuk reaktor riset. Kata kunci: Serbuk U-6Zr, pengkayaan U 19,75 %, bahan bakar, reaktor riset. ABSTRACT MAKING OF U-6Zr ALLOY POWDER WITH URANIUM ENRICHMENT OF 19.75 % FOR RESEARCH REACTOR FUEL. Making U-6Zr alloy powder with enrichment of 19.75 % for a research reactor fuel has been done. Making of U-6Zr fuel in order to find new fuels that have a high density to replace the existing fuel U3Si2-Al. The purpose of this experiment was to determine the properties of U-6Zr alloy powder obtained from the hydriding-dehydriding process as a candidate research reactor fuel. Initially was made U-6Zr ingot by melting U and Zr metals using electric arc melting furnace. U-6Zr ingots were found then cut into pieces and put into hydriding-dehydriding equipment that operates at pressures 14,46054 psi and hydriding temperatures of 350 oC. Ingots that out of the hydriding-dehydriding equipment becomes brittle then was crushed so becomes powder. Powder that obtained subjected to the test, including chemical composition testing, density, hydrogen content, phase and thermal properties. The results testing show that chemical composition of some elements such as Al, Ca, Cu, and Ni exceeded the permitted limit, each of which contained elements of 202.21 ppm, 214.05 ppm, 61.25 ppm, and 134.53 ppm. On testing the density obtained that U-6Zr powder density of 13.57 g/cc and the testing of hydrogen residual content obtained hydrogen residual content of 0.16 %. For the testing phase were obtained the αU and δU phases, while in testing of the transformation temperature, there are two peaks, the first peak occurs at a temperature of 274 to 311 °C and a second peak occurs at a temperature of 493 to 527 oC. The first peak occurs endothermic reaction by absorbing heat of 6.23 cal/g but not formed a new phase, while the second

  9. Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

    International Nuclear Information System (INIS)

    Ozdemir, Levent; Acar, Banu Bulut; Zabunoglu, Okan H.

    2011-01-01

    When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of 239 Pu and 241 Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

  10. Biološka razgradnja i mehaničko ponašanje sinteriranih kompakata od Co-Cr legure u prahu dopiranih sa ZrO2 rabljenih u stomatologiji

    OpenAIRE

    Băilă, Diana-Irinel; Mocioiu, Oana-Cătălina; Truşcă, Roxana; Surdu, Adrian; Zaharia, Cătălin; Bunea, Mihaela

    2016-01-01

    Direct Metal Laser Sintering je postupak moderne tehnologije s primjenom u različitim industrijama, uključujući medicinsku industriju. Namjena ovog istraživanja bila je dobiti sinterirane kompakte Co-Cr praha dopiranog sa ZrO2 u cilju poboljšanja biološke aktivnosti implantata te ponašanja in vitro nakon uranjanja u simuliranu biološku tekućinu (SBF) tijekom 21 dana. Co-Cr prašak (ST2724G) rabljen je za izradu jednog referentnog uzorka Co-Cr i 4 kompakta Co-Cr dopirana sa ZrO2 u sljedećim pos...

  11. Resistive switching in ZrO2 based metal-oxide-metal structures

    International Nuclear Information System (INIS)

    Kaerkkaenen, Irina

    2014-01-01

    The goal of this work is a deeper understanding of the influence of the (i) metal-oxide-metal (MOM) layer stacks configuration, (ii) the oxide films microstructure, (iii) and their defect structure on the appearance of different switching modes, i.e. unipolar (UP) and bipolar (BP). The first part deals with the fabrication of ZrO 2 thin films by an industrial compatible atomic layer deposition (ALD) process, the chemical, structural and morphological characterization of the films, the growth of ZrO 2 /TiO 2 bilayers, the integration of the layers into metal-oxide-metal (MOM) devices and the electrical characterization with focus on the RS behavior. In the second part the effect of the device structure, in particular the thickness of the electrochemical active electrode (EAE) and the ZrO 2 film morphology, on the RS switching polarity of Pt/ZrO 2 /(EAE) cells is discussed. ZrO 2 films and ZrO 2 /TiO 2 bilayers were grown by ALD and were carefully structurally and electrically characterized. The ZrO 2 films grown from Zr[N(CH 3 )C 2 H 5 ] 4 (TEMA-Zr) at 240 C were polycrystalline with a mixture of cubic/tetragonal phases. ALD/H 2 O-ZrO 2 films exhibited a random oriented polycrystalline structure, whereas the ALD/O 3 -ZrO 2 films consisted of preferably oriented cubic shaped grains. Pt/ZrO 2 /Ti/Pt structures with a Ti top electrode (TE) thickness of 5 to 20 nm showed unipolar type RS behavior, while by increasing the Ti TE thickness a gradual change of switching polarity from unipolar to bipolar with a completely bipolar type RS behavior for a Ti TE thickness of 40 nm is found. The switching in Pt/ZrO 2 /TiO 2 /Ti/Pt devices was unipolar, comparable to Pt/ZrO 2 /Ti/Pt cells. In contrast, bilayers with the reverse structure, Pt/TiO 2 /ZrO 2 /Ti/Pt, showed non-switching behavior. The effect of the cells stack structure on the polarity of the RS behavior was studied in detail for 20 nm thick ZrO 2 films grown by an ozone based ALD process and integrated into Pt/ZrO 2

  12. The effects of Zr on the interdiffusion between metal fuel and cladding material

    International Nuclear Information System (INIS)

    Lee, J. T.; Joo, K. S.; Lee, Y. W.; Son, D. S.; Kim, H.; Kim, K. M.

    1999-01-01

    The interdiffusion layers of the heat-treated U-X(X=6, 8, 10, 12)wt.%Zr /HT9 diffusion couples at 725 deg C to 735 deg C was investigated in terms of Zr content. The diffusion layer of U-6Zr/HT9 formed at 725 deg C was similar to that at 700 deg C, but eutectic reactions was locally initiated along the interface. It was observed that the incipient eutectic reaction layer consisted of a two-phase U(Fe,Cr) 2 + U-rich(90-96at.%U), U-rich phase, partially decomposed Zr-rich band, a two-phase U 6 Fe + needle-shaped precipitates and Zr-rich band. The activated interdiffusion between U-Zr and HT9, is thought to be due to the eutectic liquid phase which partially dissolved Zr and decomposed Zr-rich band, and eutectic liquid phase resulted in the thick diffusion layer of a two-phase UFe 2 matrix + round-shaped U(Fe,Cr,Mo) precipitates. As Zr interrupts the interdiffusion between U-Zr and HT9 at interface, it was thought that Zr-content had an effect of suppression on eutectic reaction

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

    International Nuclear Information System (INIS)

    Venault, L.

    1998-01-01

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

  14. Microstructural and thermodynamic evaluation of as-cast U-rich U-Zr alloys

    International Nuclear Information System (INIS)

    Basak, Chandrabhanu; Prasad, G.J.; Kamath, H.S.

    2009-01-01

    The present study involves evaluation of microstructures and some basic properties of as-cast uranium rich U-Zr alloys; i.e. uranium alloys containing 2wt%, 5wt%, 7wt% and 10 wt% zirconium. Microstructural evaluation, both optical and SEM, with hardness values are reported. It was shown that a definite correlation exists between the microstructure and the hardness of the alloy. Lattice parameter and densities are determined with the help of XRD analysis. Also the phase transformation mechanism is proposed based on the microstructures and XRD analysis. Thermodynamic analysis coupled with the experimental observation reveals that the lamellar structure found in the as-cast U-rich U-Zr alloys originates from the monotectoid reaction (γ→β + γ'). As Zr concentration increases in the alloy the gamma phase can remain in the metastable state even at lower T. So, with increasing Zr content the monotectoid reaction takes place at lower temperature causing generation of finer lamellae. (author)

  15. 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.

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-02-23

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

  16. Development of supported noble metal catalyst for U(VI) to U(IV) reduction

    International Nuclear Information System (INIS)

    Tyagi, Deepak; Varma, Salil; Bhattacharyya, K.; Tripathi, A.K.; Bharadwaj, S.R.; Jain, V.K.; Sahu, Avinash; Vincent, Tessy; Jagatap, B.N.; Wattal, P.K.

    2015-01-01

    Uranium-plutonium separation is an essential step in the PUREX process employed in spent nuclear fuel reprocessing. This partitioning in the PUREX process is achieved by selective reduction of Pu(IV) to Pu(III) using uranous nitrate as reductant and hydrazine as stabilizer. Currently in our Indian reprocessing plants, the requirement of uranous nitrate is met by electrolytic reduction of uranyl nitrate. This process, however, suffers from a major drawback of incomplete reduction with a maximum conversion of ~ 60%. Catalytic reduction of U(VI) to U(IV) is being considered as one of the promising alternatives to the electro-reduction process due to fast kinetics and near total conversion. Various catalysts involving noble metals like platinum (Adams catalyst, Pt/Al 2 O 3 , Pt/SiO 2 etc.) have been reported for the reduction. Sustained activity and stability of the catalyst under harsh reaction conditions are still the issues that need to be resolved. We present here the results on zirconia supported noble metal catalyst that is developed in BARC for reduction of uranyl nitrate to uranous nitrate. Supported noble metal catalysts with varying metal loadings (0.5 - 2 wt%) were prepared via support precipitation and noble metal impregnation. The green catalysts were reduced either by chemical reduction using hydrazine hydrate or by heating in hydrogen flow or combination of both the steps. These catalysts were characterized by various techniques such as, XRD, SEM, TEM, N 2 adsorption and H 2 chemisorption. Performance of these catalysts was evaluated for U(VI) to U(IV) reduction with uranyl nitrate feed using hydrazine as reductant. The results with the most active catalyst are named as 'BARC-CAT', which was developed in our lab. (author)

  17. Measurements of the neutron-induced fission cross sections of 240Pu and 242Pu relative to 235U

    International Nuclear Information System (INIS)

    Behrens, J.W.; Browne, J.C.; Carlson, G.W.

    1976-01-01

    A continuation is given of the fission-cross-section ratio measurements in progress at the Lawrence Livermore Laboratory. Preliminary results are provided for the 240 Pu/ 235 U and 242 Pu/ 235 U ratios from 0.02 to 30 MeV and 0.1 to 30 MeV, respectively. Using the threshold-cross-section method, the ratios were normalized to the values 1.368 +- 0.030 and 1.116 +- 0.025, respectively, from 1.75 to 4.00 MeV

  18. Results from the characterisation of the Futurix-FTA metal alloy transmutation fuels

    International Nuclear Information System (INIS)

    Rory Kennedy, J.; O'Holleran, Th.; Keiser, D.

    2007-01-01

    Full text of publication follows. Idaho National Laboratory has been developing and irradiation testing a number of fuels and fuel types for actinide transmutation as part of the Advanced Fuel Cycle Initiative (AFCI). Fuel types under consideration include both fertile (fast reactor systems) and fertile-free (accelerator-driven systems) metallic alloys. Most recently, fuel fabrication was completed and the fuel pins shipped to the fast flux Phenix reactor in Marcoule, France for irradiation testing as part of the FUTURIX-FTA experiment: an international experiment involving the USA, France, the European Commission and Japan. The metal alloy fuels for this experiment are the low-fertile U-29Pu-4Am-2Np-30Zr and the non-fertile Pu-12Am-40Zr. The fresh fuels have been fully characterised for chemical composition, phase, microstructure, thermal behaviour and fuel-cladding-chemical-interaction (FCCI). Preliminary FCCI results raised some safety concerns with respect to the formation of low melting phases and cladding degradation, which could preclude a fuel from consideration. Results from diffusion couple experiments between the non-fertile fuel Pu-12Am-40Zr and the ferritic HT9 and 422 stainless steels (SS) used in the AFC experiments in the ATR reactor (USA) compared to the austenitic AIM1 SS used in the FUTURIX-FTA experiments in the Phenix reactor (France) indicate significant inter-diffusion with the AIM1 SS. Up to about a 30-fold increase in the diffusion of iron (and accompanying Ni and Cr) into the fuel at 650 C was observed compared to the 422 SS studies. Comparable studies between the low-fertile U-29Pu-4Am-2Np-30Zr fuel alloy and the AIM1 SS show virtually no inter-diffusion. The Fe (along with small amounts of Ni and Cr) appears as small precipitates in the fuel alloy with only minor concentrations identified in the fuel alloy matrix. These results will be discussed in terms of mechanisms of the inter-diffusion and the difference in behaviour between the

  19. Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

    Energy Technology Data Exchange (ETDEWEB)

    Yacout, A. M. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2016-09-16

    The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of data were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.

  20. Methylbutylmalonamide as an extractant for U(VI), Pu(IV), and Am(III)

    International Nuclear Information System (INIS)

    Nair, G.M.; Prabhu, D.R.; Mahajan, G.R.

    1994-01-01

    The unsymmetrical diamide methylbutylmalonamide has been synthesized and used in the extraction of U(VI), Pu(IV), and Am(III) in benzene medium. The distribution ratio for three cations was found to increase with increasing aqueous nitric acid concentration. U(VI) and Pu(IV) were found to be extracted as disolvates while Am(III) as a trisolvate. The thermodynamic parameters determined by the temperature variation method showed the extraction reactions to be mainly enthalphy-controlled. Am(III) was found to be back-extracted with dilute nitric acid, while Pu(IV) by dilute nitric acid - hydrofluoric acid mixture and U(VI) by dilute sodium carbonate solution. (author) 6 refs.; 3 figs.; 2 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    Venault, L

    1998-07-01

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

  2. Beta decay heat following U-235, U-238 and Pu-239 neutron fission

    Science.gov (United States)

    Li, Shengjie

    1997-09-01

    This is an experimental study of beta-particle decay heat from 235U, 239Pu and 238U aggregate fission products over delay times 0.4-40,000 seconds. The experimental results below 2s for 235U and 239Pu, and below 20s for 238U, are the first such results reported. The experiments were conducted at the UMASS Lowell 5.5-MV Van de Graaff accelerator and 1-MW swimming-pool research reactor. Thermalized neutrons from the 7Li(p,n)7Be reaction induced fission in 238U and 239Pu, and fast neutrons produced in the reactor initiated fission in 238U. A helium-jet/tape-transport system rapidly transferred fission fragments from a fission chamber to a low background counting area. Delay times after fission were selected by varying the tape speed or the position of the spray point relative to the beta spectrometer that employed a thin-scintillator-disk gating technique to separate beta-particles from accompanying gamma-rays. Beta and gamma sources were both used in energy calibration. Based on low-energy(energies 0-10 MeV. Measured beta spectra were unfolded for their energy distributions by the program FERD, and then compared to other measurements and summation calculations based on ENDF/B-VI fission-product data performed on the LANL Cray computer. Measurements of the beta activity as a function of decay time furnished a relative normalization. Results for the beta decay heat are presented and compared with other experimental data and the summation calculations.

  3. Fabrication of metallic fuel for fast breeder reactor

    International Nuclear Information System (INIS)

    Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arbind; Mittal, R.K.; Prasad, R.S.; Mahule, N.; Kumar, Arun; Prasad, G.J.

    2012-01-01

    Natural uranium oxide fuelled PHWRs comprises of first stage of Indian nuclear power programme. Liquid metal fast breeder reactors fuelled by Pu (from PHWR's) form the second stage. A shorter reactor doubling time is essential in order to accelerate the nuclear power growth in India. Metallic fuels are known to provide shorter doubling times, necessitating to be used as driver fuel for fast breeder reactors. One of the fabrication routes for metallic fuels having random grain orientation, is injection casting technique. The technique finds its basis in an elementary physical concept - the possibility of supporting a liquid column within a tube, by the application of a pressure difference across the liquid interface inside and outside the tube. At AFD, BARC a facility has been set-up for injection casting of uranium rods in quartz tube moulds, demoulding of cast rods, end-shearing of rods and an automated inspection system for inspection of fuel rods with respect to mass, length, diameter and diameter variation along the length and internal and external porosities/voids. All the above facilities have been set-up in glove boxes and have successfully been used for fabrication of uranium bearing fuel rods. The facility has been designed for fabrication and inspection of Pu-bearing metallic fuels also, if required

  4. Atomization of Cd in U+Zr matrix after chemical separation using GF-AAS

    International Nuclear Information System (INIS)

    Thulasidas, S.K.; Gupta, Santosh Kumar; Natarajan, V.

    2014-01-01

    Studies on the direct atomization of Cd in U+Zr matrix were carried out and the effect of matrix composition and matrix concentration on the analyte absorbance were investigated. Development of a method using graphite furnace atomic absorption spectrometry (GF-AAS) for determination of Cd is required for FBR fuel (U+20%Zr) materials. It was reported that the absorbance signal for Cd is reduced with matrix, 50% at 20 mg/mL of U and 10 mg/mL of Zr matrix as compared to matrix free solution. To use the method for U+Zr mixed oxide samples, effect of varying composition of Zr in U+Zr mixed matrix was studied. The results indicated that Cd absorbance signal remained unaffected in the range 0-40% Zr in (U+Zr) mixed matrix with 20 mg/mL total matrix. Based on these studies, an analytical method was developed for the direct determination of Cd with 20% Zr in 20 mg/mL of U+Zr solution with optimized experimental parameters. The range of analysis was found to be 0.005-0.1 g/mL for Cd with 20 mg/mL matrix; this leads to detection limits of 0.25 ppm. To meet the specification limits at 0.1 ppm level for Cd, it was necessary to separate the matrix from the sample using solvent extraction method. It was reported that with 30%TBP+70%CCl 4 in 7M HNO 3 , a selective simultaneous extraction of U and Zr into the organic phase can be achieved. In the present studies, same extraction procedure was used with 100 mg U+Zr sample. The effect of U+Zr in raffinate on Cd was also estimated. To validate the method, the extracted aqueous samples were also analyzed by ICP-AES SPECTRO ARCOS SOP technique independently and the results were compared. It was seen that Cd estimation was not affected in the presence of 10-50 μg/mL U+Zr by ICP-AES as well

  5. DFT and two-dimensional correlation analysis methods for evaluating the Pu3+–Pu4+ electronic transition of plutonium-doped zircon

    International Nuclear Information System (INIS)

    Bian, Liang; Dong, Fa-qin; Song, Mian-xin; Dong, Hai-liang; Li, Wei-Min; Duan, Tao; Xu, Jin-bao; Zhang, Xiao-yan

    2015-01-01

    Highlights: • Effect of Pu f-shell electron on the electronic property of zircon is calculated via DFT and 2D-CA techniques. • Reasons of Pu f-shell electron influencing on electronic properties are systematically discussed. • Phase transitions are found at two point 2.8 mol% and 7.5 mol%. - Abstract: Understanding how plutonium (Pu) doping affects the crystalline zircon structure is very important for risk management. However, so far, there have been only a very limited number of reports of the quantitative simulation of the effects of the Pu charge and concentration on the phase transition. In this study, we used density functional theory (DFT), virtual crystal approximation (VCA), and two-dimensional correlation analysis (2D-CA) techniques to calculate the origins of the structural and electronic transitions of Zr 1−c Pu c SiO 4 over a wide range of Pu doping concentrations (c = 0–10 mol%). The calculations indicated that the low-angular-momentum Pu-f xy -shell electron excites an inner-shell O-2s 2 orbital to create an oxygen defect (V O-s ) below c = 2.8 mol%. This oxygen defect then captures a low-angular-momentum Zr-5p 6 5s 2 electron to form an sp hybrid orbital, which exhibits a stable phase structure. When c > 2.8 mol%, each accumulated V O-p defect captures a high-angular-momentum Zr-4d z electron and two Si-p z electrons to create delocalized Si 4+ → Si 2+ charge disproportionation. Therefore, we suggest that the optimal amount of Pu cannot exceed 7.5 mol% because of the formation of a mixture of ZrO 8 polyhedral and SiO 4 tetrahedral phases with the orientation (10-1). This study offers new perspective on the development of highly stable zircon-based solid solution materials

  6. Run-Beyond-Cladding-Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program

    International Nuclear Information System (INIS)

    Batte, G.L.; Hoffman, G.L.

    1990-01-01

    In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab

  7. Importance of resonance parameters of fertile nuclei and of 239Pu isotope for fast power reactors

    International Nuclear Information System (INIS)

    Barre, J.Y.; Khairallah, A.

    1975-01-01

    The importance of resonance parameters of fertile nuclei and of 239 Pu isotope for fast power reactors will be restricted, in this presentation, to mixed oxide-uranium-plutonium fuelled sodium-cooled and uranium-oxide-sodium reflected fast reactors. The power range lies between 200 and 2000 MWe. Among the topics of this specialist meeting, the isotopes to be considered are, primarly 239 Pu then 238 U and 240 Pu. Resonance parameters are mainly used in fast power reactor calculations through the well-known concept of self shielding factors. After a short description of the determination and the use of these self-shielding factors, their sensitivities to resonance parameters are characterized from some specific examples: those sensitivities are small. Then, the main design parameters sensitive to the amplitude of self-shielding factors are considered: critical enrichment, global breeding gain. The relative importance of isotope, reaction rate and energy range are mentionned. In a third part, the Doppler effect, sensitive to the temperature variation of self-shielding factors, is considered in the same way. Finally, it is concluded that the present knowledge of resonance parameters for 238 U, 239 Pu and 240 Pu is sufficient for fast power reactors from a designer point of view [fr

  8. European roe deer antlers as an environmental archive for fallout 236U and 239Pu

    International Nuclear Information System (INIS)

    Froehlich, M.B.; Steier, P.; Wallner, G.; Fifield, L.K.

    2016-01-01

    Anthropogenic 236 U and 239 Pu were measured in European roe deer antlers hunted between 1955 and 1977 which covers and extends beyond the period of intensive nuclear weapons testing (1954–1962). The antlers were hunting trophies, and hence the hunting area, the year of shooting and the approximate age of each animal is given. Uranium and plutonium are known to deposit in skeletal tissue. Since antler histology is similar to bone, both elements were expected in antlers. Furthermore, roe deer shed their antlers annually, and hence antlers may provide a time-resolved environmental archive for fallout radionuclides. The radiochemical procedure is based on a Pu separation step by anion exchange (Dowex 1 × 8) and a subsequent U purification by extraction chromatography using UTEVA ® . The samples were measured by Accelerator Mass Spectrometry at the VERA facility (University of Vienna). In addition to the 236 U and 239 Pu concentrations, the 240 Pu/ 239 Pu isotopic ratios were determined with a mean value of 0.172 ± 0.023 which is in agreement with the ratio of global fallout (∼0.18). Rather high 236 U/ 238 U ratios of the order of 10 −6 were observed. These measured ratios, where the 236 U arises only from global fallout, have implications for the use of the 236 U/ 238 U ratio as a fingerprint for nuclear accidents or releases from nuclear facilities. Our investigations have shown the potential to use antlers as a temporally resolved archive for the uptake of actinides from the environment. - Highlights: • Roe deer antlers were studied as an environmental archive for the retrospective study of fallout isotopes 236 U and 239 Pu. • The rather high 236 U/ 238 U ratios of about 10 −6 suggest 236 U as a fingerprint tool for nuclear material releases. • The 240 Pu/ 239 Pu atom ratio underpins global fallout as the main anthropogenic contributor in antlers.

  9. Analysis and composition of the first U-Pu charge (0,043 per cent of Pu); Analyse et constitution du 1. jeu U-Pu (0,043 pour cent de Pu)

    Energy Technology Data Exchange (ETDEWEB)

    Brunet, J P; Lapparent, D de; Lourme, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    Checking the homogeneity in the content of plutonium of 0,043 per cent Pu-natural uranium alloy slugs has been made by Pu 240 and U 238 spontaneous fissions neutrons counting. The purpose of the test was to select groups of slugs to be correctly associated into fuel rods for critical experiments. General technic for spontaneous fissions counting, then elaboration of data in view of ranking the slugs are described. Results are given for this particular case. (authors) [French] On a effectue un controle d'homogeneite de teneur en plutonium sur des billettes d'alliage 0,043 pour cent Pu-uranium naturel, par comptage des fissions spontanees du plutonium 240 et de l'uranium 238. Le but du controle etait de permettre une association correcte de ces billettes a l'interieur des elements combustibles destines a servir dans des experiences critiques. On indique la methode generale de comptage des fissions spontanees, puis le depouillement des donnees en vue du classement des barreaux. Les resultats pour ce cas particulier sont donnes dans le rapport. (auteurs)

  10. Stability and electronic structure of Zr-based ternary metallic glasses and relevant compounds

    International Nuclear Information System (INIS)

    Hasegawa, M.; Soda, K.; Sato, H.; Suzuki, T.; Taketomi, T.; Takeuchi, T.; Kato, H.; Mizutani, U.

    2007-01-01

    The electronic structure of the Zr-based metallic glasses has been investigated by theoretical and experimental approaches. One approach is band calculations of the Zr 2 Ni (Zr 66.7 Ni 33.3 ) compound to investigate the electronic structure of the Zr 66.7 Ni 33.3 metallic glass (ΔT x = 0 K) of which the local atomic structure is similar to that of the Zr 2 Ni compound. The other is photoemission spectroscopy of the Zr 50 Cu 35 Al 15 bulk metallic glass (BMG) (ΔT x = 69 K). Here ΔT x = T x - T g where T x and T g are crystallization and glass transition temperature, respectively. Both results and previous ones on the Zr 55 Cu 30 Ni 5 Al 10 BMG indicate that there is a pseudogap at the Fermi level in the electronic structure of these Zr-based metallic glasses, independent of the value of the ΔT x . This implies that the pseudogap at the Fermi level is one of the factors that stabilize the glass phase of Zr-based metallic glasses

  11. Disagregation of (U, Pu)O2 fuels in molten sodium nitrate and oxides system

    International Nuclear Information System (INIS)

    Chou, T.S.

    1976-01-01

    An oxidation process based on the use of an alkali-nitrate melt has been considered as a possible head end step for the reprocessing of FBR spent fuels. The total alkali solubility in the nitrate melt was examined. It is influenced by the temperature. At 500 degC the alkali solubility in the sodium nitrate melt is about 17 mol %. Examining solidified mixture of sodium and nitrate or sodium oxides and nitrite by X-ray diffraction has revealed five unknown lattices. NaNO 3 .xNa 2 O 2 is cubic (a=8.71A), NaNO 2 .xNa 2 O 2 is tetragonal (a=5.939A, c=9.997A), NaNO 2 .xNa 2 O is cubic (a=10.586A). The structure of NaNO 3 .xNa 2 O and NaNO 3 .xNaO 2 could not be determined. The solubility of barium and ruthenium was briefly investigated. The reaction (U,Pu)O 2 with the alkaline sodium nitrate melt proceeds along the grain boundaries of the solid solution. Two steps have been recognized. First (U,Pu)O 2 is oxidized to (U,Pu)Osub(2+x) and in a subsequent step (U,Pu)Osub(2+x) reacts with sodium peroxide to form (U,Pu) 2 O 5 .xNa 2 O 2 . Disaggregation efficiency is a function of temperature, alkali concentration and physical properties of the pellets. High temperature and low alkali concentration lead to high efficiency. The structure of the reaction products (U,Pu)O 2 with alkaline NaNO 3 melt was shown to depend mainly on the alkali concentration. As the alkali concentration is lower than 2 mole % (U,Pu) 2 O 5 . Na 2 O 2 is the dominate phase. (U,Pu) 2 O 5 .3Na 2 O 2 corresponds to 6 mole % and over 11 mole % alkali, (U,Pu) 2 O 5 .xNa 2 O 2 becomes the main product. The solubility of the fuel (U,Pu) in the alkali sodium nitrate melt increases with the alkali concentration up to 6000-8000 ppm for uranium and 1200-1700 ppm for plutonium at 500 degC with only 5 mole % alkali. As a result of high losses of fissile material in the salt bath molten salt process must regarded as uneligible for a general head end step in fuel reprocessing. Nevertheless its application can still be

  12. Liquid–liquid extraction of Pu(IV), U(VI) and Am(III) using malonamide in room temperature ionic liquid as diluent

    International Nuclear Information System (INIS)

    Rout, Alok; Venkatesan, K.A.; Srinivasan, T.G.; Vasudeva Rao, P.R.

    2012-01-01

    Highlights: ► Extraction of actinides using malonamide in room temperature ionic liquid. ► High distribution ratios of actinides in room temperature ionic liquid. ► Ion exchange mechanism. ► Stoichiometry of extraction. ► High separation factors of U(VI) and Pu(IV) over Am(III) and fission products. - Abstract: The extraction behavior of U(VI), Pu(IV) and Am(III) from nitric acid medium by a solution of N,N-dimethyl-N,N-dioctyl-2-(2-hexyloxyethyl)malonamide (DMDOHEMA) in the room temperature ionic liquid, 1–butyl–3–methylimidazolium bis(trifluoromethanesulfonyl)imide (C 4 mimNTf 2 ), was studied. The distribution ratio of these actinides in DMDOHEMA/C 4 mimNTf 2 was measured as a function of various parameters such as the concentration of nitric acid, DMDOHEMA, NTf 2 − , alkyl chain length of ionic liquid. The extraction of actinides in the absence of DMDOHEMA was insignificant and the distribution ratio achieved in conjunction with C 4 mimNTf 2 , was remarkable. The separation factor of U(VI) and Pu(IV) achieved with the use of DMDOHEMA, ionic liquid was compared with Am(III) and other fission products. The stoichiometry of the metal-solvate was determined to be 1:2 for U(VI) and Pu(IV) and 1:3 for Am(III).

  13. Source identification of Pu and 236U deposited on Norwegian territories

    OpenAIRE

    Wendel, Cato Christian

    2013-01-01

    Plutonium (Pu) is a predominately anthropogenic element produced during neutron irradiation of U in reactors and nuclear weapon detonations. Pu has been released to the environment during nuclear weapon detonations, nuclear reactor accidents, and in association with reprocessing of spent nuclear fuel. The most important source of Pu in the environment were the 543 atmospheric nuclear detonations conducted worldwide in the period 1945 – 1980 by the former Soviet Union (FSU), USA, United Kingdo...

  14. Irradiation performance of U-Mo-Ti and U-Mo-Zr dispersion fuels in Al-Si matrixes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Hofman, G.L. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B.; Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Ryu, H.J.; Park, J.M.; Yang, J.H. [Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2012-08-15

    Performance of U-7 wt.%Mo with 1 wt.%Ti, 1 wt.%Zr or 2 wt.%Zr, dispersed in an Al-5 wt.%Si alloy matrix, was investigated through irradiation tests in the ATR at INL and HANARO at KAERI. Post-irradiation metallographic features show that the addition of Ti or Zr suppresses interaction layer growth between the U-Mo and the Al-5 wt.%Si matrix. However, higher fission gas swelling was observed in the fuel with Zr addition, while no discernable effect was found in the fuel with Ti addition as compared to U-Mo without the addition. Known to have a destabilizing effect on the {gamma}-phase U-Mo, Zr, either as alloy addition or fission product, is ascribed for the disadvantageous result. Considering its benign effect on fuel swelling, with slight disadvantage from neutron economy point of view, Ti may be a better choice for this purpose.

  15. Experimental Investigations into U/TRU Recovery using a Liquid Cadmium Cathode and Salt Containing High Rare Earth Concentrations

    Energy Technology Data Exchange (ETDEWEB)

    Shelly X. Li; Steven D. Herrmann; Michael F. Simpson

    2009-09-01

    Experimental Investigations into U/TRU Recovery using a Liquid Cadmium Cathode and Salt Containing High Rare Earth Concentrations Shelly X. Li, Steven D. Herrmann, and Michael F. Simpson Pyroprocessing Technology Department Idaho National Laboratory P.O. Box 1625, Idaho Falls, ID 83415 USA Abstract - A series of six bench-scale liquid cadmium cathode (LCC) tests was performed to obtain basic separation data with focus on the behavior of rare earth elements. The electrolyte used for the tests was a mixed salt from the Mk-IV and Mk-V electrorefiners, in which spent metal fuels from Experimental Breeder Reactor-II (EBR-II) had been processed. Rare earth (RE) chlorides, such as NdCl3, CeCl3, LaCl3, PrCl3, SmCl3, and YCl3, were spiked into the salt prior to the first test to create an extreme case for investigating rare earth contamination of the actinides collected by a LCC. For the first two LCC tests, an alloy with the nominal composition of 41U-30Pu-5Am-3Np-20Zr-1RE was loaded into the anode baskets as the feed material. The anode feed material for Runs 3 to 6 was spent ternary fuel (U-19Pu-10Zr). The Pu/U ratio in the salt varied from 0.6 to 1.3. Chemical and radiochemical analytical results confirmed that U and transuranics can be collected into the LCC as a group under the given run conditions. The RE contamination level in the LCC product was up to 6.7 wt% of the total metal collected. The detailed data for partitioning of actinides and REs in the salt and Cd phases are reported in the paper.

  16. Characterization of the interaction layer in diffusion couples U-Mo-Zr/Al and U-Mo-Zr/Al-A356 at 550 C degrees

    International Nuclear Information System (INIS)

    Komar Varela, Carolina; Arico, Sergio; Mirandou, Marcela; Balart, Silvia; Gribaudo, Luis

    2007-01-01

    Out-of-pile diffusion experiments were performed between U-7 wt.% Mo-1 wt.% Zr and Al or Al A356 (7,1 wt.% Si) at 550 C degrees. In this work morphological characterization and phase identification on both interaction layers are presented. They were carried out by the use of different techniques: optical and scanning electron microscopy, X-ray diffraction and WDS microanalysis. In the interaction layer U-7 wt.% Mo-1 wt.% Zr/Al, the phases UAl 3 , UAl 4 , Al 20 Mo 2 U and Al 43 Mo 4 U 6 were identified. Similar results in the interaction layer of the U-7 % Mo/Al at 580 C degrees were previously obtained. In the interaction layer U-7 wt.% Mo-1 wt.% Zr/Al A356, the phases U(Al,Si) 3 with 25 at.% Si and Si 5 U 3 were identified. This last phase, with a higher Si concentration, was identified with X-ray diffraction synchrotron radiation performed at the National Synchrotron Light Laboratory, Campinas, Brazil. (author) [es

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

    Energy Technology Data Exchange (ETDEWEB)

    Verbeke, J. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chapline, G. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nakae, L. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Prasad, M. K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sheets, S. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Snyderman, N. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-01-07

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

  18. Buckling and reaction rate measurements in graphite moderated lattices fuelled with plutonium-uranium oxide clusters at temperatures up to 400 deg. C

    International Nuclear Information System (INIS)

    Carter, D.H.; Gibson, M.; King, D.C.; Marshall, J.; Puckett, B.J.; Richards, A.E.; Wass, T.; Wilson, D.J.

    1965-07-01

    The Report describes a series of experiments carried out in SCORPIO I and II on sub-critical graphite moderated lattices fuelled with 21-rod clusters of PuO 2 /UO 2 fuel. Three fuel batches with nominal plutonium: uranium ratios of 0.25%, 0.8% and 1.2% were investigated at temperatures between 20 deg. C and 400 deg. C. Because of the limited amounts of the three fuels, exponential measurements were made in 2-zone stacks, the outer regions of which were loaded with suitably matched 'reference fuel'. Fine structure distributions in the lattice cell were obtained with manganese and indium foils. Pu239/U235 fission ratios were determined both by fission chambers and by fission-product counting techniques. (author)

  19. Design of a thorium fuelled Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Krishnani, P.D.

    2009-01-01

    Full text: The main objective for development of Advanced Heavy Water Reactor (AHWR) is to demonstrate thorium fuel cycle technologies, along with several other advanced technologies required for next generation reactors, so that these are readily available in time for launching the third stage. The AHWR under design is a 300 MWe vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The fuel consists of (Th-Pu)O 2 and ( 233 ThU)O 2 pins. The fuel cluster is designed to generate maximum energy out of 233 U, which is bred in-situ from thorium and has a slightly negative void coefficient of reactivity, negative fuel temperature coefficient and negative power coefficient. For the AHWR, the well -proven pressure tube technology and online fuelling have been adopted. Core heat removal is by natural circulation of coolant during normal operation and shutdown conditions. Thus, it combines the advantages of light water reactors and PHWRs and removes the disadvantages of PHWRs. It has several passive safety systems for reactor normal operation, decay heat removal, emergency core cooling, confinement of radioactivity etc. The fuel cycle is based on the in-situ conversion of naturally available thorium into fissile 233 U in self sustaining mode. The uranium in the spent fuel will be reprocessed and recycled back into the reactor. The plutonium inventory will be kept a minimum and will come from fuel irradiated in Indian PHWRs. The 233 U required initially can come from the fast reactor programme or it can be produced by specially designing the initial core of AHWR using (Th,Pu)MOX fuel. There will be gradual transition from the initial core which will not contain any 233 U to an equilibrium core, which will have ( 233 U, Th) MOX fuel pins also in a composite cluster. The self sustenance is being achieved by a differential fuel loading of low and a relatively higher Pu in the composite clusters. The AHWR burns the

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

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, Jon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Walters, L. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

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

  1. Biocorrosion Evaluation on a Zr-Cu-Ag-Ti Metallic Glass

    Science.gov (United States)

    Kumar, Shresh; Anwar, Rebin; Ryu, Wookha; Park, E. S.; Vincent, S.

    2018-04-01

    Metallic glasses are in high demand for fabrication of variety of innovative products, in particular surgical and biomedical tools and devices owing to its excellent biocompatible properties. In the present investigation, a novel Zr39.5Cu50.5Ag4Ti6 metallic glass composition was synthesized using melt spinning technique. Potentiodynamic polarization studies were conducted to investigate bio-corrosion behaviour of Zr39.5Cu50.5Ag4Ti6 metallic glass. The test were conducted in various simulated artificial body conditions such as artificial saliva solution, phosphate-buffered saline solution, artificial blood plasma solution, and Hank’s balanced saline solution. The bio-corrosion results of metallic glass were compared with traditional biomaterials. The study aims to provide bio-compatible properties of Zr39.5Cu50.5Ag4Ti6 metallic glass.

  2. Resistive switching in ZrO{sub 2} based metal-oxide-metal structures

    Energy Technology Data Exchange (ETDEWEB)

    Kaerkkaenen, Irina

    2014-07-01

    The goal of this work is a deeper understanding of the influence of the (i) metal-oxide-metal (MOM) layer stacks configuration, (ii) the oxide films microstructure, (iii) and their defect structure on the appearance of different switching modes, i.e. unipolar (UP) and bipolar (BP). The first part deals with the fabrication of ZrO{sub 2} thin films by an industrial compatible atomic layer deposition (ALD) process, the chemical, structural and morphological characterization of the films, the growth of ZrO{sub 2}/TiO{sub 2} bilayers, the integration of the layers into metal-oxide-metal (MOM) devices and the electrical characterization with focus on the RS behavior. In the second part the effect of the device structure, in particular the thickness of the electrochemical active electrode (EAE) and the ZrO{sub 2} film morphology, on the RS switching polarity of Pt/ZrO{sub 2}/(EAE) cells is discussed. ZrO{sub 2} films and ZrO{sub 2}/TiO{sub 2} bilayers were grown by ALD and were carefully structurally and electrically characterized. The ZrO{sub 2} films grown from Zr[N(CH{sub 3})C{sub 2}H{sub 5}]{sub 4} (TEMA-Zr) at 240 C were polycrystalline with a mixture of cubic/tetragonal phases. ALD/H{sub 2}O-ZrO{sub 2} films exhibited a random oriented polycrystalline structure, whereas the ALD/O{sub 3}-ZrO{sub 2} films consisted of preferably oriented cubic shaped grains. Pt/ZrO{sub 2}/Ti/Pt structures with a Ti top electrode (TE) thickness of 5 to 20 nm showed unipolar type RS behavior, while by increasing the Ti TE thickness a gradual change of switching polarity from unipolar to bipolar with a completely bipolar type RS behavior for a Ti TE thickness of 40 nm is found. The switching in Pt/ZrO{sub 2}/TiO{sub 2}/Ti/Pt devices was unipolar, comparable to Pt/ZrO{sub 2}/Ti/Pt cells. In contrast, bilayers with the reverse structure, Pt/TiO{sub 2}/ZrO{sub 2}/Ti/Pt, showed non-switching behavior. The effect of the cells stack structure on the polarity of the RS behavior was studied in

  3. Aerosols generated by 239PU and 233U droplets burning in air

    International Nuclear Information System (INIS)

    Nelson, L.S.; Raabe, O.G.

    1978-01-01

    The inhalation hazards of radioactive aerosols produced by the explosive disruption and subsequent combustion of metallic plutonium in air are not adequately understood. Results of a study to determine whether uranium can be substituted for plutonium in such a situation in which experiments were performed under identical conditions with laser-ignited, single, freely falling droplets of 239 Pu and 233 U are reported. The total amounts of aerosol produced were studied quantitatively as a function of time during the combustion. Also, particle size distributions of selected aerosols were studied with aerodynamic particle separation techniques. Results showed that the ultimate quantity of aerosols, their final particle size distributions, and depositions as a function of time are not identical mainly because of the different vapor pressures of the metals, and the unlike degrees of violence of the explosions of the droplets

  4. The effects of 5f localization on the electronic and magnetic properties of the hexagonal U{sub 3}ZrSb{sub 5}

    Energy Technology Data Exchange (ETDEWEB)

    Merabiha, O.; Seddik, T. [Laboratoire de Physique Quantique et de Modélisation Mathématique, Université de Mascara, Mascara 29000 (Algeria); Khenata, R., E-mail: khenata_rabah@yahoo.fr [Laboratoire de Physique Quantique et de Modélisation Mathématique, Université de Mascara, Mascara 29000 (Algeria); Murtaza, G., E-mail: murtaza@icp.edu.pk [Materials Modeling Lab, Department of Physics, Islamia College University, Peshawar (Pakistan); Bouhemadou, A. [Laboratory for Developing New Materials and their Characterization, Department of Physics, Faculty of Science, University of Setif, 19000 Setif (Algeria); Takagiwa, Y. [Department of Advanced Materials Science, The University of Tokyo (Japan); Bin Omran, S. [Department of Physics and Astronomy, King Saud University, P.O. Box 2455, Riyadh 11451 (Saudi Arabia); Rached, D. [Magnetic Materials Laboratory, Department of Physics, Faculty of Sciences, University of Sidi Bel-Abbes, 22000 Sidi Bel-Abbes (Algeria)

    2014-02-15

    Highlights: • The calculated structural parameters of hexagonal U{sub 3}ZrSb{sub 5} are found in good agreement with the experimental data. • U{sub 3}ZrSb{sub 5} is found to be ferromagnetic material. • Evident effects of the Hubbard U parameter in the uranium “5f” states are noted. • Our band structure calculation show the metallic behavior of this ferromagnetic compound. • The thermodynamic properties are predicted through the quasi-harmonic Debye model. -- Abstract: Structural, magnetic, electronic and thermodynamic properties of the hexagonal U{sub 3}ZrSb{sub 5} are theoretically investigated by using the full potential linearized augmented plane wave plus local orbital’s (FP-LAPW + lo) method. The exchange–correlation potential was treated with the generalized gradient approximation GGA of Wu and Cohen. Moreover, the GGA + U approximation (where U is the Hubbard correlation terms) is employed to treat the f electrons properly. The calculated structural parameters are in good agreement with the experimental data. The magnetic study reveals that U{sub 3}ZrSb{sub 5} is a ferromagnetic material. Furthermore, we present a comparative study between the band structures, electronic structures, total and partial densities of states and local moments calculated within both GGA and GGA + U schemes. Our band structure calculations show the metallic behavior of this ferromagnetic compound. The thermodynamic properties are predicted through the quasi-harmonic Debye model, in which the lattice vibrations are taken into account. The variation of relative change in volume, heat capacities and the Debye temperature with temperature and pressure are successfully achieved.

  5. Development of an analytical instrumentation for determining U, Np and Pu

    International Nuclear Information System (INIS)

    Wu Jizong; Guo Kuisheng; Liu Huanliang

    1995-01-01

    An analytical instrumentation for determining U, Np and Pu in the solution of reprocessing factory are made. The principle of the instrumentation is based on that of flow injection analysis and ion chromatography. The instrumentation is composed of controlling box and working box, the distance between the two boxes is 1.5 m. Important quantity of impurity is permitted. The determination limit is 1 mg/l. The relative standard deviation is better than 5%. The instrumentation can be used in 1AF, 1AP and other many controlling points for determining U, Np and Pu

  6. R-matrix analyses of the 235U and 239Pu neutron cross sections

    International Nuclear Information System (INIS)

    Derrien, H.; de Saussure, G.; Larson, N.M.; Leal, L.C.; Perez, R.B.

    1988-01-01

    The resonance parameter analysis code SAMMY was used to perform consistent resonance analyses of several 235 U and 239 Pu fission and capture cross section and transmission measurements up to 110 eV for 235 U and up to 1 keV for 239 Pu. The method of analysis, the measurement selection and the results are briefly outlined in this paper

  7. Resonance Region Covariance Analysis Method and New Covariance Data for Th-232, U-233, U-235, U-238, and Pu-239

    International Nuclear Information System (INIS)

    Leal, Luiz C.; Arbanas, Goran; Derrien, Herve; Wiarda, Dorothea

    2008-01-01

    Resonance-parameter covariance matrix (RPCM) evaluations in the resolved resonance region were done for 232Th, 233U, 235U, 238U, and 239Pu using the computer code SAMMY. The retroactive approach of the code SAMMY was used to generate the RPCMs for 233U, 235U. RPCMs for 232Th, 238U and 239Pu were generated together with the resonance parameter evaluations. The RPCMs were then converted in the ENDF format using the FILE32 representation. Alternatively, for computer storage reasons, the FILE32 was converted in the FILE33 cross section covariance matrix (CSCM). Both representations were processed using the computer code PUFF-IV. This paper describes the procedures used to generate the RPCM with SAMMY.

  8. Fission Product Yields of 233U, 235U, 238U and 239Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons

    Science.gov (United States)

    Laurec, J.; Adam, A.; de Bruyne, T.; Bauge, E.; Granier, T.; Aupiais, J.; Bersillon, O.; Le Petit, G.; Authier, N.; Casoli, P.

    2010-12-01

    The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for 235U(n,f), 239Pu(n,f) in a thermal spectrum, for 233U(n,f), 235U(n,f), and 239Pu(n,f) reactions in a fission neutron spectrum, and for 233U(n,f), 235U(n,f), 238U(n,f), and 239Pu(n,f) for 14.7 MeV monoenergetic neutrons. Irradiations were performed at the EL3 reactor, at the Caliban and Prospero critical assemblies, and at the Lancelot electrostatic accelerator in CEA-Valduc. Fissions were counted in thin deposits using fission ionization chambers. The number of fission products of each species were measured by gamma spectrometry of co-located thick deposits.

  9. Electron beam welding of the dissimilar Zr-based bulk metallic glass and Ti metal

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jonghyun [Department of Material Science, Kumamoto University, Kumamoto 860-8555 (Japan)], E-mail: joindoc@kumamoto-u.ac.jp; Kawamura, Y. [Department of Material Science, Kumamoto University, Kumamoto 860-8555 (Japan)

    2007-04-15

    We successfully welded 3 mm thick Zr{sub 41}Be{sub 23}Ti{sub 14}Cu{sub 12}Ni{sub 10} bulk metallic glass plate to Ti metal by electron beam welding with a beam irradiated 0.4 mm on the BMG side of the interface. There was no crystallization or defects in the weld because changes in the chemical composition of the weld metal were prevented. Bending showed that the welded sample had a higher strength than the Ti base metal. The interface had a 10 {mu}m thick interdiffusion layer of Zr and Ti.

  10. AMS measurements of global fallout U-236 and Pu in an ombrotrophic peat profile: evidence for their post depositional migration

    Energy Technology Data Exchange (ETDEWEB)

    Quinto, Francesca; Hrnecek, Erich; Krachler, Michael [European Commission Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Shotyk, William [Department of Renewable Resources, University of Alberta, 839 General Services Building, Edmonton, AB (Canada); Steier, Peter; Winkler, Stephan; Golser, Robin [VERA Laboratory, Faculty of Physics, University of Vienna, Waehringer Strasse 17, A-1090 Vienna (Austria)

    2014-07-01

    U-236, Pu-239, Pu-240, Pu-241 and Pu-242 were analysed in an ombrotrophic peat core representing the last 80 years of atmospheric deposition. The determination of these isotopes at femtogram and attogram levels was possible by using ultra-clean laboratory procedures and accelerator mass spectrometry. Since the Pu isotopic composition characteristic for global fallout, as well as anthropogenic U-236, were identified in peat samples pre-dating the period of atmospheric atom bomb testing, migration of Pu and U within the peat profile is clearly indicated. The vertical profile of the U-236/U-238 isotopic ratio represents the first observation of the U-236 bomb peak in a terrestrial environment. Comparing the abundances of the global fallout derived U-236 and Pu-239 along the peat core, the post depositional migration of plutonium exceeds that of uranium. These results highlight, for the first time, the mobility of Pu and U in a peat bog with implications for their migration in other acidic, organic rich environments.

  11. Contribution of the study of a nuclear reactor accident: residual power aspects and thermodynamic of U-UO_2 and UO_2-ZrO_2 systems

    International Nuclear Information System (INIS)

    Baichi, Mehdi

    2001-01-01

    This work is a contribution to the study of early delocalization and fission product releases during the formation of corium coming from a nuclear reactor accident. The first part deals with an analysis of corium cooling. The contribution to the power of each corium element has been calculated with time. The main elements are represented but the addition of Pu, Mo and Nb has been proposed. The last release experimental data taken into account result in a loss of residual power of 25% exclusive of corium between the emergency stop and ten days. The second part deals with the early delocalization observed during Vercors experiments. A critical selection on the U-UO_2 and UO_2-ZrO_2 systems has been carried out. In order to complete the small and inconsistent data, thermodynamic activity measurements have been performed by mass spectrometry. The UO_2 activity on UO_2-ZrO_2 presents a positive deviation from ideality at 2200 K and approximates ideality at 2400 K. All the data have been used for optimizing the systems with Thermo-Calc. This work has allowed to calculate the ternary systems and to define the required approach to analyze the metallic phase and corium oxides densities. (author) [fr

  12. Fabrication of U-10wt.%Zr Fuel slug for SFR by Injection Casting

    International Nuclear Information System (INIS)

    Kim, Jong Hwan; Song, Hoon; Kim, Hyung Tae; Ko, Young Mo; Kim, Ki Hwan; Lee, Chan B.

    2013-01-01

    The fabrication technology of metal fuel has been developed by various methods such as rolling, swaging, wire drawing, and co-extrusion, but each of these methods had process limitations requiring an additional subsequent process, and needing the fabrication equipment is complex, which is not favorable for remote use. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, in the early 1950s, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II (EBR-II) driver and the Fast Flux Test Facility (FFTF) fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, vacuum injection casting suitable for remote operation has been developed to fabricate metallic fuel for an SFR. Vacuum injection casting technique was developed to fabricate metallic fuel for an SFR. The appearance of the fabricated U-10wt.%Zr fuel was generally sound and the internal integrity was found to be satisfactory through gamma-ray radiography. Minimum fuel losses after casting relative to the initial charge amount of U-10wt.%Zr fuel slugs met the proposed goal of less than 0.1% fuel losses during fabrication. Modifications of the current facility system and advanced casting techniques are underway to produce higher quality fuel slugs

  13. Fabrication of U-10wt.%Zr Fuel slug for SFR by Injection Casting

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Hwan; Song, Hoon; Kim, Hyung Tae; Ko, Young Mo; Kim, Ki Hwan; Lee, Chan B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The fabrication technology of metal fuel has been developed by various methods such as rolling, swaging, wire drawing, and co-extrusion, but each of these methods had process limitations requiring an additional subsequent process, and needing the fabrication equipment is complex, which is not favorable for remote use. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, in the early 1950s, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II (EBR-II) driver and the Fast Flux Test Facility (FFTF) fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, vacuum injection casting suitable for remote operation has been developed to fabricate metallic fuel for an SFR. Vacuum injection casting technique was developed to fabricate metallic fuel for an SFR. The appearance of the fabricated U-10wt.%Zr fuel was generally sound and the internal integrity was found to be satisfactory through gamma-ray radiography. Minimum fuel losses after casting relative to the initial charge amount of U-10wt.%Zr fuel slugs met the proposed goal of less than 0.1% fuel losses during fabrication. Modifications of the current facility system and advanced casting techniques are underway to produce higher quality fuel slugs.

  14. Transformation temperatures γU → δUZr2 Y γU →αU in U-Zr-Nb alloys

    International Nuclear Information System (INIS)

    Komar Varela, Carolina L.; Arico, Sergio F.; Gribaudo, Luis M.; Comision Nacional de Energia Atomica, General San Martin

    2009-01-01

    The international RERTR program has as primary objective the development of new fuels for research and test reactors to satisfy the requirement of low enrichment in 235 U (lower than 20 %). It is known that the cubic-phase (γU) has an excellent behaviour under irradiation. In this context, in the Materials Department (GIDAT-GAEN-CNEA) U-Zr-Nb alloys are considered candidates for the development of a high-density monolithic-type nuclear fuel. It is necessary to evaluate the thermodynamic parameters that allow to obtain a range of concentrations of the U-Zr-Nb system in which this phase can be retained in the metastable condition with the required 235 U density. In this work, eight U alloys with concentrations ranging from 13.9 to 43.7 wt.% Zr and from 0 to 6.4 wt.% Nb, were fabricated. Dynamical measurements of electrical resistivity, with a cooling rate of 4 o C/min, were performed and the results were analyzed. Considering this cooling rate, a Nb concentration of at least 17.8 wt. % would inhibit the transformation γU→ δUZr 2 and a concentration of al least 23.3 wt % would inhibit the γU → αU transformation. (author)

  15. Transformation behavior of the γU(Zr,Nb) phase under continuous cooling conditions

    Science.gov (United States)

    Komar Varela, C. L.; Gribaudo, L. M.; González, R. O.; Aricó, S. F.

    2014-10-01

    The selected alloy for designing a high-density monolithic-type nuclear fuel with U-Zr-Nb alloy as meat and Zry-4 as cladding, has to remain in the γU(Zr,Nb) phase during the whole fabrication process. Therefore, it is necessary to define a range of concentrations in which the γU(Zr,Nb) phase does not decompose under the process conditions. In this work, several U alloys with concentrations between 28.2-66.9 at.% Zr and 0-13.3 at.% Nb were fabricated to study the possible transformations of the γU(Zr,Nb) phase under different continuous cooling conditions. The results of the electrical resistivity vs temperature experiments are presented. For a cooling rate of 4 °C/min a linear regression was determined by fitting the starting decomposition temperature as a function of Nb concentration. Under these conditions, a concentration of 45.3 at.% Nb would be enough to avoid any transformation of the γU(Zr,Nb) phase. In experiments that involve higher cooling conditions, it has been determined that this concentration can be halved.

  16. Simultaneous determination of U and Pu isotopes by alpha spectrometry

    International Nuclear Information System (INIS)

    Equillor, H.E.; Campos, J.M.

    2011-01-01

    Determination of actinides by alpha spectrometry is usually carried out after full separation of each of the components of the sample. The procedure presented in this paper permits U and Pu isotopes to be measured together allowing faster sample processing and measurement. The method consists in the extraction with tributyl phosphate of U and Pu isotopes from the rest of the matrix, followed by a cathodic electrodeposition to obtain alpha spectrometry suitable pieces. It can be applied to various environmental samples, such as water, filters and soil (about 0.5 g of solid sample for the described conditions). High-quality electroplated discs are essential for simultaneous processing, so a technique to achieve this requirement is also explained. (authors) [es

  17. First-principles investigation of U doping in ZrO2

    International Nuclear Information System (INIS)

    Huang, H.W.; Pan, Y.; Yu, C.; Yang, J.; Wang, H.; Yi, W.; Peng, J.C.

    2014-01-01

    Highlights: • U-doped system with cubic structure is more stable than the monoclinic structure. • The elastic modulus of cubic structure is higher than the monoclinic structure. • U-doped weakens the resistance to shear deformation of ZrO 2 . • U-doped enhances the resistance to shape deformation for monoclinic structure. -- Abstract: The formation energies, elastic properties, Debye temperature and electronic structure of U-doped ZrO 2 within cubic and monoclinic structures are investigated by first-principles approach. The calculated formation energies show that the U-doped with cubic structure is more stable than that of monoclinic structure. The U-doped weakens the resistance to shear deformation whether cubic structure or monoclinic structure. However, the U-doped decreases the shape deformation of cubic structure in contrast to enhances the shape deformation for monoclinic structure. The Debye temperature of U-doped system is lower than the corresponding ZrO 2 . We found that the calculated electronic structure of these systems are consistent with the variation of formation energies

  18. Electrochemical characterization of Zr-based thin film metallic glass in hydrochloric aqueous solution

    International Nuclear Information System (INIS)

    Chuang, Ching-Yen; Liao, Yi-Chia; Lee, Jyh-Wei; Li, Chia-Lin; Chu, Jinn P.; Duh, Jenq-Gong

    2013-01-01

    Recently thin film metallic glass represents a class of promising engineering materials for structural applications. In this work, the Zr-based thin film metallic glass (TFMG) was fabricated on the Si and AISI 420 substrates using a Zr–Cu–Ni–Al alloy and pure Zr metal targets by a pulsed DC magnetron sputtering system. The chemical compositions, crystalline structures, microstructures and corrosion behavior in hydrochloric (HCl) aqueous solutions of Zr-based TFMGs were investigated. The results showed that the surface morphologies of Zr-based TFMG were very smooth. A compact and dense structure without columnar structure was observed. The amorphous structure of Zr-based TFMG was characterized by the X-ray diffractometer and transmission electron microscopy analyses. After the potentiodynamic polarization test, the better corrosion resistance was achieved for the Zr-based TFMG coated AISI 420 in 1 mM HCl aqueous solution. Based on the surface morphologies and chemical analysis results of the corroded surfaces, the pitting, crevice corrosion and filiform corrosion were found. The corrosion mechanisms of the Zr-based TFMG were discussed in this work. - Highlights: ► Zr-based thin film metallic glass with amorphous structure. ► Better corrosion resistance of Zr-based thin film metallic glass observed. ► Pitting, crevice and filiform corrosion reactions revealed. ► The Cu-rich corrosion products found in the pit. ► Nanowire and flaky corrosion products formed adjacent to the filiform corrosion path

  19. First measurements of (236)U concentrations and (236)U/(239)Pu isotopic ratios in a Southern Hemisphere soil far from nuclear test or reactor sites.

    Science.gov (United States)

    Srncik, M; Tims, S G; De Cesare, M; Fifield, L K

    2014-06-01

    The variation of the (236)U and (239)Pu concentrations as a function of depth has been studied in a soil profile at a site in the Southern Hemisphere well removed from nuclear weapon test sites. Total inventories of (236)U and (239)Pu as well as the (236)U/(239)Pu isotopic ratio were derived. For this investigation a soil core from an undisturbed forest area in the Herbert River catchment (17°30' - 19°S) which is located in north-eastern Queensland (Australia) was chosen. The chemical separation of U and Pu was carried out with a double column which has the advantage of the extraction of both elements from a relatively large soil sample (∼20 g) within a day. The samples were measured by Accelerator Mass Spectrometry using the 14UD pelletron accelerator at the Australian National University. The highest atom concentrations of both (236)U and (239)Pu were found at a depth of 2-3 cm. The (236)U/(239)Pu isotopic ratio in fallout at this site, as deduced from the ratio of the (236)U and (239)Pu inventories, is 0.085 ± 0.003 which is clearly lower than the Northern Hemisphere value of ∼0.2. The (236)U inventory of (8.4 ± 0.3) × 10(11) at/m(2) was more than an order of magnitude lower than values reported for the Northern Hemisphere. The (239)Pu activity concentrations are in excellent agreement with a previous study and the (239+240)Pu inventory was (13.85 ± 0.29) Bq/m(2). The weighted mean (240)Pu/(239)Pu isotopic ratio of 0.142 ± 0.005 is slightly lower than the value for global fallout, but our results are consistent with the average ratio of 0.173 ± 0.027 for the southern equatorial region (0-30°S). Copyright © 2014 Elsevier Ltd. All rights reserved.

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

    Science.gov (United States)

    Tomczuk, Z.; Miller, W.E.

    1994-10-18

    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].

  1. DFT and two-dimensional correlation analysis methods for evaluating the Pu{sup 3+}–Pu{sup 4+} electronic transition of plutonium-doped zircon

    Energy Technology Data Exchange (ETDEWEB)

    Bian, Liang, E-mail: bianliang@ms.xjb.ac.cn [Key Laboratory of Functional Materials and Devices for Special Environments, Chinese Academy of Sciences, Urumqi 830011, Xinjiang (China); Laboratory for Extreme Conditions Matter Properties, South West University of Science and Technology, Mianyang 621010, Sichuan (China); Dong, Fa-qin; Song, Mian-xin [Laboratory for Extreme Conditions Matter Properties, South West University of Science and Technology, Mianyang 621010, Sichuan (China); Dong, Hai-liang [Department of Geology and Environmental Earth Science, Miami University, Oxford, OH 45056 (United States); Li, Wei-Min [Key Laboratory of Functional Materials and Devices for Special Environments, Chinese Academy of Sciences, Urumqi 830011, Xinjiang (China); Duan, Tao; Xu, Jin-bao [Laboratory for Extreme Conditions Matter Properties, South West University of Science and Technology, Mianyang 621010, Sichuan (China); Zhang, Xiao-yan [Key Laboratory of Functional Materials and Devices for Special Environments, Chinese Academy of Sciences, Urumqi 830011, Xinjiang (China); Laboratory for Extreme Conditions Matter Properties, South West University of Science and Technology, Mianyang 621010, Sichuan (China)

    2015-08-30

    Highlights: • Effect of Pu f-shell electron on the electronic property of zircon is calculated via DFT and 2D-CA techniques. • Reasons of Pu f-shell electron influencing on electronic properties are systematically discussed. • Phase transitions are found at two point 2.8 mol% and 7.5 mol%. - Abstract: Understanding how plutonium (Pu) doping affects the crystalline zircon structure is very important for risk management. However, so far, there have been only a very limited number of reports of the quantitative simulation of the effects of the Pu charge and concentration on the phase transition. In this study, we used density functional theory (DFT), virtual crystal approximation (VCA), and two-dimensional correlation analysis (2D-CA) techniques to calculate the origins of the structural and electronic transitions of Zr{sub 1−c}Pu{sub c}SiO{sub 4} over a wide range of Pu doping concentrations (c = 0–10 mol%). The calculations indicated that the low-angular-momentum Pu-f{sub xy}-shell electron excites an inner-shell O-2s{sup 2} orbital to create an oxygen defect (V{sub O-s}) below c = 2.8 mol%. This oxygen defect then captures a low-angular-momentum Zr-5p{sup 6}5s{sup 2} electron to form an sp hybrid orbital, which exhibits a stable phase structure. When c > 2.8 mol%, each accumulated V{sub O-p} defect captures a high-angular-momentum Zr-4d{sub z} electron and two Si-p{sub z} electrons to create delocalized Si{sup 4+} → Si{sup 2+} charge disproportionation. Therefore, we suggest that the optimal amount of Pu cannot exceed 7.5 mol% because of the formation of a mixture of ZrO{sub 8} polyhedral and SiO{sub 4} tetrahedral phases with the orientation (10-1). This study offers new perspective on the development of highly stable zircon-based solid solution materials.

  2. Buckling and reaction rate measurements in graphite moderated lattices fuelled with plutonium-uranium oxide clusters at temperatures up to 400 deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Carter, D H; Gibson, M; King, D C; Marshall, J; Puckett, B J; Richards, A E; Wass, T; Wilson, D J [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1965-07-15

    The Report describes a series of experiments carried out in SCORPIO I and II on sub-critical graphite moderated lattices fuelled with 21-rod clusters of PuO{sub 2}/UO{sub 2} fuel. Three fuel batches with nominal plutonium: uranium ratios of 0.25%, 0.8% and 1.2% were investigated at temperatures between 20 deg. C and 400 deg. C. Because of the limited amounts of the three fuels, exponential measurements were made in 2-zone stacks, the outer regions of which were loaded with suitably matched 'reference fuel'. Fine structure distributions in the lattice cell were obtained with manganese and indium foils. Pu239/U235 fission ratios were determined both by fission chambers and by fission-product counting techniques. (author) 14 refs, 30 figs, 18 tabs

  3. Crystallization behavior of Zr62Al8Ni13Cu17 Metallic Glass

    Directory of Open Access Journals (Sweden)

    Jo Mi Sun

    2017-06-01

    Full Text Available The crystallization behavior has been studied in Zr62Al8Ni13Cu17 metallic glass alloy. The Zr62Al8Ni13Cu17 metallic glass crystallized through two steps. The fcc Zr2Ni phase transformed from the amorphous matrix during first crystallization and then the Zr2Ni and residual amorphous matrix transformed into a mixture of tetragonal Zr2Cu and hexagonal Zr6Al2Ni phases. Johnson-Mehl-Avrami analysis of isothermal transformation data suggested that the formation of crystalline phase is primary crystallization by diffusion-controlled growth.

  4. Evaluation of the U-Pu residual mass from spent fuel assemblies with passive and active neutronic methods

    International Nuclear Information System (INIS)

    Bignan, G.; Martin-Deidier, L.

    1991-01-01

    The interpretation of passive and active neutronic measurements to evaluate the U-Pu residual mass in spent fuel assemblies is presented as follows: passive neutron measurements are well correlated to the plutonium mass, active neutron measurements give information linked to the fissile mass content of the assembly ( 235 U + 239 Pu + 241 Pu) and, using the passive neutron measurement, lead to the 235 U mass content of the assemblies

  5. The evaluation of the mass distribution data for 238U, 239Pu and 242Pu fission

    International Nuclear Information System (INIS)

    Liu Tingjin

    2002-01-01

    The mass distribution data for 238 U at E n =1.5, 5.5, 8.3, 11.3, 14.9, 22.0, 27.5, 50.0, 99.5, 160.0 MeV, E p =20.0, 60.0 MeV 239 Pu at E n =0.17, 7.9, 14.5 MeV and 242 Pu at E n =15.1 MeV were evaluated and recommended based on the main available experimental data up to now. The experimental data were make necessary corrections and their errors were also made necessary adjustments. The problems concerned were discussed

  6. Oxygen potentials, oxygen diffusion coefficients and defect equilibria of nonstoichiometric (U,Pu)O{sub 2±x}

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Masato, E-mail: kato.masato@jaea.go.jp [Fukushima Fuels and Materials Department, Japan Atomic Energy Agency, 4002 Narita-chou, O-arai machi, Ibaraki 311-1919 (Japan); Fast Reactor Fuel Cycle Technology Development Division, Japan Atomic Energy Agency, 4-33 Muramatsu Tokai-mura, Ibaraki 319-1194 (Japan); Watanabe, Masashi [Fukushima Fuels and Materials Department, Japan Atomic Energy Agency, 4002 Narita-chou, O-arai machi, Ibaraki 311-1919 (Japan); Fast Reactor Fuel Cycle Technology Development Division, Japan Atomic Energy Agency, 4-33 Muramatsu Tokai-mura, Ibaraki 319-1194 (Japan); Matsumoto, Taku; Hirooka, Shun; Akashi, Masatoshi [Fast Reactor Fuel Cycle Technology Development Division, Japan Atomic Energy Agency, 4-33 Muramatsu Tokai-mura, Ibaraki 319-1194 (Japan)

    2017-04-15

    Oxygen potential of (U,Pu)O{sub 2±x} was evaluated based on defect chemistry using an updated experimental data set. The relationship between oxygen partial pressure and deviation x in (U,Pu)O{sub 2±x} was analyzed, and equilibrium constants of defect formation were determined as functions of Pu content and temperature. Brouwer's diagrams were constructed using the determined equilibrium constants, and a relational equation to determine O/M ratio was derived as functions of O/M ratio, Pu content and temperature. In addition, relationship between oxygen potential and oxygen diffusion coefficients were described. - Highlights: •Brouwer’s diagrams for (U,Pu)O2 were constructed using the updated oxygen potential experimental data set. •Equilibrium constants of defect formation were determined as functions of Pu content and temperature. •Oxygen potential, oxygen diffusion coefficients, point defect concentration were described as functions of O/M ratio, Pu content and temperature.

  7. GRSIS program to predict fission gas release and swelling behavior of metallic fast reactor fuel

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Lee, Byung Ho; Nam, Cheol; Sohn, Dong Seong

    1999-03-01

    A mechanistic model of fission gas release and swelling for the U-(Pu)-Zr metallic fuel in the fast reactor, GRSIS (Gas Release and Swelling in ISotropic fuel matrix) was developed. Fission gas bubbles are assumed to nucleate isotropically from the gas atoms in the metallic fuel matrix since they can nucleate at both the grain boundaries and the phase boundaries which are randomly distributed inside the grain. Bubbles can grow to larger size by gas diffusion and coalition with other bubbles so that they are classified as three classes depending upon their sizes. When bubble swelling reaches the threshold value, bubbles become interconnected each other to make the open channel to the external free space, that is, the open bubbles and then fission gases inside the interconnected open bubbles are released instantaneously. During the irradiation, fission gases are released through the open bubbles. GRSIS model can take into account the fuel gap closure by fuel bubble swelling. When the fuel gap is closed by fuel swelling, the contact pressure between fuel and cladding in relation to the bubble swelling and temperature is calculated. GRSIS model was validated by comparison with the irradiation test results of U-(Pu)-Zr fuels in ANL as well as the parametric studies of the key variable in the model. (author). 13 refs., 1 tab., 22 figs

  8. GRSIS program to predict fission gas release and swelling behavior of metallic fast reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Lee, Byung Ho; Nam, Cheol; Sohn, Dong Seong

    1999-03-01

    A mechanistic model of fission gas release and swelling for the U-(Pu)-Zr metallic fuel in the fast reactor, GRSIS (Gas Release and Swelling in ISotropic fuel matrix) was developed. Fission gas bubbles are assumed to nucleate isotropically from the gas atoms in the metallic fuel matrix since they can nucleate at both the grain boundaries and the phase boundaries which are randomly distributed inside the grain. Bubbles can grow to larger size by gas diffusion and coalition with other bubbles so that they are classified as three classes depending upon their sizes. When bubble swelling reaches the threshold value, bubbles become interconnected each other to make the open channel to the external free space, that is, the open bubbles and then fission gases inside the interconnected open bubbles are released instantaneously. During the irradiation, fission gases are released through the open bubbles. GRSIS model can take into account the fuel gap closure by fuel bubble swelling. When the fuel gap is closed by fuel swelling, the contact pressure between fuel and cladding in relation to the bubble swelling and temperature is calculated. GRSIS model was validated by comparison with the irradiation test results of U-(Pu)-Zr fuels in ANL as well as the parametric studies of the key variable in the model. (author). 13 refs., 1 tab., 22 figs.

  9. Sequential separation of transuranic elements and fission products from uranium metal ingots in electrolytic reduction process of spent PWR fuels

    International Nuclear Information System (INIS)

    Chang Heon Lee; Kih Soo Joe; Won Ho Kim; Euo Chang Jung; Kwang Yong Jee

    2009-01-01

    A sequential separation procedure has been developed for the determination of transuranic elements and fission products in uranium metal ingot samples from an electrolytic reduction process for a metallization of uranium dioxide to uranium metal in a medium of LiCl-Li 2 O molten salt at 650 deg C. Pu, Np and U were separated using anion-exchange and tri-n-butylphosphate (TBP) extraction chromatography. Cs, Sr, Ba, Ce, Pr, Nd, Sm, Eu, Gd, Zr and Mo were separated in several groups from Am and Cm using TBP and di(2-ethylhexyl)phosphoric acid (HDEHP) extraction chromatography. Effect of Fe, Ni, Cr and Mg, which were corrosion products formed through the process, on the separation of the analytes was investigated in detail. The validity of the separation procedure was evaluated by measuring the recovery of the stable metals and 239 Pu, 237 Np, 241 Am and 244 Cm added to a synthetic uranium metal ingot dissolved solution. (author)

  10. Interdiffusion between U-Zr-Mo and stainless steel cladding

    International Nuclear Information System (INIS)

    Hwang, J. Y.; Lee, B. S.; Lee, J. T.; Kang, Y. H.

    1998-01-01

    Interdiffusion investigations were carried out at 700 deg C for 200 hours for the diffusion couples assembled with the U-Zr-Mo ternary fuel versus austenitic stainless steel D9 and the U-Zr-Mo ternary fuel versus martensitic stainless steel HT9 respectively to investigate the fuel-cladding compatibility. SEM-EDS analysis was utilized to determine the composition and the penetration depths of the reaction layers. In the case of Fuel/D9 couple, (Fe, Cr, Ni) of the cladding elements formed the precipitates with the Zr, Mo and diminished the U concentration upto 800μ length from the fuel side. Composition of the precipitates was varied with the penetrated elements. In Fuel/HT9 couple, reaction layer was smaller than that of D9 couples and was less affected by cladding elements. The eutectic reaction appeared partially in the Fuel/HT9 diffusion couple

  11. Atomic structure and formation of CuZrAl bulk metallic glasses and composites

    International Nuclear Information System (INIS)

    Kaban, I.; Jóvári, P.; Escher, B.; Tran, D.T.; Svensson, G.; Webb, M.A.; Regier, T.Z.; Kokotin, V.; Beuneu, B.; Gemming, T.; Eckert, J.

    2015-01-01

    Graphical abstract: Partial radial distribution functions for Cu 47.5 Zr 47.5 Al 5 metallic glass and relevant crystal structures. - Abstract: Cu 47.5 Zr 47.5 Al 5 metallic glass is studied experimentally by high-energy X-ray diffraction, neutron diffraction with isotopic substitution, electron diffraction and X-ray absorption spectroscopy. The atomic structure of the glass is modeled by reverse Monte-Carlo and molecular dynamics simulations. RMC modeling of seven experimental datasets enabled reliable separation of all partial pair distribution functions for Cu 47.5 Zr 47.5 Al 5 metallic glass. A peculiar structural feature of the ternary alloy is formation of the strong Al–Zr bonds, which are supposed to determine its high viscosity and enhanced bulk glass formation. Analysis of the local atomic order in Cu 47.5 Zr 47.5 Al 5 glass and Cu 10 Zr 7 , CuZr 2 and CuZr B2 crystalline structures elucidates their similarities and differences explaining the phase formation sequence by devitrification of the glass.

  12. Mössbauer spectroscopic studies in U-Fe and U-Fe-Zr alloys

    Energy Technology Data Exchange (ETDEWEB)

    Panda, Alaka; Singh, L. Herojit; Rajagopalan, S.; Govindaraj, R., E-mail: govind@igcar.gov.in; Ramachandran, Renjith; Kalavathi, S.; Amarendra, G. [Materials Science Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2016-05-23

    {sup 57}Fe Mössbauer studies have been carried out in an alloy of U and Fe with atomic percentage in the ratio of 68%:32% in order to understand the local structure and valence of Fe atoms associated with different phases that may get formed. The effect of changes in the hyperfine parameters such as isomer shift and quadrupole splitting at Fe sites due to additional alloying of Zr has been studied in an alloy of U, Fe and Zr in the ratio of 44%:33%:23% respectively with respect to that of the U-Fe alloy chosen in the present study. Possible effect of solute clustering in these systems has been addressed in an analogous alloy of uranium and zirconium using positron lifetime spectroscopy.

  13. Hydrogen Fuelling Stations

    DEFF Research Database (Denmark)

    Rothuizen, Erasmus Damgaard

    . A system consisting of one high pressure storage tank is used to investigate the thermodynamics of fuelling a hydrogen vehicle. The results show that the decisive parameter for how the fuelling proceeds is the pressure loss in the vehicle. The single tank fuelling system is compared to a cascade fuelling......This thesis concerns hydrogen fuelling stations from an overall system perspective. The study investigates thermodynamics and energy consumption of hydrogen fuelling stations for fuelling vehicles for personal transportation. For the study a library concerning the components in a hydrogen fuelling...... station has been developed in Dymola. The models include the fuelling protocol (J2601) for hydrogen vehicles made by Society of Automotive Engineers (SAE) and the thermodynamic property library CoolProp is used for retrieving state point. The components in the hydrogen fuelling library are building up...

  14. Experimental determination of U diffusion in α-Zr

    Energy Technology Data Exchange (ETDEWEB)

    Gordillo, Jorge A. [Gerencia de Materiales, Comisión Nacional de Energía Atómica (CNEA), Avenida General Paz 1499, B1650KNA San Martín, Pcia. de Buenos Aires (Argentina); Perez, Rodolfo A., E-mail: rodperez@cnea.gov.ar [Gerencia de Materiales, Comisión Nacional de Energía Atómica (CNEA), Avenida General Paz 1499, B1650KNA San Martín, Pcia. de Buenos Aires (Argentina); Consejo Nacional de Investigaciones Científicas y Técnicas – CONICET, Avda. Rivadavia 1917, 1033 CABA (Argentina); Instituto Sabato-UNSAM/CNEA, Avda. Gral. Paz 1499, 1650 San Martín (Argentina); Iribarren, Manuel [Gerencia de Materiales, Comisión Nacional de Energía Atómica (CNEA), Avenida General Paz 1499, B1650KNA San Martín, Pcia. de Buenos Aires (Argentina); Instituto Sabato-UNSAM/CNEA, Avda. Gral. Paz 1499, 1650 San Martín (Argentina); Di Lalla, N. [Consejo Nacional de Investigaciones Científicas y Técnicas – CONICET, Avda. Rivadavia 1917, 1033 CABA (Argentina)

    2015-07-15

    Highlights: • Diffusion of U in α-Zr was measured for the first time. • The used technique was α-spectrometry. • An extended temperature range was studied 763–1123 K. • A downward curvature in the Arrhenius plot was observed. • The non-Arrhenius behaviour is similar to self-diffusion one. - Abstract: U bulk diffusion in α-Zr was measured by mean α-spectrometry in the temperature range 763–1123 K (540–850 °C). A deviation from the Arrhenius law consistent in a downward curvature was observed; such anomaly is similar to the self and hetero substitutional diffusion previously measured in α-Zr matrix. The measurements are compatible with the existences of migrating Fe–vacancy complex that could be competitive with a simplest single vacancy mechanism for substitutional diffusers. The possibility that this could be the reason for the non Arrhenius behaviour is discussed.

  15. Atomic transport properties in UO2 and mixed oxides (U,Pu)O2

    International Nuclear Information System (INIS)

    Matzke, H.

    1987-01-01

    Atomic diffusion processes in UO 2 and in the fast-breeder reactor fuel, (U,Pu)O 2 are reviewed. Emphasis is given to the slower-moving species, i.e. U and Pu. Self-diffusion, chemical diffusion, diffusion in a thermal gradient, enhancement of diffusion by radiation and fission and the operative diffusion mechanisms are discussed. The main parameter, besides the temperature, is the oxygen-to-metal ratio (O/M ratio) of the oxide. The experimental results are compared with recent calculations reported elsewhere in this volume. Also treated are effects of the possible lambda-transition at ca.2600 K in UO 2 on high-temperature kinetic processes. The present knowledge on the diffusion and mobility of fission products with emphasis on volatile and gaseous elements, and of other actinides with emphasis on their valence states are treated. Gaps in our knowledge are pointed out and the relevance of the available results for oxide fuel during reactor operation is discussed. Whereas much is known for the as-produced 'virgin' fuel, more results are urgently needed for oxides with higher burn-ups containing a few per cent fission products. Finally, technological applications of the diffusion results are treated. As an example, important savings in cost, energy and time in fuel sintering were recently achieved based on basic studies of diffusion properties of UO 2 . (author)

  16. Estimation of uncertainties in resonance parameters of {sup 56}Fe, {sup 239}Pu, {sup 240}Pu and {sup 238}U

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tsuneo; Shibata, Keiichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-05-01

    Uncertainties have been estimated for the resonance parameters of {sup 56}Fe, {sup 239}Pu, {sup 240}Pu and {sup 238}U contained in JENDL-3.2. Errors of the parameters were determined from the measurements which the evaluation was based on. The estimated errors have been compiled in the MF32 of the ENDF format. The numerical results are given in tables. (author)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  18. Di-hydroxyurea-a Promising Reducing Reagent for the U/Pu split in the PUREX process

    Energy Technology Data Exchange (ETDEWEB)

    Taihong, Yan; Weifang, Zheng; Guoan, Ye; Yu, Zhang; Liang, Xian; Ying, Di; Xiaoyan, Bian [Department of Radiochemistry, China Institute of Atomic Energy - CIAE, Beijing 102413 (China)

    2009-06-15

    In the reprocessing of spent nuclear fuel by the Purex process, the separation of U and Pu is a major stage. This is commonly achieved by a redox process, in which a reducing agent (e.g. U(IV) or (FeII)) and a stabiliser (e.g. N{sub 2}H{sub 4} or NH{sub 2}SO{sub 3}H) are added to reduce extractable Pu{sup 4+} to un-extractable Pu{sup 3+}. The stabiliser prevents the nitrous acid catalysed re-oxidation of Pu(III) back to Pu(IV). One of the key objectives is to reduce both the number of solvent extraction cycles and the waste stream volumes [1]. One option for Advanced Purex flowsheets is to adopt a new salt-free reductant in the U/Pu split. Di-hydroxyurea(DHU)-a new Reducing reagent was synthesized with tri-associated solid phosgene (Bis(trichloromethyl)Carbonate) solved in dioxane and hydroxylamine hydrochloride solved in potassium acetate solution. The Reduction of Pu(IV) by DHU was investigated using UV-Vis spectrophotometer. The reduction back-extraction behavior of Pu(IV) in 30%TBP /OK was firstly investigated under conditions of different temperature, different concentration of DHU and HNO{sub 3} and various phase contract time respectively.The results showed that Pu(IV) in organic phase can be stripped rapidly to aqueous phase by DHU. Simulating the 1B contactor of the Purex process by DHU with nitric acid solution as the stripping agent,the separation factors of uranium/plutonium can reach 2.1 10{sup 4}. This indicates that DHU is a promising salt free agent for uranium/plutonium separation. (authors)

  19. Initial and transition cycle development for KALIMER uranium fueled core

    International Nuclear Information System (INIS)

    Kim, Sang Ji; Kim, Young In; Kim, Young Jin; Park, Chang Kue

    1998-01-01

    An economic and safe equilibrium Uranium metallic fuelled core having been established, strategic loading schemes for initial and transition cycles to early reach target equilibrium cycles are suggested for U-U and U-Pu transition cycles. An iterative method to find initial core enrichment splits is developed. With non-uniform feed enrichments at the initial core adopted, this iterative method shows KALIMER can reach Uranium equilibrium cycles just after 4 reloads, keeping feed enrichment unchanged from cycle 2. Recycling of self-generated Pu is not sufficient to make KALIMER a pure Pu equilibrium core even after 56 reloads. equilibrium cycles are suggested for U-U and U-Pu transition cycles. An iterative method to find initial core enrichment splits is developed. With non-uniform feed enrichments at the initial core adopted, this iterative method shows KALIMER can reach Uranium equilibrium cycles just after 4 reloads, keeping feed enrichment unchanged from cycle 2. Recycling of self-generated Pu is not sufficient to make KALIMER a pure Pu equilibrium core even after 56 reloads

  20. Study of phase equilibria in function of temperature in UO2-PuO2-Pu2O3 system for nuclear ceramics with high plutonium contents

    International Nuclear Information System (INIS)

    Truphemus, Thibaut

    2013-01-01

    In the UO 2 -PuO 2 -Pu 2 O 3 section, a monophasic (U 1-y ,Pu y )O 2-x domain is stable for y≤0,20 at 25 C and up to solid-liquid equilibrium. At higher Pu content, phase equilibria are more unclear with a phase separation process. The main objective of this work consisted in upgrading the representation of this system for 0,15≤y≤0,65 and 25≤T(C)≤1500. At 25 C, a miscibility gap composed by two different (U 1-y ,Pu y )O 2-x phases has been observed for y≤0,45, with one very closed to stoichiometric state (Oxygen/Metal=2) and one other very reduced. For the first time, a triphasic domain has been characterized at higher Pu contents, with two (U 1-y ,Pu y )O 2-x phases near y=0,45 and one (U 1-y ,Pu y ) 2 O 3 phase with a low U content inside. Concerning the study in function of temperature, we have demonstrated that phase separation temperature increase when Pu content grows. Several representations have been established. At 200 C, the representation is closed to that at 25 C. At 400 C, the phase separation have been specified at a lower Pu content than that of literature: y=0,35. At 600 C, our results have clarified the section, until then very unclear, with a phase separation appearing at y=0,60.The microstructural analysis has clearly demonstrated the significant impact of the phase separation on the material. Indeed many cracks have been observed in our samples, and quantity of these defects increases when Pu content grows. (author) [fr

  1. Fission Product Yields for 14 MeV Neutrons on 235U, 238U and 239Pu

    International Nuclear Information System (INIS)

    Mac Innes, M.; Chadwick, M.B.; Kawano, T.

    2011-01-01

    We report cumulative fission product yields (FPY) measured at Los Alamos for 14 MeV neutrons on 235 U, 238 U and 239 Pu. The results are from historical measurements made in the 1950s–1970s, not previously available in the peer reviewed literature, although an early version of the data was reported in the Ford and Norris review. The results are compared with other measurements and with the ENDF/B-VI England and Rider evaluation. Compared to the Laurec (CEA) data and to ENDF/B-VI evaluation, good agreement is seen for 235 U and 238 U, but our FPYs are generally higher for 239 Pu. The reason for the higher plutonium FPYs compared to earlier Los Alamos assessments reported by Ford and Norris is that we update the measured values to use modern nuclear data, and in particular the 14 MeV 239 Pu fission cross section is now known to be 15–20% lower than the value assumed in the 1950s, and therefore our assessed number of fissions in the plutonium sample is correspondingly lower. Our results are in excellent agreement with absolute FPY measurements by Nethaway (1971), although Nethaway later renormalized his data down by 9% having hypothesized that he had a normalization error. The new ENDF/B-VII.1 14 MeV FPY evaluation is in good agreement with our data.

  2. The fission cross sections of /sup 230/Th, /sup 232/Th, /sup 233/U, /sup 234/U, /sup 236/U, /sup 238/U, /sup 237/Np, /sup 239/Pu and /sup 242/Pu relative /sup 235/U at 14. 74 MeV neutron energy

    Energy Technology Data Exchange (ETDEWEB)

    Meadows, J.W.

    1986-12-01

    The measurement of the fission cross section ratios of nine isotopes relative to /sup 235/U at an average neutron energy of 14.74 MeV is described with particular attention to the determination of corrections and to sources of error. The results are compared to ENDF/B-V and to other measurements of the past decade. The ratio of the neutron induced fission cross section for these isotopes to the fission cross section for /sup 235/U are: /sup 230/Th - 0.290 +- 1.9%; /sup 232/Th - 0.191 +- 1.9%; /sup 233/U - 1.132 +- 0.7%; /sup 234/U - 0.998 +- 1.0%; /sup 236/U - 0.791 +- 1.1%; /sup 238/U - 0.587 +- 1.1%; /sup 237/Np - 1.060 +- 1.4%; /sup 239/Pu - 1.152 +- 1.1%; /sup 242/Pu - 0.967 +- 1.0%. 40 refs., 11 tabs., 9 figs.

  3. Calorimetric measurements on plutonium rich (U,Pu)O2 solid solutions

    International Nuclear Information System (INIS)

    Kandan, R.; Babu, R.; Nagarajan, K.; Vasudeva Rao, P.R.

    2008-01-01

    Enthalpy increments of U (1-y) Pu y O 2 solid solutions with y = 0.45, 0.55 and 0.65 were measured using a high-temperature differential calorimeter by employing the method of inverse drop calorimetry in the temperature range 956-1803 K. From the fit equations for the enthalpy increments, other thermodynamic functions such as heat capacity, entropy and Gibbs energy function have been computed in the temperature range 298-1800 K. The results are presented and compared with the data available in the literature. The results indicate that the enthalpies of U (1-y) Pu y O 2 solid solutions with y = 0.45, 0.55 and 0.65 obey the Neumann-Kopp's molar additivity rule

  4. Modeling and preliminary analysis on the temperature profile of the (TRU-Zr)-Zr dispersion fuel rod for HYPER

    International Nuclear Information System (INIS)

    Lee, B. W.; Hwang, W.; Lee, B. S.; Park, W. S.

    2000-01-01

    Either TRU-Zr metal alloy or (TRU-Zr)-Zr dispersion fuel is considered as a blanket fuel for HYPER(Hybrid Power Extraction Reactor). In order to develop the code for dispersion fuel rod performance analysis under steady state condition, the fuel temperature distribution model which is the one of the most important factors in a fuel performance code has been developed in this paper,. This developed model computes the one dimensional radial temperature distribution of a cylindrical fuel rod. The temperature profile results by this model are compared with the temperature distributions of U 3 Si-A1 dispersion fuel and TRU-Zr metal alloy fuel. This model will be installed in performance analysis code for dispersion fuel

  5. Multiscale structural characterizations of mixed U(iv)-An(iii) oxalates (An(iii) = Pu or Am) combining XAS and XRD measurements.

    Science.gov (United States)

    Arab-Chapelet, B; Martin, P M; Costenoble, S; Delahaye, T; Scheinost, A C; Grandjean, S; Abraham, F

    2016-04-28

    Mixed actinide(III,IV) oxalates of the general formula M2.2UAn(C2O4)5·nH2O (An = Pu or Am and M = H3O(+) and N2H5(+)) have been quantitatively precipitated by oxalic precipitation in nitric acid medium (yield >99%). Thorough multiscale structural characterization using XRD and XAS measurements confirmed the existence of mixed actinide oxalate solid solutions. The XANES analysis confirmed that the oxidation states of the metallic cations, tetravalent for uranium and trivalent for plutonium and americium, are maintained during the precipitation step. EXAFS measurements show that the local environments around U(+IV), Pu(+III) and Am(+III) are comparable, and the actinides are surrounded by ten oxygen atoms from five bidentate oxalate anions. The mean metal-oxygen distances obtained by XAS measurements are in agreement with those calculated from XRD lattice parameters.

  6. Formation and reduction behaviors of zirconium oxide compounds in LiCl–Li{sub 2}O melt at 923 K

    Energy Technology Data Exchange (ETDEWEB)

    Sakamura, Yoshiharu, E-mail: sakamura@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (CRIEPI), 2-11-1 Iwadokita, Komae-shi, Tokyo 201-8511 (Japan); Iizuka, Masatoshi [Central Research Institute of Electric Power Industry (CRIEPI), 2-11-1 Iwadokita, Komae-shi, Tokyo 201-8511 (Japan); Kitawaki, Shinichi; Nakayoshi, Akira; Kofuji, Hirohide [International Research Institute for Nuclear Decommissioning (IRID), 2-23-1 Nishi-shimbashi, Minato-ku, Tokyo 105-0003 (Japan); Japan Atomic Energy Agency (JAEA), 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1194 (Japan)

    2015-11-15

    The reduction behaviors of ZrO{sub 2}, Li{sub 2}ZrO{sub 3} and (U,Pu,Zr)O{sub 2} in a LiCl–Li{sub 2}O salt bath at 923 K were investigated. This study was conducted as part of a feasibility study on the pyrochemical treatment of damaged fuel debris generated by severe accidents at light water reactors. It was demonstrated in electrolytic reduction tests that the uranium in synthetic corium specimens of (U,Pu,Zr)O{sub 2} with various ZrO{sub 2} contents could be reduced to the metallic form and that part of the zirconium was converted to Li{sub 2}ZrO{sub 3}. Zirconium metal and Li{sub 2}ZrO{sub 3} were obtained by the reduction of ZrO{sub 2}. The reduction of Li{sub 2}ZrO{sub 3} did not proceed even in LiCl containing no Li{sub 2}O. Moreover, the stable chemical forms of the ZrO{sub 2}–Li{sub 2}O complex oxide were investigated as a function of the Li{sub 2}O concentration in LiCl. ZrO{sub 2} was converted to Li{sub 2}ZrO{sub 3} at a Li{sub 2}O concentration of 0.018 wt%. As the Li{sub 2}O concentration was increased, Li{sub 2}ZrO{sub 3} was converted to Li{sub 6}Zr{sub 2}O{sub 7} and then to Li{sub 8}ZrO{sub 6}. It is suggested that the removal of Li{sub 2}ZrO{sub 3} from the reduction product is a key point in the pyrochemical treatment of corium. - Highlights: • The uranium in (U,Pu,Zr)O{sub 2} could be reduced to the metallic form in LiCl–Li{sub 2}O. • Part of the zirconium was converted to Li{sub 2}ZrO{sub 3} during electrolytic reduction. • Li{sub 6}Zr{sub 2}O{sub 7} and Li{sub 8}ZrO{sub 6} formed at high Li{sub 2}O concentrations in LiCl.

  7. Comparison of potential radiological impacts of 233U and 239Pu fuel cycles

    International Nuclear Information System (INIS)

    Meyer, H.R.; Little, C.A.; Witherspoon, J.P.; Till, J.E.

    1979-01-01

    Nuclear fuel cycles utilizing 233 U are currently the subject of considerable interest in the United States. This paper focuses on the identification of significant differences between the off-site radiological hazards posed by 232 Th/ 233 U (Th/U) and 238 U/ 239 Pu (U/Pu) fuel cycles, and represents a portion of our involvement in the Nonproliferation Alternative Systems Assessment Program (NASAP), to be used in support of the International Fuel Cycle Evaluation (INFCE). The major contributors to radiological dose are likely to be uranium mining and milling (58.5% of total fuel cycle dose), reprocessing (33.9%), and light-water reactor power generation (7.3%). The remainder of the cycle, including enrichment processes, fuel fabrication, transportation, and waste management, contributes only 0.3% to total estimated fuel cycle dose

  8. Mechanical and dynamical behaviors of ZrSi and ZrSi{sub 2} bulk metallic glasses: A molecular dynamics study

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Shin-Pon, E-mail: jushin-pon@mail.nsysu.edu.tw [Department of Mechanical and Electro-Mechanical Engineering, National Sun Yat-sen University, Kaohsiung 804, Taiwan (China); Department of Medicinal and Applied Chemistry, Kaohsiung Medical University, Kaohsiung 807, Taiwan (China); Wu, Tsang-Yu; Liu, Shih-Hao [Department of Mechanical and Electro-Mechanical Engineering, National Sun Yat-sen University, Kaohsiung 804, Taiwan (China)

    2015-03-14

    The mechanical and dynamical properties of ZrSi and ZrSi{sub 2} bulk metallic glasses (BMGs) have been investigated by molecular dynamics simulation. The Honeycutt-Anderson (HA) index analysis indicates that the major indexes in ZrSi and ZrSi{sub 2} bulk metallic glasses are 1551, 1541, and 1431, which refers to the liquid structure. For uniaxial tension, the results show that the ZrSi and ZrSi{sub 2} BMGs are more ductile than their crystal counterparts. The evolution of the distribution of atomic local shear strain clearly shows the initialization of shear transformation zones (STZs), the extension of STZs, and the formation of shear bands along a direction 45° from the tensile direction when the tensile strain gradually increases. The self-diffusion coefficients of ZrSi and ZrSi{sub 2} BMGs at temperatures near their melting points were calculated by the Einstein equation according to the slopes of the MSD profiles at the long-time limit. Because the HA fraction summation of icosahedral-like structures of ZrSi BMG is higher than that of ZrSi{sub 2} BMG, and these local structures are more dense, the self-diffusion coefficients of the total, Zr, and Si atoms of ZrSi{sub 2} BMG are larger than those of ZrSi BMG. This can be attributed to the cage effect, where a denser local structure has a higher possibility of atoms jumping back to form a backflow and then suppress atomic diffusivity. For ZrSi{sub 2} BMG, the self-diffusion coefficient of Si increases with temperature more significantly than does that of Zr, because more open packing rhombohedra structures are formed by the Si-Si pair.

  9. The influence of Zr substitution for Nb on the corrosion behaviors of the Ni-Nb-Zr bulk metallic glasses

    Science.gov (United States)

    Li, DengKe; Zhu, ZhengWang; Zhang, HaiFeng; Wang, AiMin; Hu, ZhuangQi

    2012-12-01

    The influence of Zr content on corrosion behaviors of the Ni61.5Nb38.5- x Zr x ( x=1, 3, 5, 7, 9 at.%) bulk metallic glasses (BMGs) in 1 M HCl aqueous solution was investigated by potentiodynamic polarization measurements and X-ray photo-electron spectroscopy (XPS). It was found that these BMG alloys possess superior corrosion resistance, that is, with large passive region of about 1.5 V and low passive current density (as low as 0.05 Am-2 for Ni61.5Nb31.5Zr7). XPS analysis indicates that the high corrosion resistance is attributed to the formation of Nb- and Zr-enriched surface films formed in the aggressive acid solution. The Zr substitution for Nb effectively reduces the Ni content, particularly the metallic state Ni content in the surface films, which depresses the electrical conduction of the surface films and reduces the passive current density, thus leading to the enhancement of the corrosion resistance of these Ni-Nb-Zr BMGs. These alloys may potentially be useful for engineering applications.

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

    International Nuclear Information System (INIS)

    Freiboth, Cameron J.; Gibbs, Frank E.

    2000-01-01

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

  11. The corrosion properties of Zr-Cr-NM alloy metallic waste form for longterm disposal

    Energy Technology Data Exchange (ETDEWEB)

    Han, Seung Youb; Jang, Seon Ah; Eun, Hee Chul; Choi, Jung Hoon; Lee, Ki Rak; Park, Hwan Seo; Ahn, Do Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    KAERI is conducting research on spent cladding hulls and additive metals to generate a solidifcation host matrix for the noble metal fssion product waste in anode sludge from the electro-refning process to minimize the volume of waste that needs to be disposed of. In this study, alloy compositions Zr-17Cr, Zr-22Cr, and Zr-27Cr were prepared with or without eight noble metals representing fuel waste using induction melting. The microstructures of the resulting alloys were characterized and electrochemical corrosion tests were conducted to evaluate their corrosion characteristics. All the compositions had better corrosion characteristics than other Zr-based alloys that were evaluated for comparison. Analysis of the leach solution after the corrosion test of the Zr-22Cr-8NM specimen indicated that the noble metals were not leached during corrosion under 500 mV imposed voltage, which simulates a highly oxidizing disposal environment. The results of this study confrm that Zr-Cr based compositions will likely serve as chemically stable waste forms.

  12. Advanced chemical quality control techniques for use in the manufacture of (U-Pu) MOX fuels

    International Nuclear Information System (INIS)

    Panakkal, J.P.; Prakash, Amrit

    2010-01-01

    Analytical chemistry plays a very important role for nuclear fuel cycle activities be it fuel fabrication, waste management or reprocessing. Nuclear fuels are selected based on the type of reactor. The nuclear fuel has to conform to various stringent chemical specifications like B, rare earths, H, O/M heavy metal content etc. Selection of technique is very important to determine the true specification. This is important particularly when the analyses has to be performed inside leak tight enclosure. The present paper describes the details of the advanced techniques being developed and used in the manufacture of (U,Pu) MOX fuels. (author)

  13. The interaction of Np(V), Pu(VI) and Tc(VII) with metal in alkaline solutions

    International Nuclear Information System (INIS)

    Silin, V.I.; Kareta, A.V.

    1998-01-01

    The interaction of Np(V), Pu(VI) and Tc(VII) with metal reductants Zn, Cr, Sn and their alloys was investigated in 0.5-4 mol l -1 NaOH solutions in static and dynamic conditions (by filtration of solutions through the column filled with grains of metal). In this paper, it was found that the reduction and succeeding precipitation hydroxides of these elements, on the surface of metal grains from 0.5 to 4 mol l -1 NaOH solutions, gives a decontamination factor (DF) from 1.1 to 67. The best result was achieved for Pu (DF=67) on Cr grains after 2.5 h contact at 60 C with 0.5 mol l -1 NaOH solution containing Pu(VI). Increasing the NaOH concentration, and the addition of chromate ions and complex-forming agents to alkaline solution results in a decrease of the decontamination factor (DF). A better result for Np sorption from 1 mol l -1 NaOH solutions was achieved after longer contact, than for Pu, with Cr and Zn grains. The maximum DF=8.9 was achieved for Tc on a column with Zn grains after filtration with a 3.5 mol l -1 NaOH solution containing Tc(VII). Washing out of Np and Pu, sorbed on the Cr grain surfaces, was achieved using an acid solution (1 mol l -1 HNO 3 ). The technetium was desorbed from metal surface by 10% H 2 O 2 solution. (orig.)

  14. Water reactivity with mixed oxide (U,Pu)O2 surfaces

    International Nuclear Information System (INIS)

    Gaillard, Jeremy

    2013-01-01

    The interaction of water with actinides oxide surfaces remains poorly understood. The adsorption of water on PuO 2 surface and (U,Pu)O 2 surface leads to hydrogen generation through radiolysis but also surface evolution. The study of water interaction with mixed oxide (U,Pu)O 2 and PuO 2 surfaces requires the implementation of non intrusive techniques. The study of the hydration of CeO 2 surface is used to study the effectiveness of different techniques. The results show that the water adsorption leads to the surface evolution through the formation of a hydroxide superficial layer. The reactivity of water on the surface depends on the calcination temperature of the oxide precursor. The thermal treatment of hydrated surfaces can regenerate the surface. The study on CeO 2 hydration emphasizes the relevancies of these techniques in studying the hydration of surfaces. The hydrogen generation through water radiolysis is studied with an experimental methodology based on constant relative humidity in the radiolysis cell. The hydrogen accumulation is linear for the first hours and then tends to a steady state content. A mechanism of hydrogen consumption is proposed to explain the existence of the steady state of hydrogen content. This mechanism enables to explain also the evolution of the oxide surface during hydrogen generation experiments as shown by the evolution of hydrogen accumulation kinetics. The accumulation kinetics depends on the dose rate, specific surface area and the relative humidity but also on the oxide aging. The plutonium percentage appears to be a crucial parameter in hydrogen accumulation kinetics. (author) [fr

  15. Stability study of the γ phase in U-Nb-Zr alloys

    International Nuclear Information System (INIS)

    Arico, S.F; Hermida, J.D; Gribaudo, L.M

    2006-01-01

    The development of new low enrichment nuclear fuels for research and radioisotope production reactors imposes the knowledge of properties and behaviors about a series of alloys which the reducing of U 235 (fissionable) concentration is compensated with a greater density of this element inside the fuel. One of these series is composed by U alloys with different contents of alloying, that allow to retain the body centered cubic structure solid solution recognized as phase α in metastable condition at low temperatures. For the present work 10 U based alloys were manufactured with different concentrations containing up to 43,7 % zirconium weight and up to 7,3 % niobium weight. An arch oven was utilized with argon atmosphere. The identification of the present phases in massive samples from the melting was carried out through X-rays diffraction analysis. The results obtained in this work are compared with others results published since the year 1957. In the samples melted the intermetallic UZr 2 diminishes in quantity with the reduction of the composition of Zr in the alloys. In all of them were identified, besides, Zr 6 Fe 3 O, ZrO 0,35 , α and U 3 O 8 present in quantities reduced. The quantity of the two last phases diminishes at the same time with the content in Zr. The parameter of network of the cubic phase γU in these alloys can be represented for the equation: α=(3,5796 -0,1616.x Nb +0,1155.x Zr )/(1.0306+0,003.x Nb -0,0068.x Zr . The parameter of network of the γ phase was measured. Comparing it measured with the value calculated, for eight alloys, the proposed equation showed a very good adjustment (HC)

  16. Characterization of a glass-bonded ceramic waste form loaded with U and Pu

    International Nuclear Information System (INIS)

    Sinkler, W.; O'Holleran, T. P.; Frank, S. M.; Richmann, M. K.; Johnson, S. G.

    1999-01-01

    This paper presents microscopic characterization of four samples of a ceramic waste form (CWF) developed for disposal of actinide-containing electrorefiner salts. The four samples were prepared to investigate the influence of water content and the Pu:U ratio on CWF microstructure and performance. While the overall phase content is not strongly influenced by either variable, the presence of water in the initial zeolite has a detectable effect on CWF microstructure. It is found to influence the distribution of the major actinide host phase, a (U,Pu)O 2 mixed oxide

  17. Technological advances in (U,Pu)O2 CRO recycling using microwave heating

    International Nuclear Information System (INIS)

    Das, D.K.; Singh, G.; Khot, P.M; Kumar, S.; Mishra, A.K.; Behere, P.G.; Afzal, Mohd; Kumar, Arun

    2014-01-01

    A batch type wet recycling process viz. microwave direct de-nitration and calcination technique (MWDDC) has been developed at Advanced Fuel Fabrication Facility (AFFF), BARC, Tarapur, India. The process was developed for complete and multiple recycling of PFBR clean rejected (U,Pu)O 2 MOX fuel pellets (CRO) up to 30(wt%) of PuO 2 . The complete recycling of CRO containing higher Pu content with conventional dry recycling was difficult to achieve and certain amount of virgin powder is always needed to obtain the required product characteristics. The conditioned co-de-nitrated powder via MWDDC process have more or less similar characteristics to that of virgin powder with respect to particle size, apparent and tap density, surface area. This paper presents an insight into MWDDC process details and recent advancements made for improvement of powder and product characteristics. Low temperature microwave calcination (LTMC) was incorporated to improve the quality of co-de-nitrated powder with regard to volatile impurities and nitrate content. MWDDC powder and pellets were subjected to extensive chemical and physical characterization as per PFBR specification document. MOX pellets were fabricated from virgin and MWDDC powder via powder oxide pelletizing route and characterized. The homogeneity in the MOX pellets fabricated from MWDDC powder was found as good as that of virgin. Industrial microwave heating systems are indigenously developed and have advanced applicator and wave transmission designs to achieve high throughput, precise control of microwave power hence the temperature during the course of the process. It was demonstrated that MWDDC is a novel technique for (U,Pu)O 2 MOX rejects recycling in view of complete and multiple recycling. Key words: (U,Pu)O 2 MOX, CRO, Recycling, MWDDC. (authors)

  18. Fermiology of PuCoGa{sub 5} and of related Pu-115 compounds

    Energy Technology Data Exchange (ETDEWEB)

    Oppeneer, P.M. [Department of Physics, Box 530, Uppsala University, S-751 21 Uppsala (Sweden)], E-mail: peter.oppeneer@fysik.uu.se; Shick, A.B. [Institute of Physics, Academy of Sciences of the Czech Republic, CZ-18221 Prague (Czech Republic); Rusz, J. [Department of Electronic Structures, Faculty of Mathematics and Physics, Charles University, Ke Karlovu 5, CZ-121 16 Prague 2 (Czech Republic); Lebegue, S. [Laboratoire de Cristallographie et de Modelisation des Materiaux Mineraux et Biologiques, CNRS-Universite Henri Poincare, B.P. 239, F-54506 Vandoeuvre-les-Nancy (France); Eriksson, O. [Department of Physics, Box 530, Uppsala University, S-751 21 Uppsala (Sweden)

    2007-10-11

    We report computational investigations of the electronic structures of the superconducting Pu-compounds PuCoGa{sub 5} as well as of the non-superconducting compounds PuFeGa{sub 5} and PuNiGa{sub 5}. To capture the localization behavior of the Pu 5f electrons, we apply two computational approaches, which are both rooted in the density-functional theory: the local spin-density approximation (LSDA) and the around mean field (AMF) LSDA+U approach. The latter is applicable to moderately localized 5f electrons while the former is applicable to delocalized 5f electrons. Our LSDA calculations show that the Fermi surfaces of the three Pu-115 compounds are sensitive to the amount of band filling, i.e., the number of electrons of the 3d element. Precisely at the electron filling corresponding to PuCoGa{sub 5} the Fermi surface has a particularly two-dimensional shape. AMF-LSDA+U calculations (with a Coulomb U of about 3 eV and exchange J of 0.6 eV) lead to a non-magnetic ground state for PuCoGa{sub 5}, in which the 5f states are shifted to a higher binding energy, in better agreement with photoemission data. The Fermi surface of PuCoGa{sub 5} computed with the AMF-LSDA+U approach is nonetheless rather two-dimensional and similar to the LSDA Fermi surface. The AMF-LSDA+U approach with a Coulomb U of {approx}3 eV would thus predict an electronic structure for PuCoGa{sub 5} in accord with several experimental data.

  19. Comparison of results analysis of chemical composition of alloys inside the U-Zr-Nb by XRF and AAS techniques

    International Nuclear Information System (INIS)

    Masrukan; Tri Yulianto; Anwar Muchsin

    2011-01-01

    U-Zr-Nb alloy chemical composition analysis using X Ray Fluorescence (XRF) and Atomic Absorption Spectroscopy (AAS) techniques have been conducted, where U-Zr- Nb alloy was chosen as candidates for new high-density fuel for future research reactors . Composition analysis is necessary because the composition of elements in the fuel will determine the characteristics of fuel during the fabrication process and in the reactor. The use of two kinds of analysis techniques were designed to obtain accurate analysis results. The experiment was conducted to determine the major element composition and impurities in the alloy U-Zr-Nb. First U-Zr-Nb varying alloy composition Nb were respectively 1%, 4%, 7% (U10% Zr1% Nb, U10% Zr4% Nb and U10% 7% Nb) as results of the melting process of measuring the diameter of 120 mm crushed on the surface bottom. Once on the bottom surface is smooth, then analyzed using XRF techniques. To analyze the elements using AAS techniques, alloy U-Zr-Nb cut into 10 mm x 5 mm then dissolved using HF and nitric acid. Solution that occurred were analyzed using AAS technique. From the analysis using the XRF technique is obtained the alloy U-10% Zr-1% Nb, U-10% Zr-4% Nb and Zr-10% U-7% Nb) had a content of each element as follows: U (87.8858%), Zr (2.6097%) and Nb (0.2206%), U (87.8556%), Zr (2.6302%), and Nb (0.6573%); U (84.6334%), Zr (2.5773%), and Nb (1.0940) weight. Results of analysis using AAS techniques on samples obtained third consecutive Zr content of 9.25%, 8.90% and 9.80% while the content of Nb was not detected. Meanwhile, the results of elemental analysis of impurities in all three samples showed that almost all the elements are still qualify as fuel except Zn element. Element Zn at the three samples of each alloys U-10% Zr-1% Nb, U-10% Zr-4% Nb and U-10% Zr-7%Nb is 1.3266%, 3.2756% and 1.0927% weight. It could be concluded that the results of analysis of elemental content and impurities in the alloy U-Nb-Zr using both XRF and AAS visible

  20. Photon-induced Fission Product Yield Measurements on 235U, 238U, and 239Pu

    Science.gov (United States)

    Krishichayan, Fnu; Bhike, M.; Tonchev, A. P.; Tornow, W.

    2015-10-01

    During the past three years, a TUNL-LANL-LLNL collaboration has provided data on the fission product yields (FPYs) from quasi-monoenergetic neutron-induced fission of 235U, 238U, and 239Pu at TUNL in the 0.5 to 15 MeV energy range. Recently, we have extended these experiments to photo-fission. We measured the yields of fission fragments ranging from 85Kr to 147Nd from the photo-fission of 235U, 238U, and 239Pu using 13-MeV mono-energetic photon beams at the HIGS facility at TUNL. First of its kind, this measurement will provide a unique platform to explore the effect of the incoming probe on the FPYs, i.e., photons vs. neutrons. A dual-fission ionization chamber was used to determine the number of fissions in the targets and these samples (along with Au monitor foils) were gamma-ray counted in the low-background counting facility at TUNL. Details of the experimental set-up and results will be presented and compared to the FPYs obtained from neutron-induced fission at the same excitation energy of the compound nucleus. Work supported in part by the NNSA-SSAA Grant No. DE-NA0001838.

  1. Transformation behavior of the γU(Zr,Nb) phase under continuous cooling conditions

    Energy Technology Data Exchange (ETDEWEB)

    Komar Varela, C.L., E-mail: cavarela@cnea.gov.ar [Instituto Sabato, UNSAM-CNEA, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina); Gerencia de Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina); Gribaudo, L.M. [Gerencia de Materiales, GAEN, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina); González, R.O.; Aricó, S.F. [Instituto Sabato, UNSAM-CNEA, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina); Gerencia de Materiales, GAEN, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, B1650KNA San Martín, Buenos Aires (Argentina)

    2014-10-15

    The selected alloy for designing a high-density monolithic-type nuclear fuel with U–Zr–Nb alloy as meat and Zry-4 as cladding, has to remain in the γU(Zr,Nb) phase during the whole fabrication process. Therefore, it is necessary to define a range of concentrations in which the γU(Zr,Nb) phase does not decompose under the process conditions. In this work, several U alloys with concentrations between 28.2–66.9 at.% Zr and 0–13.3 at.% Nb were fabricated to study the possible transformations of the γU(Zr,Nb) phase under different continuous cooling conditions. The results of the electrical resistivity vs temperature experiments are presented. For a cooling rate of 4 °C/min a linear regression was determined by fitting the starting decomposition temperature as a function of Nb concentration. Under these conditions, a concentration of 45.3 at.% Nb would be enough to avoid any transformation of the γU(Zr,Nb) phase. In experiments that involve higher cooling conditions, it has been determined that this concentration can be halved.

  2. Transformation behavior of the γU(Zr,Nb) phase under continuous cooling conditions

    International Nuclear Information System (INIS)

    Komar Varela, C.L.; Gribaudo, L.M.; González, R.O.; Aricó, S.F.

    2014-01-01

    The selected alloy for designing a high-density monolithic-type nuclear fuel with U–Zr–Nb alloy as meat and Zry-4 as cladding, has to remain in the γU(Zr,Nb) phase during the whole fabrication process. Therefore, it is necessary to define a range of concentrations in which the γU(Zr,Nb) phase does not decompose under the process conditions. In this work, several U alloys with concentrations between 28.2–66.9 at.% Zr and 0–13.3 at.% Nb were fabricated to study the possible transformations of the γU(Zr,Nb) phase under different continuous cooling conditions. The results of the electrical resistivity vs temperature experiments are presented. For a cooling rate of 4 °C/min a linear regression was determined by fitting the starting decomposition temperature as a function of Nb concentration. Under these conditions, a concentration of 45.3 at.% Nb would be enough to avoid any transformation of the γU(Zr,Nb) phase. In experiments that involve higher cooling conditions, it has been determined that this concentration can be halved

  3. Characterization of a Pu-bearing zirconolite-rich synroc

    International Nuclear Information System (INIS)

    Buck, E.C.; Ebbinghaus, B.; Bakel, A.J.; Bates, J.K.

    1996-01-01

    A titanate-based ceramic waste form, rich in phases structurally related to zirconolite (CaZrTi 2 O 7 ), is being developed as a possible method for immobilizing excess plutonium from dismantled nuclear weapons. As part of this program, Lawrence Livermore National Laboratory (LLNL) produced several ceramics that were then characterized at Argonne National Laboratory (ANL). The plutonium- loaded ceramic was found to contain a Pu-Gd zirconolite phase but also contained plutonium titanates, Gd-polymignyte, and a series of other phases. In addition, much of the Pu was remained as PuO 2- x . The Pu oxidation state in the zirconolite was determined to be mainly Pu 4+ , although some Pu 3+ was believed to be present

  4. Combined exposure of F344 rats to beryllium metal and 239PuO2 aerosols

    International Nuclear Information System (INIS)

    Finch, G.L.; Carlton, W.W.; Rebar, A.H.; Hahn, F.F.; Hoover, M.D.; Griffith, W.C.; Mewhinney, J.A.; Cuddihy, R.G.

    1994-01-01

    Nuclear weapons industry workers have the potential for inhalation exposures to plutonium (Pu) and other agents, such as beryllium (Be) metal. Inhaled Pu deposited in the lung delivers high linear energy transfer alpha particle radiation and is known to induce pulmonary cancer in laboratory animals. Although the epidemiological evidence implicating Be in the induction of human lung cancer is weak and controversial, various studies in laboratory animals have demonstrated the pulmonary carcinogenicity of Be; Be is currently classified as a suspect human carcinogen in the United States and as a demonstrated human carcinogen by the International Agency for Research on Cancer. The purpose of this study is to investigate the potential interactions between Pu and Be in the production of lung tumors in rats exposed by inhalation to particles of plutonium dioxide ( 239 PuO 2 ), Be metal, or these agents in combination

  5. 236U and 239,240Pu ratios from soils around an Australian nuclear weapons test site

    International Nuclear Information System (INIS)

    Tims, S.G.; Froehlich, M.B.; Fifield, L.K.; Wallner, A.; De Cesare, M.

    2016-01-01

    The isotopes 236 U, 239 Pu and 240 Pu are present in surface soils as a result of global fallout from nuclear weapons tests carried out in the 1950's and 1960's. These isotopes potentially constitute artificial tracers of recent soil erosion and sediment movement. Only Accelerator Mass Spectrometry has the requisite sensitivity to measure all three isotopes at these environmental levels. Coupled with its relatively high throughput capabilities, this makes it feasible to conduct studies of erosion across the geographical extent of the Australian continent. In the Australian context, however, global fallout is not the only source of these isotopes. As part of its weapons development program the United Kingdom carried out a series of atmospheric and surface nuclear weapons tests at Maralinga, South Australia in 1956 and 1957. The tests have made a significant contribution to the Pu isotopic abundances present in the region around Maralinga and out to distances ∼1000 km, and impact on the assessment techniques used in the soil and sediment tracer studies. Quantification of the relative fallout contribution derived from detonations at Maralinga is complicated owing to significant contamination around the test site from numerous nuclear weapons safety trials that were also carried out around the site. We show that 236 U can provide new information on the component of the fallout that is derived from the local nuclear weapons tests, and highlight the potential of 236 U as a new fallout tracer. - Highlights: • Measured 236 U inventories around the Maralinga Test Nuclear weapons test site. • Comparison of 236 U and 239 Pu soil depth profiles at Maralinga. • Differences in 236 U and 239 Pu inventories indicate most Pu fallout is from the safety trials, rather than the weapons tests.

  6. An XRD technique for quantitative phase analysis of Al-U-Zr alloy

    International Nuclear Information System (INIS)

    Khan, K.B.; Kulkarni, N.K.; Jain, G.C.

    2003-01-01

    In several nuclear research reactors all over the world, Al-U alloy is used as fuel. To stabilise less brittle phase UAl 3 in Al-U alloy, a small amount of Zr (1 to 3 wt% ) is added. A rapid, non destructive and simple x-ray diffraction technique has been developed for quantitative phase analysis Al-U-Zr alloy system containing UAl 4 , UAl 3 and Al. (author)

  7. Effects of low central fuelling on density and ion temperature profiles in reversed shear plasmas on JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Takenaga, H; Ide, S; Sakamoto, Y; Fujita, T [Japan Atomic Energy Agency, Naka Ibaraki 311-0193 (Japan)], E-mail: takenaga.hidenobu@jaea.go.jp

    2008-07-15

    Effects of low central fuelling on density and ion temperature profiles have been investigated using negative ion based neutral beam injection and electron cyclotron heating (ECH) in reversed shear plasmas on JT-60U. Strong internal transport barrier (ITB) was maintained in density and ion temperature profiles, when central fuelling was decreased by switching positive ion based neutral beam injection to ECH after the strong ITB formation. Similar density and ion temperature ITBs were formed for the low and high central fuelling cases during the plasma current ramp-up phase. Strong correlation between the density gradient and the ion temperature gradient was observed, indicating that particle transport and ion thermal transport are strongly coupled or the density gradient assists the ion temperature ITB formation through suppression of drift wave instabilities such as ion temperature gradient mode. These results support that the density and ion temperature ITBs can be formed under reactor relevant conditions.

  8. Effects of low central fuelling on density and ion temperature profiles in reversed shear plasmas on JT-60U

    Science.gov (United States)

    Takenaga, H.; Ide, S.; Sakamoto, Y.; Fujita, T.; JT-60 Team

    2008-07-01

    Effects of low central fuelling on density and ion temperature profiles have been investigated using negative ion based neutral beam injection and electron cyclotron heating (ECH) in reversed shear plasmas on JT-60U. Strong internal transport barrier (ITB) was maintained in density and ion temperature profiles, when central fuelling was decreased by switching positive ion based neutral beam injection to ECH after the strong ITB formation. Similar density and ion temperature ITBs were formed for the low and high central fuelling cases during the plasma current ramp-up phase. Strong correlation between the density gradient and the ion temperature gradient was observed, indicating that particle transport and ion thermal transport are strongly coupled or the density gradient assists the ion temperature ITB formation through suppression of drift wave instabilities such as ion temperature gradient mode. These results support that the density and ion temperature ITBs can be formed under reactor relevant conditions.

  9. Report on the Synchrotron Characterization of U-Mo and U-Zr Alloys and the Modeling Results

    Energy Technology Data Exchange (ETDEWEB)

    Okuniewski, Maria A. [Purdue Univ., West Lafayette, IN (United States); Ganapathy, Varsha [Purdue Univ., West Lafayette, IN (United States); Hamilton, Brenden [Purdue Univ., West Lafayette, IN (United States); Cassutt, Paul [Purdue Univ., West Lafayette, IN (United States); Zhang, Fan [Purdue Univ., West Lafayette, IN (United States); Velaquez, Daniel [Illinois Inst. of Technology, Chicago, IL (United States); Seibert, Rachel [Illinois Inst. of Technology, Chicago, IL (United States); Terry, Jeff [Illinois Inst. of Technology, Chicago, IL (United States); Sprouster, David [Brookhaven National Lab. (BNL), Upton, NY (United States); Ecker, Lynne [Brookhaven National Lab. (BNL), Upton, NY (United States); Elbakhshwan, Mohamed [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-09-01

    ABSTRACT Uranium-molybdenum (U-Mo) and uranium-zirconium (U-Zr) are two promising fuel candidates for nuclear transmutation reactors which burn long-lived minor actinides and fission products within fast spectrum reactors. The objectives of this research are centered on understanding the early stages of fuel performance through the examination of the irradiation induced microstructural changes in U-Zr and U-Mo alloys subjected to low neutron fluences. Specimens that were analyzed include those that were previously irradiated in the Advanced Test Reactor at INL. This most recent work has focused on a sub-set of the irradiated specimens, specifically U-Zr and U-Mo alloys that were irradiated to 0.01 dpa at temperatures ranging from (150-800oC). These specimens were analyzed with two types of synchrotron techniques, including X-ray absorption fine structure and X-ray diffraction. These techniques provide non-destructive microstructural analysis, including phase identification and quantitation, lattice parameters, crystallite sizes, as well as bonding, structure, and chemistry. Preliminary research has shown changes in the phase fractions, crystallite sizes, and lattice parameters as a function of irradiation and temperature. Future data analyses will continue to explore these microstructural changes.

  10. Modulated synthesis of zirconium-metal organic framework (Zr-MOF) for hydrogen storage applications

    CSIR Research Space (South Africa)

    Ren, Jianwei

    2014-01-01

    Full Text Available A modulated synthesis of Zr-metal organic framework (Zr-MOF) with improved ease of handling and decreased reaction time is reported to yield highly crystalline Zr-MOF with well-defined octahedral shaped crystals for practical hydrogen storage...

  11. Reactivity feedback components of a homogeneous U10Zr-fueled 900 MWt LMR

    International Nuclear Information System (INIS)

    Meneghetti, D.; Kucera, D.A.

    1988-01-01

    The linear and Doppler feedback components of the regional contributions of the power-reactivity-decrement (PRD) and of the temperature coefficient of reactivity for a 900 MWt homogeneous U10Zr-fueled sodium-cooled reactor are calculated. The PRD components are separated into power dependent and power-to-flow dependent parts. The values of PRD and temperature coefficient components are compared with corresponding quantities calculated for the Experimental Breeder Reactor II. The implications of these comparisons upon inherent safety characteristics of metal-fueled sodium-cooled reactors are discussed

  12. Cumulative damage fraction design approach for LMFBR metallic fuel elements

    International Nuclear Information System (INIS)

    Johnson, D.L.; Einziger, R.E.; Huchman, G.D.

    1979-01-01

    The cumulative damage fraction (CDF) analytical technique is currently being used to analyze the performance of metallic fuel elements for proliferation-resistant LMFBRs. In this technique, the fraction of the total time to rupture of the cladding is calculated as a function of the thermal, stress, and neutronic history. Cladding breach or rupture is implied by CDF = 1. Cladding wastage, caused by interactions with both the fuel and sodium coolant, is assumed to uniformly thin the cladding wall. The irradiation experience of the EBR-II Mark-II driver fuel with solution-annealed Type 316 stainless steel cladding provides an excellent data base for testing the applicability of the CDF technique to metallic fuel. The advanced metal fuels being considered for use in LMFBRs are U-15-Pu-10Zr, Th-20Pu and Th-2OU (compositions are given in weight percent). The two cladding alloys being considered are Type 316 stainless steel and a titanium-stabilized Type 316 stainless steel. Both are in the cold-worked condition. The CDF technique was applied to these fuels and claddings under the assumed steady-state operating conditions

  13. Correlation between local structure and stability of supercooled liquid state in Zr-based metallic glasses

    International Nuclear Information System (INIS)

    Saida, Junji; Imafuku, Muneyuki; Sato, Shigeo; Sanada, Takashi; Matsubara, Eiichiro; Inoue, Akihisa

    2007-01-01

    The correlation between the local structure and stability of supercooled liquid state is investigated in the Zr 70 (Ni, Cu) 30 binary and Zr 70 Al 10 (Ni, Cu) 20 (numbers indicate at.%) ternary metallic glasses. The Zr 70 Ni 30 binary amorphous alloy with a low stability of supercooled liquid state has a tetragonal Zr 2 Ni-like local structure around Ni atom. Meanwhile, the Zr 70 Cu 30 binary metallic glass has a different local structure of tetragonal Zr 2 Cu, where we suggest the icosahedral local structure by the quasicrystallization behavior in addition of a very small amount of noble metals. The effect of Al addition on the local structure in the Zr-Ni alloy is also examined. We have investigated that the dominant local structure changes in the icosahedral-like structure from the tetragonal Zr 2 Ni-like local structure by the Al substitution with Ni accompanying with the significant stabilization of supercooled liquid state. It is concluded that the formation of icosahedral local structure contributes to the enhancement of stability of supercooled liquid state in the Zr-based alloys

  14. The effect of actinides on the microstructural development in a metallic high-level nuclear waste form

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D. D., Jr.; Sinkler, W.; Abraham, D. P.; Richardson, J. W., Jr.; McDeavitt, S. M.

    1999-10-25

    Waste forms to contain material residual from an electrometallurgical treatment of spent nuclear fuel have been developed by Argonne National Laboratory. One of these waste forms contains waste stainless steel (SS), fission products that are noble to the process (e.g., Tc, Ru, Pd, Rh), Zr, and actinides. The baseline composition of this metallic waste form is SS-15wt.% Zr. The metallurgy of this baseline alloy has been well characterized. On the other hand, the effects of actinides on the alloy microstructure are not well understood. As a result, SS-Zr alloys with added U, Pu, and/or Np have been cast and then characterized, using scanning electron microscopy, transmission electron microscopy, and neutron diffraction, to investigate the microstructural development in SS-Zr alloys that contain actinides. Actinides were found to congregate non-uniformally in a Zr(Fe,Cr,Ni){sub 2+x} phase. Apparently, the actinides were contained in varying amounts in the different polytypes (C14, C15, and C36) of the Zr(Fe,Cr,Ni){sub 2+x} phase. Heat treatment of an actinide-containing SS-15 wt.% Zr alloy showed the observed microstructure to be stable.

  15. Exchange of Th, U and Pu on macroporous ion exchange resins

    International Nuclear Information System (INIS)

    Nadkarni, M.N.; Mayankutty, P.C.; Pillai, N.S.; Shinde, S.S.

    1977-01-01

    Absorption of Th, U and Pu on macroporous ion exchangers, Amberlyst 15 (cationic) and Amberlyst A-26 (anionic) were studied in nitric acid solutions and the results were found comparable with those on their microreticular counter parts, Dowex 50x8 and Dowex IX4. With a view to evalute the efficiency of Amberlyst A-26 for the final purification of plutonium from the purex process stream, detailed studies conducted to determine the breakthrough capacity of Pu(IV) from 7.2 M nitric acid, elution by 0.5 M nitric acid and the decontamination factors for uranium and zirconium-95. Because of its faster kinetics, Amberlyst A-26 exhibited a much more efficient elution of Pu(IV) by 0.5 M nitric acid than Dowex IX4. (author)

  16. DIMAC program user's manual

    International Nuclear Information System (INIS)

    Lee, Byoung Oon; Song, Tae Young

    2003-11-01

    DIMAC (A DIspersion Metallic fuel performance Analysis Code) is a computer program for simulating the behavior of dispersion fuel rods under normal operating conditions of HYPER. It computes the one-dimensional temperature distribution and the thermo-mechanical characteristics of fuel rod under the steady state operation condition, including the swelling and rod deformation. DIMAC was developed based on the experience of research reactor fuel. DIMAC consists of the temperature calculation module, the mechanical swelling calculation module, and the fuel deformation calculation module in order to predict the deformation of a dispersion fuel as a function of power history. Because there are a little of available U-TRU-Zr or TRU-Zr characteristics, the material data of U-Pu-Zr or Pu-Zr are used for those of U-TRU-Zr or TRU-Zr. This report is mainly intended as a user's manual for the DIMAC code. The general description on this code, the description on input parameter, the description on each subroutine, the sample problem and the sample input and partial output are written in this report

  17. DIMAC program user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byoung Oon; Song, Tae Young

    2003-11-01

    DIMAC (A DIspersion Metallic fuel performance Analysis Code) is a computer program for simulating the behavior of dispersion fuel rods under normal operating conditions of HYPER. It computes the one-dimensional temperature distribution and the thermo-mechanical characteristics of fuel rod under the steady state operation condition, including the swelling and rod deformation. DIMAC was developed based on the experience of research reactor fuel. DIMAC consists of the temperature calculation module, the mechanical swelling calculation module, and the fuel deformation calculation module in order to predict the deformation of a dispersion fuel as a function of power history. Because there are a little of available U-TRU-Zr or TRU-Zr characteristics, the material data of U-Pu-Zr or Pu-Zr are used for those of U-TRU-Zr or TRU-Zr. This report is mainly intended as a user's manual for the DIMAC code. The general description on this code, the description on input parameter, the description on each subroutine, the sample problem and the sample input and partial output are written in this repo0008.

  18. Metallic Fuels Handbook

    Energy Technology Data Exchange (ETDEWEB)

    Janney, Dawn E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Papesch, Cynthia A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Burkes, Douglas E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cole, James I. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Fielding, Randall S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Frank, Steven M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hartmann, Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hyde, Timothy A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Keiser, Jr., Dennis D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, J. Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States); Maddison, Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mariani, Robert D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Middlemas, Scott C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Holleran, Thomas P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sencer, Bulent H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Squires, Leah N. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-08-07

    This is not a typical External Report--It is a Handbook. No Abstract is involved. This includes both Parts 1 and 2. The Metallic Fuels Handbook summarizes currently available information about phases and phase diagrams, heat capacity, thermal expansion, and thermal conductivity of elements and alloys in the U-Pu-Zr-Np-Am-La-Ce-Pr-Nd system. Although many sections are reviews and updates of material in previous versions of the Handbook [1, 2], this revision is the first to include alloys with four or more elements. In addition to presenting information about materials properties, the handbook attempts to provide information about how well each property is known and how much variation exists between measurements. Although it includes some results from models, its primary focus is experimental data.

  19. Thermodynamic database development: Al-Am-Ga-Pu-U

    Energy Technology Data Exchange (ETDEWEB)

    Perron, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Physical and Life Sciences; Turchi, P. E. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Physical and Life Sciences; Landa, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Physical and Life Sciences; Soderlind, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Physical and Life Sciences

    2014-03-17

    The goal of this study is about the thermodynamic re-assessment of the Plutonium- Uranium (Pu-U) system as a first step leading to the development of a plutonium-based thermodynamic database (i.e., Pu with Al, Am, Ga, Mo, U…) with resulting phase diagrams and associated thermodynamic data. Indeed, phase stability trends and phase diagrams of multi-component nuclear materials are crucial for predicting properties and performance under normal, hypothetical or even accidental conditions. This work is based on a coupling between ab initio energetics, phenomenological thermodynamics models - based on the CALPHAD (CALculation of PHAse Diagrams) approach - and the use of the Thermo-Calc software, together with experimental data (whenever available). The present report summarizes results obtained (quarter period: 10/07/2013-01/07/2014) under the auspices of an agreement between CEA/DAM and NNSA/DP on cooperation in fundamental science supporting stockpile stewardship (P182).

  20. Assessment of the physico-chemical properties of phases in the Na-U-Pu-O system

    International Nuclear Information System (INIS)

    Kleykamp, H.

    1990-05-01

    A critical review is given on the physico-chemical properties of phases in the Na-O, Na-U-O, Na-Pu-O and Na-U-Pu-O systems. This includes the phase diagrams as well as the crystallographic, mechanical, thermal, thermodynamic, transport, optical and chemical properties. This data is to be used for the modelling of the thermal, mechanical and chemical behaviour of defective LMFBR mixed oxide pins during and after reactor operation. (orig.) [de

  1. Calculated critical parameters in simple geometries for oxide and nitrate water mixtures of U-233, U-235 and Pu-239 with thorium. Final report

    International Nuclear Information System (INIS)

    Converse, W.E.; Bierman, S.R.

    1979-11-01

    Calculations have been performed on water mixtures of oxides and nitrates of 233 U, 235 U, and 239 Pu with chemically similar thorium compounds to determine critical dimensions for simple geometries (sphere, cylinder, and slab). Uranium enrichments calculated were 100%, 20%, 10%, and 5%; plutonium calculations assumed 100% 239 Pu. Thorium to uranium or plutonium weight ratios (Th: U or Pu) calculated were 0, 1, 4, and 8. Both bare and full water reflection conditions were calculated. The results of the calculations are plotted showing a critical dimension versus the uranium or plutonium concentration. Plots of K-infinity and material buckling for each material type are also shown

  2. Interdiffusion and reactions between U-Mo and Zr at 650 °C as a function of time

    Science.gov (United States)

    Park, Y.; Keiser, D. D.; Sohn, Y. H.

    2015-01-01

    Development of monolithic U-Mo alloy fuel (typically U-10 wt.%Mo) for the Reduced Enrichment for Research and Test Reactors (RERTR) program entails a use of Zr diffusion barrier to eliminate the interdiffusion-reactions between the fuel alloy and Al-alloy cladding. The application of Zr barrier to the U-Mo fuel system requires a co-rolling process that utilizes a soaking temperature of 650 °C, which represents the highest temperature the fuel system is exposed to during both fuel manufacturing and reactor application. Therefore, in this study, development of phase constituents, microstructure and diffusion kinetics of U-10 wt.%Mo and Zr was examined using solid-to-solid diffusion couples annealed at 650 °C for 240, 480 and 720 h. Phase constituents and microstructural development were analyzed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Concentration profiles were mapped as diffusion paths on the isothermal ternary phase diagram. Within the diffusion zone, single-phase layers of β-Zr and β-U were observed along with a discontinuous layer of Mo2Zr between the β-Zr and β-U layers. In the vicinity of Mo2Zr phase, islands of α-Zr phases were also found. In addition, acicular α-Zr and U6Zr3Mo phases were observed within the γ-U(Mo) terminal alloy. Growth rate of the interdiffusion-reaction zone was determined to be 7.75 (± 5.84) × 10-16 m2/s at 650 °C, however with an assumption of a certain incubation period.

  3. Benchmark Specification for an HTR Fuelled with Reactor-grade Plutonium (or Reactor-grade Pu/Th and U/Th). Proposal version 2

    International Nuclear Information System (INIS)

    Hosking, J.G.; Newton, T.D.; Morris, P.

    2007-01-01

    This benchmark proposal builds upon that specified in NEA/NSC/DOC(2003)22 report. In addition to the three phases described in that report, another two phases have now been defined. Additional items for calculation have also been added to the existing phases. It is intended that further items may be added to the benchmark after consultation with its participants. Although the benchmark is specifically designed to provide inter-comparisons for plutonium- and thorium-containing fuels, it is proposed that phases considering simple calculations for a uranium fuel cell and uranium core be included. The purpose of these is to identify any increased uncertainties, relative to uranium fuel, associated with the lesser-known fuels to be investigated in different phases of this benchmark. The first phase considers an infinite array of fuel pebbles fuelled with uranium fuel. Phase 2 considers a similar array of pebbles but for plutonium fuel. Phase 3 continues the plutonium fuel inter-comparisons within the context of whole core calculations. Calculations for Phase 4 are for a uranium-fuelled core. Phase 5 considers an infinite array of pebbles containing thorium. In setting the benchmark the requirements in the definition of the LEUPRO-12 PROTEUS benchmark have been considered. Participants were invited to submit both deterministic results as well as, where appropriate, results from Monte Carlo calculations. Fundamental nuclear data, Avogadro's number, natural abundance data and atomic weights have been taken from the references indicated in the document

  4. Influence of alkali metal hydroxides on corrosion of Zr-base alloys

    International Nuclear Information System (INIS)

    Jeong, Yong Hwan

    1996-01-01

    The influence of group-1 alkali hydroxides on different Zr-based alloys have been carried out in static autoclaves at 350 deg C in pressurized water, conditioned in low(0.32 mmol), medium(4.3 mmol) and high(31.5 mmol) equimolar concentration of Li-, Na-, K-, Rb- and Cs-hydroxide. Two types of alloys have been investigated: Zr-Sn-(TRM, Transition metal) and Zr-Sn-Nb-(TRM, Transition metal). From the experiments the cation could be identified as the responsible species for corrosion of Zr alloy in alkalized water. The radius of the cation governs the accelerated corrosion in the pre-transition region of Zr alloy. Incorporation of alkali cation into the zirconium oxide lattice is probably the mechanism which allows the corrosion enhancement for Li and Na and the significant lower effect for the other bases. Nb containing alloys showed lower corrosion resistance than Zr-Sn-TRM alloys in all alkali solutions. Both types of alloys were corroded significantly more in LiOH and NaOH than in the other alkali environments. Lowest corrosive aggressiveness has been found for CsOH followed by KOH. Concluding from the corrosion behavior in the different alkali environments and taking into account the tendency to accelerate the corrosion of Zr alloys, CsOH and KOH are possible alternate alkali for PWR (Pressurized Water Reactor) application. (author)

  5. Development of metallic fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Ho; Lee, Chong Yak; Lee, Myung Ho and others

    1999-03-01

    With the vacuum melting and casting of the U-10wt%Zr alloy which is metallic fuel for liquid metal fast breeder reactor, we studied the microstructure of the alloy and the parameters of the melting and casting for the fuel rods. Internal defects of the U-10wt%Zr fuel by gravity casting, were inspected by non-destructive test. U-10wt%Zr alloy has been prepared for the thermal stability test in order to estimate the decomposition of the lamellar structure with relation to swelling under irradiation condition. (author)

  6. Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on 239Pu, 235U, 238U

    International Nuclear Information System (INIS)

    Selby, H.D.; Mac Innes, M.R.; Barr, D.W.; Keksis, A.L.; Meade, R.A.; Burns, C.J.; Chadwick, M.B.; Wallstrom, T.C.

    2010-01-01

    We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400-600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4-1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for 99 Mo, 95 Zr, 137 Cs, 140 Ba, 141,143 Ce, and 147 Nd. Modest incident-energy dependence exists for the 147 Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by ∼5-7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried over to the ENDF/B-VII.0 library, except

  7. Co-extraction and co-stripping of U(VI) and Pu(IV) using tri-iso-amyl phosphate and tri-n-butyl phosphate in n-dodecane from nitric acid media under high loading conditions

    Energy Technology Data Exchange (ETDEWEB)

    Sreenivasulu, Balija; Suresh, Ammath; Sivaraman, Nagarajan; Rao, P.R. Vasudeva [Indira Gandhi Centre for Atomic Research, Kalpakkam (India). Chemistry Group

    2016-08-01

    The extraction of Pu(IV) using 1.1 M solution of tri-iso-amyl phosphate (TiAP)/n-dodecane (DD) from plutonium nitrate solutions in nitric acid media was examined as a function of equilibrium aqueous phase metal ion concentration and equilibrium aqueous phase acidity at 303 K. The nitric acid concentration in the organic phase was measured as a function of equilibrium organic phase plutonium concentration. The co-extraction of U(VI) and Pu(IV) was studied using 1.1 M TiAP/DD system as a function of their equilibrium aqueous phase metal ion concentration and compared with 1.1 M tri-n-butyl phosphate (TBP)/n-DD system under identical conditions. Co-extraction and co-stripping of U(VI) and Pu(IV) were studied using 1.1 M TiAP/DD and 1.1 M TBP/DD systems in cross current mode to evaluate the number of stages required for the extraction and stripping of heavy metal ions (uranium and plutonium). The extraction and stripping efficiencies were calculated for both the systems. The saturation limit of the organic phase was also established in these studies.

  8. Co-extraction and co-stripping of U(VI) and Pu(IV) using tri-iso-amyl phosphate and tri-n-butyl phosphate in n-dodecane from nitric acid media under high loading conditions

    International Nuclear Information System (INIS)

    Sreenivasulu, Balija; Suresh, Ammath; Sivaraman, Nagarajan; Rao, P.R. Vasudeva

    2016-01-01

    The extraction of Pu(IV) using 1.1 M solution of tri-iso-amyl phosphate (TiAP)/n-dodecane (DD) from plutonium nitrate solutions in nitric acid media was examined as a function of equilibrium aqueous phase metal ion concentration and equilibrium aqueous phase acidity at 303 K. The nitric acid concentration in the organic phase was measured as a function of equilibrium organic phase plutonium concentration. The co-extraction of U(VI) and Pu(IV) was studied using 1.1 M TiAP/DD system as a function of their equilibrium aqueous phase metal ion concentration and compared with 1.1 M tri-n-butyl phosphate (TBP)/n-DD system under identical conditions. Co-extraction and co-stripping of U(VI) and Pu(IV) were studied using 1.1 M TiAP/DD and 1.1 M TBP/DD systems in cross current mode to evaluate the number of stages required for the extraction and stripping of heavy metal ions (uranium and plutonium). The extraction and stripping efficiencies were calculated for both the systems. The saturation limit of the organic phase was also established in these studies.

  9. Study of phase transformation of U-2,5Zr-7,5Nb e U-3Zr-9Nb alloys for application in advanced nuclear fuel; Estudo das transformacoes de fases nas ligas U-2,5Zr-7,5Nb e U-3Zr-9Nb para aplicacao em combustivel nuclear avancado

    Energy Technology Data Exchange (ETDEWEB)

    Pais, Rafael Witter Dias

    2015-07-01

    Metal fuels are relevant in the nuclear area due to the versatility of its use in the nuclear fuel cycle. Among the alloys of uranium investigated with high potential for use in nuclear power reactors, U-Zr-Nb alloys appear as an important alternative because of their superior physico-chemical and metallurgical properties. These alloys have also potential for use in nuclear testing, research and production radioisotopes of high performance nuclear reactors. Therefore, the development of these alloys is strategic since they are planned to be used in national reactors as RMB (Brazilian Multipurpose Reactor) and LABGENE (Electrical Generation Core Laboratory), currently under development in Brazil. In this work it was realized a extensive study in the scope of the manufacturing, heat treatment and phase transformations of U-2,5Zr-7,5Nb (m/m%) and U-3ZR-9NB (m/m%) fuel alloys. Ingots of both alloys were produced employing a specific methodology developed in this study. This methodology comprised the melting process in a vacuum induction furnace at high temperatures (1500 °C) and thermal-mechanical processing to break the as-cast structure. Samples with typical dimensions (17 x 7 x 2.5 mm) free from macrostructural defects were homogenized at 1000 °C in vacuum of 10{sup -5} torr for 17.5 hours with a 10°C/min cooling rate until to 820 °C and, subsequently, quenched in water. The samples, randomly selected, were subjected to isothermal treatment tests under different conditions of time and temperature. Isothermal treatments for transformation and retention phases were carried out in a special assembly designed for this work. After the tests, the samples were characterized by the usual phase characterization techniques with particular emphasis for the X-ray diffraction technique. In this way, the Rietveld refinement method was applied. In the case of uranium based alloys it is quite challenging due to the lack of data in the literature. In this work a strategy for the

  10. Synthesis, spectral, DFT modeling, cytotoxicity and microbial studies of novel Zr(IV), Ce(IV) and U(VI) piroxicam complexes

    Science.gov (United States)

    El-Shwiniy, Walaa H.; Zordok, Wael A.

    2018-06-01

    The Zr(IV), Ce(IV) and U(VI) piroxicam anti-inflammatory drug complexes were prepared and characterized using elemental analyses, conductance, IR, UV-Vis, magnetic moment, IHNMR and thermal analysis. The ratio of metal: Pir is found to be 1:2 in all complexes estimated by using molar ratio method. The conductance data reveal that Zr(IV) and U(VI) chelates are non-electrolytes except Ce(IV) complex is electrolyte. Infrared spectroscopic confirm that the Pir behaves as a bidentate ligand co-ordinated to the metal ions via the oxygen and nitrogen atoms of ν(Cdbnd O)carbonyl and ν(Cdbnd N)pyridyl, respectively. The kinetic parameters of thermogravimetric and its differential, such as activation energy, entropy of activation, enthalpy of activation, and Gibbs free energy evaluated using Coats-Redfern and Horowitz-Metzger equations for Pir and complexes. The geometry of the piroxicam drug in the Free State differs significantly from that in the metal complex. In the time of metal ion-drug bond formation the drug switches-on from the closed structure (equilibrium geometry) to the open one. The antimicrobial tests were assessed towards some types of bacteria and fungi. The in vitro cell cytotoxicity of the complexes in comparison with Pir against colon carcinoma (HCT-116) cell line was measured. Optimized geometrical structure of piroxicam ligand by using DFT calculations.

  11. Studies on extraction behaviour of U (VI) and Pu (IV) by Aliquat-336 encapsulated in microporous polymer beads

    International Nuclear Information System (INIS)

    Parikh, K.J.; Kedari, C.S.; Pandit, S.S.; Tripathi, S.C.; Dwivedi, C.; Singh, K.K.; Kumar, M.; Bajaj, P.N.

    2009-01-01

    Aliquat-336 encapsulated microporous polymeric beads (AEPB) were prepared to investigate their applicability for the solid-liquid extraction Pu (IV) from nitric acid solutions. Batch equilibration studies on the extractions of U (VI) and Pu (IV) from aqueous solutions using AEPB have been carried out at different concentrations of nitric acid and sodium nitrate. Extraction of Pu (IV) increases with increasing concentration of nitrate ions in the aqueous phase where as extraction of U (VI) remains less than 5% for all the aqueous conditions studied. Polymeric beads appeared to be less stable at higher nitric acid concentrations as the extraction of Pu (IV) was found to be lowered above 4 M HNO 3 concentration. The maximum Pu (IV) uptake by AEPB was 0.84 μg per mg of beads. Pu (IV)from loaded polymer can be back extracted using dilute nitric acid or ascorbic acid solution. (author)

  12. Measurement of the fission cross-section of {sup 235}U and {sup 239}Pu for thermal neutrons; Mesures des sections de fission de {sup 235}U et de {sup 239}Pu en neutrons thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Fraysse, G; Prosdocimi, A; Netter, F; Samour, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    Improved techniques of fast detection have been applied for determining the fission cross-sections of {sup 235}U and {sup 239}Pu with reference to the absorption cross-section of Boron. Monochromatic neutron beams of 0.0322 eV, 0.0626 eV and 0.275 eV have been employed. Use has been made of a Xe-filled gaseous scintillator and of a low-geometry solid state ion chamber. Both measured alpha and fission rates. The results at the reference energy of 0.0253 eV are: ({sigma}{sub F}){sub 0} {sup 235}U = 588 {+-} 10 barns ({sigma}{sub F}){sub 0} {sup 239}Pu = 738 {+-} 7 barns. (authors) [French] Des techniques avancees de comptage rapide ont ete mise en oeuvre pour determiner la section efficace de fission de {sup 235}U et de {sup 239}Pu par rapport a celle d'absorption du bore. Des faisceaux de neutrons monochromatiques de 0,0322 eV, 0,0626 eV et 0,275 eV ont ete employes. Les detecteurs utilises sont un scintillateur gazeux rempli de xenon et une chambre d'ionisation a etat solide a basse geometrie. Les deux ont mesure les taux des desintegrations alpha et des fissions. Les resultats a l'energie de reference de 0,0253 eV sont: ({sigma}{sub F}){sub 0} {sup 235}U = 588 {+-} 10 barns ({sigma}{sub F}){sub 0} {sup 239}Pu = 738 {+-} 7 barns. (auteurs)

  13. Chemical Engineering Division fuel cycle programs. Quarterly progress report, April-June 1979. [Pyrochemical/dry processing; waste encapsulation in metal; transport in geologic media

    Energy Technology Data Exchange (ETDEWEB)

    Steindler, M.J.; Ader, M.; Barletta, R.E.

    1980-09-01

    For pyrochemical and dry processing materials development included exposure to molten metal and salt of Mo-0.5% Ti-0.07% Ti-0.01% C, Mo-30% W, SiC, Si/sub 2/ON/sub 2/, ZrB/sub 2/-SiC, MgAl/sub 2/O/sub 4/, Al/sub 2/O/sub 3/, AlN, HfB/sub 2/, Y/sub 2/O/sub 3/, BeO, Si/sub 3/N/sub 4/, nickel nitrate-infiltrated W, W-coated Mo, and W-metallized alumina-yttria. Work on Th-U salt transport processing included solubility of Th in liquid Cd, defining the Cd-Th and Cd-Mg-Th phase diagrams, ThO/sub 2/ reduction experiments, and electrolysis of CaO in molten salt. Work on pyrochemical processes and associated hardware for coprocessing U and Pu in spent FBR fuels included a second-generation computer model of the transport process, turntable transport process design, work on the U-Cu-Mg system, and U and Pu distribution coefficients between molten salt and metal. Refractory metal vessels are being service-life tested. The chloride volatility processing of Th-based fuel was evaluated for its proliferation resistance, and a preliminary ternary phase diagram for the Zn-U-Pu system was computed. Material characterization and process analysis were conducted on the Exportable Pyrochemical process (Pyro-Civex process). Literature data on oxidation of fissile metals to oxides were reviewed. Work was done on chemical bases for the reprocessing of actinide oxides in molten salts. Flowsheets are being developed for the processing of fuel in molten tin. Work on encapsulation of solidified radioactive waste in metal matrix included studies of leach rate of crystalline waste materials and of the impact resistance of metal-matrix waste forms. In work on the transport properties of nuclear waste in geologic media, adsorption of Sr on oolitic limestone was studied, as well as the migration of Cs in basalt. Fitting of data on the adsorption of iodate by hematite to a mathematical model was attempted.

  14. Investigation of Partially Crystalline Zr77Ni23 Metallic Glass

    Directory of Open Access Journals (Sweden)

    Amra Salčinović Fetić

    2016-08-01

    Full Text Available This paper presents the results of an extensive research of partially crystalline Zr77Ni23 metallic glass (indicated numbers refer to atomic percentages. The partially crystalline Zr77Ni23 samples were prepared by melt-spinning using a device constructed in the Metal Physics Laboratory, Faculty of Science in Sarajevo. XRD pattern shows crystalline peaks which correspond to an orthorhombic structure of Zr3Ni superimposed on an amorphous pattern. Homogeneity and chemical composition were investigated using scanning electron microscopy (SEM and energy-dispersive X-ray spectroscopy (EDX. Crystallization was studied by differential scanning calorimetry (DSC. DSC analysis indicated a simple thermally activated process. Overall activation energy of the crystallization was calculated using Kissinger's model for nonisothermal process and compared with those given by the Augis-Bennett model. By monitoring of the electrical resistance in the temperature range 80 – 270 K a small and negative thermal coefficient of electrical resistance was observed. This means that electrical resistance varies slightly with temperature and it makes this metallic glass suitable for application in electronic circuits for which this property is an important requirement.

  15. Dynamics of an unusual cone-building trachyte eruption at Pu`u Wa`awa`a, Hualālai volcano, Hawai`i

    Science.gov (United States)

    Shea, Thomas; Leonhardi, Tanis; Giachetti, Thomas; Lindoo, Amanda; Larsen, Jessica; Sinton, John; Parsons, Elliott

    2017-04-01

    The Pu`u Wa`awa`a pyroclastic cone and Pu`u Anahulu lava flow are two prominent monogenetic eruptive features assumed to result from a single eruption during the trachyte-dominated early post-shield stage of Hualālai volcano (Hawaíi). Púu Wa`awa`a is composed of complex repetitions of crudely cross-stratified units rich in dark dense clasts, which reversely grade into coarser pumice-rich units. Pyroclasts from the cone are extremely diverse texturally, ranging from glassy obsidian to vesicular scoria or pumice, in addition to fully crystalline end-members. The >100-m thick Pu`u Anahulu flow is, in contrast, entirely holocrystalline. Using field observations coupled with whole rock analyses, this study aimed to test whether the Pu`u Wa`awa`a tephra and Pu`u Anahulu lava flows originated from the same eruption, as had been previously assumed. Crystal and vesicle textures are characterized along with the volatile contents of interstitial glasses to determine the origin of textural variability within Pu`u Wáawáa trachytes (e.g., magma mixing vs. degassing origin). We find that (1) the two eruptions likely originated from distinct vents and magma reservoirs, despite their proximity and similar age, (2) the textural diversity of pyroclasts forming Pu`u Wa`awa`a can be fully explained by variable magma degassing and outgassing within the conduit, (3) the Pu`u Wa`awa`a cone was constructed during explosions transitional in style between violent Strombolian and Vulcanian, involving the formation of a large cone and with repeated disruption of conduit plugs, but without production of large pyroclastic density currents (PDCs), and (4) the contrasting eruption styles of Hawaiian trachytes (flow-, cone-, and PDC-forming) are probably related to differences in the outgassing capacity of the magmas prior to reaching the surface and not in intrinsic compositional or temperature properties. These results further highlight that trachytes are "kinetically faster" magmas compared

  16. Ni-free Zr-Cu-Al-Nb-Pd bulk metallic glasses with different Zr/Cu ratios for biomedical applications.

    Science.gov (United States)

    Huang, Lu; Yokoyama, Yoshihiko; Wu, Wei; Liaw, Peter K; Pang, Shujie; Inoue, Akihisa; Zhang, Tao; He, Wei

    2012-08-01

    Zr-based bulk metallic glasses (BMGs) possess attractive properties for prospective biomedical applications. The present study designs Ni-free Zr-Cu-Al-Nb-Pd BMGs and investigates their in vitro biocompatibility by studying mechanical properties, bio-corrosion resistance, and cellular responses. The Ti-6Al-4V alloy is used as a reference material. It is found that the Zr-based BMGs exhibit good mechanical properties, including high strengths above 1600 MPa, high hardness over 4700 MPa, and low elastic moduli of 85-90 GPa. The Zr-based BMGs are corrosion resistant in a simulated body environment, as revealed by wide passive regions, low passive current densities, and high pitting overpotentials. The formation of ZrO(2)-rich surface passive films of the Zr-based BMGs contributes to their high corrosion resistance, whereas their pitting corrosion in the phosphate buffered saline solution can be attributed to the sensitivity of the ZrO(2) films to the chloride ion. The general biosafety of the Zr-based BMGs is revealed by normal cell adhesions and cell morphologies. Moreover, the Zr/Cu content ratio in the alloy composition affects the biocompatibility of the Zr-based BMGs, by increasing their corrosion resistance and surface wettability with the increase of the Zr/Cu ratio. Effects of Zr/Cu ratios can be used to guide the future design of biocompatible Zr-based BMGs. Copyright © 2012 Wiley Periodicals, Inc.

  17. Hot-isostatic pressing of U-10Zr by grain boundary diffusion and creep cavitation. Part 2: Theory and data analysis

    International Nuclear Information System (INIS)

    McDeavitt, S.M.; Solomon, A.A.

    1997-01-01

    Uranium-10 wt % zirconium (U-10Zr) is a fuel alloy that has been used in the Experimental Breeder Reactor-II (EBR-II). The high burnup that was desired in this fuel system made high demands on the mechanical compatibility between fuel and cladding both during normal operation and during safety-related transients when rapid differential expansion may cause high stresses. In general, this mechanical stress can be reduced by cladding deformation if the cladding is sufficiently ductile at high burnup, and/or by fuel hot-pressing. Fortunately, the fuel is very porous when it contacts the cladding, but this porosity gradually fills with solid fission products (primarily lanthanides) that may limit the fuel's compressibility. If the porosity remains open, gaseous fission products are released and the porous fuel creeps rather than hot-presses under contact stresses. If the pores are closed by sintering or by solid fission products, the porous fuel will hot-isostatic press (HIP), as represented by the models to be discussed. HIP experiments performed at 700 C on U-10Zr samples with different impurity phase contents (Part 1) are analyzed in terms of several creep cavitation models. The coupled diffusion/creep cavitation model of Chen and Argon shows good quantitative agreement with measured HIP rates for hydride- and metal-derived U-10Zr materials, assuming that pores are uniformly distributed on grain boundaries and are of modal size, and that far-field strain rates are negligible. The analysis predicts, for the first time, an asymmetry between HIP and swelling at identical pressure-induced driving forces due to differences in grain boundary stresses. The differences in compressibility of hydride- and metal-derived U-10Zr can be partially explained by differences in pore size and spacing. The relevance of the experiments to description of in-reactor densification under external pressure or contact stress due to fuel/cladding mechanical interaction is discussed

  18. Preparation and characterization of Zr-based bulk metallic glasses in form of plate

    International Nuclear Information System (INIS)

    Pilarczyk, Wirginia

    2014-01-01

    Highlights: • Zr-based BMGs in form of plate was successful produced by die pressure casting method. • Many techniques have been used to characterize the structure of Zr 55 Cu 30 Ni 5 Al 10 alloy. • The calculated GFA parameters show that the alloy exhibits satisfactory GFA. • The studies reveal that tested as-cast Zr-based alloy is in amorphous state. - Abstract: Zr-based bulk metallic glasses present an interesting combination of physical, chemical and mechanical properties. During the last decade, intensive progress has been made and a number of applications have been suggested for these materials. In order to successfully apply these materials, it is necessary to accurately characterize their structure, thermal stability and other properties accurately. The aim of the presented work is the manufacturing, examination of the structure of selected Zr-based bulk metallic alloys and confirmation of an amorphous structure using X-ray analysis, microscopic observation and thermal analysis. In this work, the Zr-based bulk metallic glasses in form of plate was successful produced by die pressure casting method. Designed scientific station for casting zirconium based amorphous alloys in the form of plates and rods with selected dimensions is in our university a comprehensive method for achieving amorphous materials which enables us to maintain repeatability of as-cast samples with the amorphous structure and the assumed dimensions range. The diffraction pattern and exothermic reaction as well as the fracture surface morphology reveal that studied as-cast Zr-based alloy is in amorphous state. The calculated GFA parameters show that the alloy exhibits satisfactory glass-forming ability in form of studied plate. These obtained values can suggest that studied alloys are suitable materials for further planned practical application at welding process. The success of Zr-based bulk metallic glasses production in form of plate with obtained sizes is important for future

  19. Preparation and characterization of Zr-based bulk metallic glasses in form of plate

    Energy Technology Data Exchange (ETDEWEB)

    Pilarczyk, Wirginia, E-mail: wirginia.pilarczyk@polsl.pl

    2014-12-05

    Highlights: • Zr-based BMGs in form of plate was successful produced by die pressure casting method. • Many techniques have been used to characterize the structure of Zr{sub 55}Cu{sub 30}Ni{sub 5}Al{sub 10} alloy. • The calculated GFA parameters show that the alloy exhibits satisfactory GFA. • The studies reveal that tested as-cast Zr-based alloy is in amorphous state. - Abstract: Zr-based bulk metallic glasses present an interesting combination of physical, chemical and mechanical properties. During the last decade, intensive progress has been made and a number of applications have been suggested for these materials. In order to successfully apply these materials, it is necessary to accurately characterize their structure, thermal stability and other properties accurately. The aim of the presented work is the manufacturing, examination of the structure of selected Zr-based bulk metallic alloys and confirmation of an amorphous structure using X-ray analysis, microscopic observation and thermal analysis. In this work, the Zr-based bulk metallic glasses in form of plate was successful produced by die pressure casting method. Designed scientific station for casting zirconium based amorphous alloys in the form of plates and rods with selected dimensions is in our university a comprehensive method for achieving amorphous materials which enables us to maintain repeatability of as-cast samples with the amorphous structure and the assumed dimensions range. The diffraction pattern and exothermic reaction as well as the fracture surface morphology reveal that studied as-cast Zr-based alloy is in amorphous state. The calculated GFA parameters show that the alloy exhibits satisfactory glass-forming ability in form of studied plate. These obtained values can suggest that studied alloys are suitable materials for further planned practical application at welding process. The success of Zr-based bulk metallic glasses production in form of plate with obtained sizes is

  20. Direct spectrophotometric analysis of low level Pu (III) in Pu(IV) nitrate solution

    International Nuclear Information System (INIS)

    Mageswaran, P.; Suresh Kumar, K.; Kumar, T.; Gayen, J.K.; Shreekumar, B.; Dey, P.K.

    2010-01-01

    Among the various methods demonstrated for the conversion of plutonium nitrate to its oxide, the oxalate precipitation process either as Pu (III) or Pu (IV) oxalate gained wide acceptance. Since uranous nitrate is the most successful partitioning agent used in the PUREX process for the separation of Pu from the bulk amount of U, the Pu (III) oxalate precipitation of the purified nitrate solution will not give required decontamination from U. Hence Pu IV oxalate precipitation process is a better option to achieve the end user's specified PuO 2 product. Prior to the precipitation process, ensuring of the Pu (IV) oxidation state is essential. Hence monitoring of the level of Pu oxidation state either Pu (III) or Pu (IV) in the feed solution plays a significant role to establish complete conversion of Pu (III). The method in vogue to estimate Pu(lV) content is extractive radiometry using Theonyl Trifluoro Acetone (TTA). As the the method warrants a sample preparation with respect to acidity, a precise measurement of Pu (IV) without affecting the Pu(III) level in the feed sample is difficult. Present study is focused on the exploration of direct spectrophotometry using an optic fiber probe of path length of 40mm to monitor the low level of Pu(III) after removing the bulk Pu(lV) which interfere in the Pu(III) absorption spectrum, using TTA-TBP synergistic mixture without changing the sample acidity

  1. Speciation analysis of 129I, 137Cs, 232Th, 238U, 239Pu and 240Pu in environmental soil and sediment

    DEFF Research Database (Denmark)

    Qiao, Jixin; Hansen, Violeta; Hou, Xiaolin

    2012-01-01

    The environmental mobility and bioavailability of radionuclides are related to their physicochemical forms, namely species. We here present a speciation analysis of important radionuclides including 129I (also 127I), 137Cs, 232Th, 238U and plutonium isotopes (239Pu and 240Pu) in soil (IAEA-375......) and sediment (NIST-4354) standard reference materials and two fresh sediment samples from Øvre Heimdalsvatnet Lake, Norway. A modified sequential extraction protocol was used for the speciation analysis of these samples to obtain fractionation information of target radionuclides. Analytical results reveal...

  2. Fabrication of cermet fuel for fast reactor

    International Nuclear Information System (INIS)

    Mishra, Sudhir; Kumar, Arun; Kutty, T.R.G.; Kamath, H.S.

    2011-01-01

    Mixed oxide (MOX) (U,Pu)O 2 , and metallic (U,Pu ,Zr) fuels are considered promising fuels for the fast reactor. The fuel cycle of MOX is well established. The advantages of the oxide fuel are its easy fabricability, good performance in the reactor and a well established reprocessing technology. However the problems lie in low thermal conductivity , low density of the fuel leading to low breeding ratio and consequently longer doubling time. The metallic fuel has the advantages of high thermal conductivity, higher metal density and higher coefficient of linear expansion. The higher coefficient of linear expansion is good from the safety consideration (negative reactivity factor). Because of higher metal density it offers highest breeding ratio and shortest doubling time. Metallic fuel disadvantages comprise large swelling at high burnup, fuel cladding interaction and lower margin between operating and melting temperature. The optimal solution may lie in cermet fuel (U, PuO 2 ), where PuO 2 is dispersed in U metal matrix and combines the favorable features of both the fuel types. The advantages of this fuel include high thermal conductivity, larger margin between melting and operating temperature, ability to retain fission product etc. The matrix being of high density metal the advantage of high breeding ratio is also maintained. In this report some results of fabrication of cermet pellet comprising of UO 2 /PuO 2 dispersed in U metal powder through classical powder metallurgy route and characterization are presented. (author)

  3. Biocompatible Ni-free Zr-based bulk metallic glasses with high-Zr-content: compositional optimization for potential biomedical applications.

    Science.gov (United States)

    Hua, Nengbin; Huang, Lu; Chen, Wenzhe; He, Wei; Zhang, Tao

    2014-11-01

    The present study designs and prepares Ni-free Zr60+xTi2.5Al10Fe12.5-xCu10Ag5 (at.%, x=0, 2.5, 5) bulk metallic glasses (BMGs) by copper mold casting for potential biomedical application. The effects of Zr content on the in vitro biocompatibility of the Zr-based BMGs are evaluated by investigating mechanical properties, bio-corrosion behavior, and cellular responses. It is found that increasing the content of Zr is favorable for the mechanical compatibility with a combination of low Young's modulus, large plasticity, and high notch toughness. Electrochemical measurements demonstrate that the Zr-based BMGs are corrosion resistant in a phosphate buffered saline solution. The bio-corrosion resistance of BMGs is improved with the increase in Zr content, which is attributed to the enrichment in Zr and decrease in Al concentration in the surface passive film of alloys. Regular cell responses of mouse MC3T3-E1 cells, including cell adhesion and proliferation, are observed on the Zr-Ti-Al-Fe-Cu-Ag BMGs, which reveals their general biosafety. The high-Zr-based BMGs exhibit a higher cell proliferation activity in comparison with that of pure Zr and Ti-6Al-4V alloy. The effects of Zr content on the in vitro biocompatibility can be used to guide the future design of biocompatible Zr-based BMGs. Copyright © 2014 Elsevier B.V. All rights reserved.

  4. Melting metal waste for volume reduction and decontamination

    International Nuclear Information System (INIS)

    Copeland, G.L.; Heshmatpour, B.; Heestand, R.L.

    1980-01-01

    Melt-slagging was investigated as a technique for volume reduction and decontamination of radioactively contaminated scrap metals. Experiments were conducted using several metals and slags in which the partitioning of the contaminant U or Pu to the slag was measured. Concentrations of U or Pu in the metal product of about 1 ppM were achieved for many metals. A volume reduction of 30:1 was achieved for a typical batch of mixed metal scrap. Additionally, the production of granular products was demonstrated with metal shot and crushed slag

  5. Sequential Injection Method for Rapid and Simultaneous Determination of 236U, 237Np, and Pu Isotopes in Seawater

    DEFF Research Database (Denmark)

    Qiao, Jixin; Hou, Xiaolin; Steier, Peter

    2013-01-01

    An automated analytical method implemented in a novel dual-column tandem sequential injection (SI) system was developed for simultaneous determination of 236U, 237Np, 239Pu, and 240Pu in seawater samples. A combination of TEVA and UTEVA extraction chromatography was exploited to separate and purify...... target analytes, whereupon plutonium and neptunium were simultaneously isolated and purified on TEVA, while uranium was collected on UTEVA. The separation behavior of U, Np, and Pu on TEVA–UTEVA columns was investigated in detail in order to achieve high chemical yields and complete purification...

  6. Electronic structure of bulk AnO{sub 2} (An = U, Np, Pu) and water adsorption on the (111) and (110) surfaces of UO{sub 2} and PuO{sub 2} from hybrid density functional theory within the periodic electrostatic embedded cluster method

    Energy Technology Data Exchange (ETDEWEB)

    Wellington, Joseph P.W. [Department of Chemistry, University College London, 20 Gordon Street, London WC1H 0AJ (United Kingdom); Kerridge, Andrew [Department of Chemistry, Lancaster University, Bailrigg, Lancaster LA1 4YP (United Kingdom); Austin, Jonathan [National Nuclear Laboratory, Chadwick House, Warrington Road, Birchwood Park, Warrington WA3 6AE (United Kingdom); Kaltsoyannis, Nikolas, E-mail: nikolas.kaltsoyannis@manchester.ac.uk [School of Chemistry, The University of Manchester, Oxford Road, Manchester M13 9PL (United Kingdom)

    2016-12-15

    Generalised gradient approximation (PBE) and hybrid (PBE0) density functional theory (DFT) within the periodic electrostatic embedded cluster method have been used to study AnO{sub 2} bulk and surfaces (An = U, Np, Pu). The electronic structure has been investigated by examining the projected density of states (PDOS). While PBE incorrectly predicts these systems to be metallic, PBE0 finds them to be insulators, with the composition of the valence and conduction levels agreeing well with experiment. Molecular and dissociative water adsorption on the (111) and (110) surfaces of UO{sub 2} and PuO{sub 2} has been investigated, with that on the (110) surface being stronger than on the (111). Similar energies are found for molecular and dissociative adsorption on the (111) surfaces, while on the (110) there is a clear preference for dissociative adsorption. Adsorption energies and geometries on the (111) surface of UO{sub 2} are in good agreement with recent periodic DFT studies using the GGA+U approach, and our data for dissociative adsorption on the (110) surface of PuO{sub 2} match experiment rather well, especially when dispersion corrections are included. - Graphical abstract: The electronic structures of AnO{sub 2} (An = U, Np, Pu) are studied computationally with hybrid density functional theory, and the geometries and energetics of water adsorption on the low index surfaces are presented.

  7. Chemical speciation of U, Fe, and Pu in melt glass from nuclear weapons testing

    Energy Technology Data Exchange (ETDEWEB)

    Pacold, J. I.; Lukens, W. W.; Booth, C. H.; Shuh, D. K. [Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, California 94720 (United States); Knight, K. B.; Eppich, G. R. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Holliday, K. S. [Materials Science Division, Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

    2016-05-21

    Nuclear weapons testing generates large volumes of glassy materials that influence the transport of dispersed actinides in the environment and may carry information on the composition of the detonated device. We determine the oxidation state of U and Fe (which is known to buffer the oxidation state of actinide elements and to affect the redox state of groundwater) in samples of melt glass collected from three U.S. nuclear weapons tests. For selected samples, we also determine the coordination geometry of U and Fe, and we report the oxidation state of Pu from one melt glass sample. We find significant variations among the melt glass samples and, in particular, find a clear deviation in one sample from the expected buffering effect of Fe(II)/Fe(III) on the oxidation state of uranium. In the first direct measurement of Pu oxidation state in a nuclear test melt glass, we obtain a result consistent with existing literature that proposes Pu is primarily present as Pu(IV) in post-detonation material. In addition, our measurements imply that highly mobile U(VI) may be produced in significant quantities when melt glass is quenched rapidly following a nuclear detonation, though these products may remain immobile in the vitrified matrices. The observed differences in chemical state among the three samples show that redox conditions can vary dramatically across different nuclear test conditions. The local soil composition, associated device materials, and the rate of quenching are all likely to affect the final redox state of the glass. The resulting variations in glass chemistry are significant for understanding and interpreting debris chemistry and the later environmental mobility of dispersed material.

  8. Chemical speciation of U, Fe, and Pu in melt glass from nuclear weapons testing

    Science.gov (United States)

    Pacold, J. I.; Lukens, W. W.; Booth, C. H.; Shuh, D. K.; Knight, K. B.; Eppich, G. R.; Holliday, K. S.

    2016-05-01

    Nuclear weapons testing generates large volumes of glassy materials that influence the transport of dispersed actinides in the environment and may carry information on the composition of the detonated device. We determine the oxidation state of U and Fe (which is known to buffer the oxidation state of actinide elements and to affect the redox state of groundwater) in samples of melt glass collected from three U.S. nuclear weapons tests. For selected samples, we also determine the coordination geometry of U and Fe, and we report the oxidation state of Pu from one melt glass sample. We find significant variations among the melt glass samples and, in particular, find a clear deviation in one sample from the expected buffering effect of Fe(II)/Fe(III) on the oxidation state of uranium. In the first direct measurement of Pu oxidation state in a nuclear test melt glass, we obtain a result consistent with existing literature that proposes Pu is primarily present as Pu(IV) in post-detonation material. In addition, our measurements imply that highly mobile U(VI) may be produced in significant quantities when melt glass is quenched rapidly following a nuclear detonation, though these products may remain immobile in the vitrified matrices. The observed differences in chemical state among the three samples show that redox conditions can vary dramatically across different nuclear test conditions. The local soil composition, associated device materials, and the rate of quenching are all likely to affect the final redox state of the glass. The resulting variations in glass chemistry are significant for understanding and interpreting debris chemistry and the later environmental mobility of dispersed material.

  9. Consistent Data Assimilation of Actinide Isotopes: 235U and 239Pu

    International Nuclear Information System (INIS)

    Palmiottti, G.; Hiruta, H.; Salvatores, M.

    2011-01-01

    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 were analyzed using the EMPIRE evaluated files for 235 U, 238 U, and 239 Pu. Inmost cases the results have shown quite large worse results with respect to the corresponding existing evaluations available for ENDF/B-VII. The observed discrepancies between calculated and experimental results were used in conjunction with the computed sensitivity coefficients and covariance matrix for nuclear parameters in a consistent data assimilation. Only the GODIVA and JEZEBEL experimental results were used, in order to exploit information relative to the isotope of interest that are, in this particular case: 235 U and 239 Pu. The results obtained by the consistent data assimilation indicate that with reasonable modifications (mostly within the initial standard deviation) it is possible to eliminate the original large discrepancies on the K eff of the two critical configurations. However, some residual discrepancy remains for a few fission spectral indices that are, most likely, to be attributed to the detector cross sections.

  10. Chemical thermodynamics of the system Cs--U--Zr--H--I--O in the LWR fuel-clad gap

    International Nuclear Information System (INIS)

    Besmann, T.M.; Lindemer, T.B.

    1978-01-01

    Equilibrium thermodynamic calculations were performed on the are Cs-U-Zr-H-I-O system that is assumed to exist in the fuel-clad gap of light water reactor fuel under in-reactor, steam, and 50% steam--50% air conditions. The in-reactor oxygen potential is assumed to be controlled by UO/sub 2+x/ rather than Zr + ZrO 2 . Thus, the important condensed phases present are UO/sub 2+x/, Cs 2 UO 4 , and CsI, and the major gaseous species are Cs, CsI, and Cs 2 I 2 . The presence of steam does not alter the species present, although CsOH also becomes a major gaseous species. In a 50% steam--50% air mixture, the condensed phases U 3 O 8 or UO 3 , Cs 2 U 15 O 46 , and ZrI 3 or liquid ZrI 2 are present at equilibrium, and the gaseous species ZrI 2 , ZrI 3 , and ZrI 4 have large partial pressures

  11. Calculated k-effectives for light water reactor typical, U + Pu nitrate solution critical experiments

    International Nuclear Information System (INIS)

    Primm, R.T. III; Mincey, J.F.

    1982-01-01

    The Department of Energy's Consolidated Fuel Reprocessing Program has as a goal the design of nuclear fuel reprocessing equipment. In order to validate computer codes used for criticality analyses in the design of such equipment, k-effectives have been calculated for several U + Pu nitrate solution critical experiments. As of January 1981, descriptions of 45 unpoisoned, U + Pu solution experiments were available in the open literature. Twelve of these experiments were performed with solutions which have physical characteristics typical of dissolved, light water reactor fuel. This paper contains a discussion of these twelve experiments, a review of the calculational procedure used to determine k-effectives, and the results of the calculations

  12. Cross sections and neutron yields for U233, U235 and Pu239 at 2200 m/sec

    International Nuclear Information System (INIS)

    Sjoestrand, N.G.; Story, J.S.

    1960-04-01

    The experimental information on the 2200 m/sec values for σ abs , σ f , α, ν and η for 233 U , 235 U and 23 been collected and discussed. The values will later be used in an evaluation of a 'best' set of data. In appendix the isotopic abundances of the uranium isotopes are discussed and also the alpha activities of the uranium isotopes and Pu-239

  13. Interatomic potential to predict the favored and optimized compositions for ternary Cu-Zr-Hf metallic glasses

    International Nuclear Information System (INIS)

    Luo, S. Y.; Cui, Y. Y.; Dai, Y.; Li, J. H.; Liu, B. X.

    2012-01-01

    Under the framework of smoothed and long range second-moment approximation of tight-binding, a realistic interatomic potential was first constructed for the Cu-Zr-Hf ternary metal system. Applying the constructed potential, Monte Carlo simulations were carried out to compare the relative stability of crystalline solid solution versus its disordered counterpart over the entire composition triangle of the system (as a function of alloy composition). Simulations not only reveal that the origin of metallic glass formation but also determine, in the composition triangle, a quadrilateral region, within which metallic glass formation is energetically favored. It is proposed to define the energy differences between the crystalline solid solutions and disordered states as the driving force for amorphization and the corresponding calculations pinpoint an optimized composition locating at an composition of Cu 55 Zr 10 Hf 35 , around which the driving force for metallic glass formation reaches its maximum, suggesting that the ternary Cu-Zr-Hf metallic glasses designed to have the compositions around Cu 55 Zr 10 Hf 35 could be more stable than other alloys in the system. Moreover, for the Cu 55 Zr 10 Hf 35 metallic glass, the Voronoi tessellation calculations reveal some interesting features of its atomic configurations and coordination polyhedra distribution.

  14. Development of Metallic Fuels for Actinide Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, Steven Lowe [Idaho National Laboratory; Fielding, Randall Sidney [Idaho National Laboratory; Benson, Michael Timothy [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory; Carmack, William Jonathan [Idaho National Laboratory

    2015-09-01

    Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimized for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10

  15. Effect of Nb element content in U-Zr-Nb alloy on hardness, microstructure and phase formation

    International Nuclear Information System (INIS)

    Masrukan; M Husna Al Hasa; Jan Setiawan; Slamet Pribadi

    2015-01-01

    Experiments to determine the effect of Nb element in the U-Zr alloys on hardness, microstructure and phase formation has been done. The addition of Nb element would effect the hardness, microstructure and phase which formed. The U-Zr-Nb alloy was made with the variation of Nb 2%, 5% and 8% by melting in an electric arc melting furnace that equipped with water cooling and the argon atmosphere. The U-Zr-Nb alloy to be cut divided to some testing, such as hardness test, microstructure, and phase analysis. Hardness testing was done by Vickers hardness testing equipment, microstructure by an optical microscope, and diffraction pattern by XRD and phase analysis was done by GSAS. Hardness testing results showed that the addition of 2% to 5% Nb element in U-Zr alloys will increased in hardness, but the addition of Nb element over 5% the hardness was decreased. Observations the microstructure showed that the addition of 2% to 5% Nb element, grains were formed from fine into coarse. Phase analysis for diffraction pattern showed that the phase changed from αU and γU (Zr,Nb)at 2% Nb to be αU, γU (Zr,Nb) and δ1 (UZr_2) phase at 5% and 8% Nb. Phase changes was followed by changes in its compositions. The composition of αU at 2% Nb was 40% increased to 81% at 5% Nb and decreased to 3.9% at 8% Nb. The composition of γU decreased from 59,86% to 14,91% with increased Nb from 2% to 5% and further increased to 52,74% at 8% Nb. (author)

  16. Reactivity prediction of uniform PuO2-UO2 fuelled lattices and Pu(NO3)4 solutions in light water

    International Nuclear Information System (INIS)

    Mohankrishnan, P.; Huria, H.C.

    A theoretical analysis of the reactivities of the experimentally measured uniform light water moderated and reflected PuO 2 in UO 2 lattices and Pu(NO 3 ) 4 solutions is presented here. The mixed oxide single rod lattices are homogenised by the use of multigroup integral transport theory and diffusion theory is used for the cylindrical core calculations. The cross-sections are derived from the WTIS library. The homogeneous spherical Pu(NO 3 ) 4 solutions are analysed by discrete ordinate transport theory. Due to the small size of these criticals, it is necessary that one dimensional core calculations also be performed with a cross-section energy group structure which can represent neutron slowing down and thermalisation at the core reflector interface accurately. Due to the absence of such core calculation in the BNWL analyses of the mixed oxide lattices, the agreement of or predictions for these lattices with measurement is considered to be more satisfactory. These reactivity predictions are found to agree generally within +- 0.6% of measurements for the mixed oxide lattices and within 1% for the solution system. (author)

  17. Formation of hydroxyapatite on Ti-coated Ti-Zr-Cu-Pd bulk metallic glass

    International Nuclear Information System (INIS)

    Qin, F.X.; Wang, X.M.; Wada, T.; Xie, G.Q.; Asami, K.; Inoue, A.

    2009-01-01

    In this research, Ti coating was conducted on Ti 40 Zr 10 Cu 36 Pd 14 bulk metallic glass (BMG) in order to increase the formation rate of hydroxyapatite layer. The formation behavior of bone-like hydroxyapatite on Ti-coated and uncoated Ti 40 Zr 10 Cu 36 Pd 14 bulk metallic glasses (BMGs) was studied. The surface morphology of Ti-coated and uncoated Ti 40 Zr 10 Cu 36 Pd 14 BMG was investigated by scanning electron microscopy and energy dispersive X-ray spectroscopy. The results revealed that the alkali pretreatment in 5 M NaOH solution at 60degC for 24 h had a beneficial effect on the formation of porous sodium titanate on Ti-coated Ti 40 Zr 10 Cu 36 Pd 14 BMG. A bone-like hydroxyapatite layer was able to form on the alkali-treated Ti-coated Ti 40 Zr 10 Cu 36 Pd 14 BMG after a short-time immersion in simulated body fluid (SBF). On the contrary, hydroxyapatite formation was not observed on the uncoated Ti 40 Zr 10 Cu 36 Pd 14 BMG after the same chemical treatments. (author)

  18. On uncertainties and fluctuations of averaged neutron cross sections in unresolved resonance energy region for 235U, 238U, 239Pu

    International Nuclear Information System (INIS)

    Van'kov, A.A.; Blokhin, A.I.; Manokhin, V.N.; Kravchenko, I.V.

    1985-01-01

    This paper analyses the reasons for the differences which exist between group-averaged evaluated cross-section data from different evaluated data files for U235, U238 and Pu239 in the unresolved resonance energy region. (author)

  19. Pyrochemical processes for the recovery of weapons grade plutonium either as a metal or as PuO2 for use in mixed oxide reactor fuel pellets

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  20. Sodium-cooled Fast Reactor Cores using Uranium-Free Metallic Fuels for Maximizing TRU Support Ratio

    International Nuclear Information System (INIS)

    You, WuSeung; Hong, Ser Gi

    2014-01-01

    The depleted uranium plays important roles in the SFR burner cores because it substantially contributes to the inherent safety of the core through the negative Doppler coefficient and large delayed neutron. However, the use of depleted uranium as a diluent nuclide leads to a limited value of TRU support ratio due to the generation of TRUs through the breeding. In this paper, we designed sodium cooled fast reactor (SFR) cores having uranium-free fuels 3,4 for maximization of TRU consumption rate. However, the uranium-free fuelled burner cores can be penalized by unacceptably small values of the Doppler coefficient and small delayed neutron fraction. In this work, metallic fuels of TRU-(W or Ni)-Zr are considered to improve the performances of the uranium-free cores. The objective of this work is to consistently compare the neutronic performances of uranium-free sodium cooled fast reactor cores having TRU-Zr metallic fuels added with Ni or W and also to clarify what are the problematic features to be resolved. In this paper, a consistent comparative study of 400MWe sodium cooled burner cores having uranium-based fuels and uranium-free fuels was done to analyze the relative core neutronic features. Also, we proposed a uranium-free metallic fuel based on Nickel. From the results, it is found that tungsten-based uranium-free metallic fuel gives large negative Doppler coefficient due to high resonance of tungsten isotopes but this core has large sodium void worth and small effective delayed neutron fraction while the nickel-based uranium-free metallic fuelled core has less negative Doppler coefficient but smaller sodium void worth and larger effective delayed neutron fraction than the tungsten-based one. On the other hand, the core having TRU-Zr has very high burnup reactivity swing which may be problematic in compensating it using control rods and the least negative Doppler coefficient

  1. Biocompatibility study on Ni-free Ti-based and Zr-based bulk metallic glasses

    Energy Technology Data Exchange (ETDEWEB)

    Li, T.H. [Institute of Material Science and Engineering, National Central University, Taoyuan, Taiwan (China); Wong, P.C. [Department of Biomedical Engineering, National Yang-Ming University, Taipei, Taiwan (China); Chang, S.F. [Department of Mechanical Engineering, National Central University, Taoyuan, Taiwan (China); Tsai, P.H. [Institute of Material Science and Engineering, National Central University, Taoyuan, Taiwan (China); Jang, J.S.C., E-mail: jscjang@ncu.edu.tw [Institute of Material Science and Engineering, National Central University, Taoyuan, Taiwan (China); Department of Mechanical Engineering, National Central University, Taoyuan, Taiwan (China); Huang, J.C. [Department of Materials and Optoelectronic Science, National Sun Yat-Sen University, Kaohsiung, Taiwan (China)

    2017-06-01

    Safety and reliability are crucial issues for medical instruments and implants. In the past few decays, bulk metallic glasses (BMGs) have drawn attentions due to their superior mechanical properties, good corrosion resistance, antibacterial and good biocompatibility. However, most Zr-based and Ti-based BMGs contain Ni as an important element which is prone to human allergy problem. In this study, the Ni-free Ti-based and Zr-based BMGs, Ti{sub 40}Zr{sub 10}Cu{sub 36}Pd{sub 14}, and Zr{sub 48}Cu{sub 36}Al{sub 8}Ag{sub 8}, were selected for systematical evaluation of their biocompatibility. Several biocompatibility tests, co-cultural with L929 murine fibroblast cell line, were carried out on these two BMGs, as well as the comparison samples of Ti6Al4V and pure Cu. The results in terms of cellular adhesion, cytotoxicity, and metallic ion release affection reveal that the Ti{sub 40}Zr{sub 10}Cu{sub 36}Pd{sub 14} BMG and Ti6Al4V exhibit the optimum biocompatibility; cells still being attached on the petri dish with good adhesion and exhibiting the spindle shape after direct contact test. Furthermore, the Ti{sub 40}Zr{sub 10}Cu{sub 36}Pd{sub 14} BMG showed very low Cu ion release level, in agreement with the MTT results. Based on the current findings, it is believed that Ni-free Ti-based BMG can act as an ideal candidate for medical implant. - Highlight: • Ni-free bulk metallic glass is promising material for medical implants. • Ni-free Ti-based BMG presents similar cellular adhesion as Ti6Al4V. • Ni-free Ti-based BMG shows less cytotoxicity, and metallic ion release than Ti6Al4V.

  2. Thermal-Neutron-Induced Fission of U235, U233 and Pu239

    International Nuclear Information System (INIS)

    Thomas, T.D.; Gibson, W.M.; Safford, G.J.

    1965-01-01

    We have used solid-state detectors to measure the kinetic energies of the coincident fission fragments in the thermal-neutron-induced fission of U 235 , U 233 and Pu 239 . Special care has been taken to eliminate spurious-events near symmetry to give an accurate measure of such quantities as the average total kinetic energy at symmetry. For each fissioning system over 10 6 events were recorded. As a result the statistics are good enough to see definite evidence for fine structure over a wide range of masses and energies. The data have been analysed to give mass yield curves, average kinetic energies as a function of mass, and other quantities of interest. For each fissioning system the average total kinetic energy goes through a maximum for a heavy fragment mass of about 132 and for the corresponding light fragment mass. There is a pronounced minimum at symmetry, although not as deep as that found in time-of-flight experiments. The difference between the maximum average kinetic energy and that at symmetry is about 32 MeV for U 235 , 18 MeV for U 233 and 20 MeV for Pu 239 . The dispersion of kinetic energies at symmetry is also smaller than that found in time-of-flight experiments. Fine structure is apparent in two different representations of the data. The energy spectrum of heavy fragments in coincidence with light fragment energies is greater than the most probable value. This structure becomes more pronounced as the light fragment energy increases. The mass yield curves for a given total kinetic energy show a structure suggesting a preference for fission fragments with masses ∼134, ∼140 and ∼145 (and their light fragment partners). Much of the structure observed can be understood by considering a semi-empirical mass surface and a simple model for the nuclear configuration at the saddle point. (author) [fr

  3. Fracture toughness study of new Zr-based Be-bearing bulk metallic glasses

    OpenAIRE

    Kim, C. Paul; Suh, Jin-Yoo; Wiest, Aaron; Lind, Mary Laura; Conner, R. Dale; Johnson, William L.

    2009-01-01

    Three new compositional variants of the Zr–Ti–Be–LTM (late transition metal) family of metallic glasses are discussed. Thermal stability, ΔT = T_x−T_g, was increased from 82 °C for Zr_(41.2)Ti_(13.8)Cu_(12.5)Ni_(10)Be_(22.5) (Viterloy 1) to 141 °C for Zr_(44)Ti_(11)Cu_(20)Be_(25). It is found that fracture toughness is the most distinguishing parameter characterizing the alloys in contrast to other mechanical properties. Quaternary alloys consistently had fracture toughness values exceeding 8...

  4. Study of phase transformation of U-2,5Zr-7,5Nb e U-3Zr-9Nb alloys for application in advanced nuclear fuel

    International Nuclear Information System (INIS)

    Pais, Rafael Witter Dias

    2015-01-01

    Metal fuels are relevant in the nuclear area due to the versatility of its use in the nuclear fuel cycle. Among the alloys of uranium investigated with high potential for use in nuclear power reactors, U-Zr-Nb alloys appear as an important alternative because of their superior physico-chemical and metallurgical properties. These alloys have also potential for use in nuclear testing, research and production radioisotopes of high performance nuclear reactors. Therefore, the development of these alloys is strategic since they are planned to be used in national reactors as RMB (Brazilian Multipurpose Reactor) and LABGENE (Electrical Generation Core Laboratory), currently under development in Brazil. In this work it was realized a extensive study in the scope of the manufacturing, heat treatment and phase transformations of U-2,5Zr-7,5Nb (m/m%) and U-3ZR-9NB (m/m%) fuel alloys. Ingots of both alloys were produced employing a specific methodology developed in this study. This methodology comprised the melting process in a vacuum induction furnace at high temperatures (1500 °C) and thermal-mechanical processing to break the as-cast structure. Samples with typical dimensions (17 x 7 x 2.5 mm) free from macrostructural defects were homogenized at 1000 °C in vacuum of 10 -5 torr for 17.5 hours with a 10°C/min cooling rate until to 820 °C and, subsequently, quenched in water. The samples, randomly selected, were subjected to isothermal treatment tests under different conditions of time and temperature. Isothermal treatments for transformation and retention phases were carried out in a special assembly designed for this work. After the tests, the samples were characterized by the usual phase characterization techniques with particular emphasis for the X-ray diffraction technique. In this way, the Rietveld refinement method was applied. In the case of uranium based alloys it is quite challenging due to the lack of data in the literature. In this work a strategy for the

  5. Microstructural studies of suck cast (Zr-SS)-3 and 5 AI alloys for nuclear metallic waste form

    International Nuclear Information System (INIS)

    Kumar, P.; Das, N.; Sengupta, P.; Arya, A.; Dey, G.K.

    2015-01-01

    Management of radioactive metallic waste using 'alloy melting route' is currently being investigated. For disposal of Zr and SS base nuclear metallic wastes, Zr-stainless steel (SS) hybrid alloys are being considered as baseline alloys for developing metallic-waste-form (MWF) alloys. In this context Zr-16 wt. %55 has been selected for MWF alloy in our previous study. In present study, to include amorphous phase in this alloy, 3 and 5 wt. % Al has been added in order to improve desirable properties and useful features of MWF and the two alloys have been prepared by suck casting techniques. Microstructure of these alloys have been investigated by optical and electron microscopy which shows occurrence of two different phases, e.g. dark grey and white phases, in (Zr-16 SS)-3 Al and three different phases, e.g. grey, dark grey and white phases in (Zr-16 SS)-5 AI. Electron diffraction and X-ray diffraction (XRD) analyses of these two alloy specimens revealed the occurrence of Zr (Fe, Cr, AI) (dark grey) and Zr 2 (Fe, Cr, AI) (white) phases in (Zr-16 SS)-3 Al whereas, Zr (Fe, Cr, AI) (dark grey), Zr 2 (Fe, Cr, AI) (grey) and Zr 3 (Fe, Cr, AI) (white) phases were found in (Zr-16 SS)-5 AI. In addition, presence of amorphous phase was indicated by XRD analysis that could be confirmed by transmission electron microscopy of these two alloys. (author)

  6. Characterization of structural properties of U and Pu in model systems by advanced synchrotron based X-ray spectroscopy

    International Nuclear Information System (INIS)

    Pidchenko, Ivan

    2016-01-01

    This dissertation presents the investigations of different U and Pu model systems relevant for safety assessment studies of nuclear waste repositories using the X-ray based synchrotron techniques: U and Pu L_3/M_4_,_5 edges HR-XANES, L_3 edge EXAFS and 3d4f RIXS as well as other complementary techniques, including XPS, XRD, SEM, TEM and UV-Vis-NIR techniques.

  7. Contribution to the thermodynamic description of the corium - The U-Zr-O system

    Science.gov (United States)

    Quaini, A.; Guéneau, C.; Gossé, S.; Dupin, N.; Sundman, B.; Brackx, E.; Domenger, R.; Kurata, M.; Hodaj, F.

    2018-04-01

    In order to understand the stratification process that may occur in the late phase of the fuel degradation during a severe accident in a PWR, the thermodynamic knowledge of the U-Zr-O system is crucial. The presence of a miscibility gap in the U-Zr-O liquid phase may lead to a stratified configuration, which will impact the accidental scenario management. The aim of this work was to obtain new experimental data in the U-Zr-O liquid miscibility gap. New tie-line data were provided at 2567 ± 25 K. The related thermodynamic models were reassessed using present data and literature values. The reassessed model will be implemented in the TAF-ID international database. The composition and density of phases potentially formed during stratification will be predicted by coupling current thermodynamic model with thermal-hydraulics codes.

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

    Science.gov (United States)

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

    2016-02-01

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

  9. Microstructural analysis of as-processed U-10 wt.%Mo monolithic fuel plate in AA6061 matrix with Zr diffusion barrier

    Energy Technology Data Exchange (ETDEWEB)

    Perez, E.; Yao, B. [Advanced Materials Processing and Analysis Center, Department of Mechanical, Materials and Aerospace Engineering, University of Central Florida, Orlando, FL 32816 (United States); Keiser, D.D. [Nuclear Fuels and Materials Division, Idaho National Laboratory, Scoville, ID 83415 (United States); Sohn, Y.H., E-mail: ysohn@mail.ucf.ed [Advanced Materials Processing and Analysis Center, Department of Mechanical, Materials and Aerospace Engineering, University of Central Florida, Orlando, FL 32816 (United States)

    2010-07-01

    For higher U-loading in low-enriched U-10 wt.%Mo fuels, monolithic fuel plate clad in AA6061 is being developed as a part of Reduced Enrichment for Research and Test Reactor (RERTR) program. This paper reports the first characterization results from a monolithic U-10 wt.%Mo fuel plate with a Zr diffusion barrier that was fabricated as part of a plate fabrication campaign for irradiation testing in the Advanced Test Reactor (ATR). Both scanning and transmission electron microscopy (SEM and TEM) were employed for analysis. At the interface between the Zr barrier and U-10 wt.%Mo, going from Zr to U(Mo), UZr{sub 2}, {gamma}-UZr, Zr solid-solution and Mo{sub 2}Zr phases were observed. The interface between AA6061 cladding and Zr barrier plate consisted of four layers, going from Al to Zr, (Al, Si){sub 2}Zr, (Al, Si)Zr{sub 3} (Al, Si){sub 3}Zr, and AlSi{sub 4}Zr{sub 5}. Irradiation behavior of these intermetallic phases is discussed based on their constituents. Characterization of as-fabricated phase constituents and microstructure would help understand the irradiation behavior of these fuel plates, interpret post-irradiation examination, and optimize the processing parameters of monolithic fuel system.

  10. Microstructural analysis of as-processed U-10 wt.%Mo monolithic fuel plate in AA6061 matrix with Zr diffusion barrier

    Science.gov (United States)

    Perez, E.; Yao, B.; Keiser, D. D., Jr.; Sohn, Y. H.

    2010-07-01

    For higher U-loading in low-enriched U-10 wt.%Mo fuels, monolithic fuel plate clad in AA6061 is being developed as a part of Reduced Enrichment for Research and Test Reactor (RERTR) program. This paper reports the first characterization results from a monolithic U-10 wt.%Mo fuel plate with a Zr diffusion barrier that was fabricated as part of a plate fabrication campaign for irradiation testing in the Advanced Test Reactor (ATR). Both scanning and transmission electron microscopy (SEM and TEM) were employed for analysis. At the interface between the Zr barrier and U-10 wt.%Mo, going from Zr to U(Mo), UZr 2, γ-UZr, Zr solid-solution and Mo 2Zr phases were observed. The interface between AA6061 cladding and Zr barrier plate consisted of four layers, going from Al to Zr, (Al, Si) 2Zr, (Al, Si)Zr 3 (Al, Si) 3Zr, and AlSi 4Zr 5. Irradiation behavior of these intermetallic phases is discussed based on their constituents. Characterization of as-fabricated phase constituents and microstructure would help understand the irradiation behavior of these fuel plates, interpret post-irradiation examination, and optimize the processing parameters of monolithic fuel system.

  11. Photonuclear reactions of U-233 and Pu-239 near threshold induced by thermal neutron capture gamma rays

    International Nuclear Information System (INIS)

    Moraes, M.A.P. de.

    1990-01-01

    The photonuclear cross sections of U-293 and Pu-239 have been studied by using monochromatic and discrete photons, in the energy interval from 5.49 to 9.72 MeV, produced by thermal neutron capture. The gamma fluxes incident on the samples were measured using a ( 3 x 3 )'' NaI (TI) crystal. The photofission fragments were detected in Makrofol-Kg (SSNTD). A possible structure was observed in the U-233 cross sections, near 7.23 MeV. The relative fissionability of the nuclides was determined at each excitation energy and shown to be energy independent: ( 2.12 ± 0.25) for U-233 and ( 3.32 ± 0.41 ) for Pu-239. The angular distribution of photofission fragments of Pu-239 were measured at two mean excitation energies of 5.43 and 7.35 MeV. An anisotropic distribution of ( 12.2 ± 3.6 ) % was observed at 5.43 MeV. The total neutron cross sections were measured by using a long counter detector. The photoneutron cross sections were calculated by using energy dependent neutron multiplicities values, γ(E), obtained in the literature. The competition Γn/γf was also determined at each excitation energy, and shown to be energy independent: ( 0.54 ± 0.05 ) for U-233 and ( 0.44 ± 0.05 ) for Pu-239, and were correlated to the parameters Z sup(2)/A, ( Ef'-Bn'), A. According to the FUJIMOTO-YAMAGUCHI and CONSTANT NUCLEAR TEMPERATURE models, the nuclear temperatures were calculated. The total photoabsorption cross sections were also calculated as a sum of the photofission and photoneutron cross sections at each energy excitation. From these results the competition Γf/ΓA, called fission probability Pf, were obtained: ( 0.66 ± 0.02) for U-233 and ( 0.70 ± 0.02 ) for Pu-239. (author)

  12. Cross sections and neutron yields for U-233, U-235 and Pu-239 at 2200 m/sec

    Energy Technology Data Exchange (ETDEWEB)

    Sjoestrand, N G; Story, J S

    1960-04-15

    The experimental information on the 2200 m/sec values for {sigma}{sub abs}, {sigma}{sub f}, {alpha}, {nu} and {eta} for {sup 233}U , {sup 235}U and {sup 23} been collected and discussed. The values will later be used in an evaluation of a 'best' set of data. In appendix the isotopic abundances of the uranium isotopes are discussed and also the alpha activities of the uranium isotopes and Pu-239.

  13. Management feature of transuranic for HTGR and LWR

    International Nuclear Information System (INIS)

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

    Long-lived actinides from spent fuels can cause potential long-term environ- mental hazards. The generation and incineration of transuranic in different closed fuel cycles were studied. U and Pu were recycled from spent fuel in the 250 MW high-temperature gas-cooled reactor-pebble-bed-module (HTR-PM) U-Pu fuelled core, and then PuO 2 and MOX fuel elements were designed based on this recycled U and Pu. These fuel elements were used to build up a new PuO 2 or MOX fuelled core with the same geometry of the original reactor. Characteristics of transuranic incineration with HTGR open and closed fuel cycles were studied with VSOP code, and the corresponding results from the light water reactor were compared and analyzed. The transuranic generation with HTGR open fuel cycle is almost half of the corresponding result of the light water reactor. Thus, HTGR closed fuel cycles can effectively burn transuranic. (authors)

  14. Tumor production in Syrian hamsters following inhalation of PuO2--ZrO2 aerosol

    International Nuclear Information System (INIS)

    Thomas, R.G.; Smith, D.M.

    1978-01-01

    Syrian golden hamsters of both sexes were exposed to aerosols of ZrO 2 containing PuO 2 . The starting material in the aerosol generator also had a small amount of 57 Co added as a tracer. The mixture of all three constituents was nebulized and the droplets passed through a heating column at 1000 0 C. Aerosol sampling was accomplished with a cascade impactor and electrostatic precipitator. The median aerodynamic diameters in all inhalation runs were approximately 2 μm with a geometric standard deviation of 2. One exposed group of 60 hamsters had 6-day lung burdens averaging 100 nCi. This group had a lung tumor incidence of 44% with an even distribution of adenomas and carcinomas. Two other groups had average 6-day lung burdens of 80 to 90 nCi plus 55 nCi of intravenously injected spheres localized in the lung. These animals had tumor incidences of approximately 30%

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

    International Nuclear Information System (INIS)

    Bakel, A.J.; Buck, E.C.; Wolf, S.F.; Chamberlain, D.B.; Bates, J.K.; Ebbinghaus, B.B.

    1997-01-01

    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

  16. Biocompatible Ni-free Zr-based bulk metallic glasses with high-Zr-content: Compositional optimization for potential biomedical applications

    Energy Technology Data Exchange (ETDEWEB)

    Hua, Nengbin, E-mail: flower1982cn@126.com [Department of Materials Science and Engineering, Fujian University of Technology, 350118 Fuzhou (China); Department of Materials Science and Engineering, Fuzhou University, 350116 Fuzhou (China); Huang, Lu [Department of Materials Science and Engineering, The University of Tennessee, Knoxville, TN 37996-2100 (United States); Chen, Wenzhe [Department of Materials Science and Engineering, Fujian University of Technology, 350118 Fuzhou (China); Department of Materials Science and Engineering, Fuzhou University, 350116 Fuzhou (China); He, Wei [Department of Materials Science and Engineering, The University of Tennessee, Knoxville, TN 37996-2100 (United States); Department of Mechanical, Aerospace and Biomedical Engineering, The University of Tennessee, Knoxville, TN 37996-2200 (United States); Zhang, Tao [Key Laboratory of Aerospace Materials and Performance (Ministry of Education), School of Materials Science and Engineering, Beihang University, 100191 Beijing (China)

    2014-11-01

    The present study designs and prepares Ni-free Zr{sub 60+x}Ti{sub 2.5}Al{sub 10}Fe{sub 12.5−x}Cu{sub 10}Ag{sub 5} (at.%, x = 0, 2.5, 5) bulk metallic glasses (BMGs) by copper mold casting for potential biomedical application. The effects of Zr content on the in vitro biocompatibility of the Zr-based BMGs are evaluated by investigating mechanical properties, bio-corrosion behavior, and cellular responses. It is found that increasing the content of Zr is favorable for the mechanical compatibility with a combination of low Young's modulus, large plasticity, and high notch toughness. Electrochemical measurements demonstrate that the Zr-based BMGs are corrosion resistant in a phosphate buffered saline solution. The bio-corrosion resistance of BMGs is improved with the increase in Zr content, which is attributed to the enrichment in Zr and decrease in Al concentration in the surface passive film of alloys. Regular cell responses of mouse MC3T3-E1 cells, including cell adhesion and proliferation, are observed on the Zr–Ti–Al–Fe–Cu–Ag BMGs, which reveals their general biosafety. The high-Zr-based BMGs exhibit a higher cell proliferation activity in comparison with that of pure Zr and Ti-6Al-4V alloy. The effects of Zr content on the in vitro biocompatibility can be used to guide the future design of biocompatible Zr-based BMGs. - Highlights: • Ni-free Zr{sub 60+x}Ti{sub 2.5}Al{sub 10}Fe{sub 12.5−x}Cu{sub 10}Ag{sub 5} (at.%, x = 0, 2.5, 5) BMGs were fabricated. • Plasticity and notch toughness of BMGs are enhanced by high-Zr-content. • The high-Zr-based BMGs exhibit excellent bio-corrosion resistance in PBS solution. • The biosafety of BMGs is revealed by regular cell adhesion and proliferation. • High-Zr-bearing BMGs are favorable for potential applications as biomaterials.

  17. Local structure of Th1-xMO2 solid solutions (M = U, Pu)

    International Nuclear Information System (INIS)

    Hubert, S.; Heisbourg, G.; Moisy, Ph.; Dacheux, N.; Purans, J.E.

    2004-01-01

    X-ray absorption spectroscopy of Th 1-x U x O 2 and Th 1-x Pu x O 2 solid solutions was carried out on the Th, U L 3 -edges, and Pu L 3 edge to study the local structure environment of actinide mixed oxides. Various compositions of Th 1-x M x O 2 solid solutions have been prepared through the coprecipitation of the mixed oxalates from chloride or nitrate solutions: x = 0.11, 0.24, 0.37, 0.53, 0.67, 0.81, 0.91 and 1 for Th 1-x U x O 2 , and x = 0.13, 0.32, 0.66 and 1 for Th 1-x Pu x O 2 . They were characterized using X- ray diffraction. XRD analysis allowed to confirm that the variation of the lattice parameters varies linearly with the composition between the end members, suggesting that the atomic volume was conserved regardless of the details of the local distortions of the lattice, following the Vegard's law. Extending X-ray absorption fine structure (EXAFS) provides a direct characterization of the local distortions present in solid solutions. We found that opposite to the lattice parameter obtained by XRD, the interatomic distances given by EXAFS do not follow completely to neither the Vegard's law nor the virtual crystal approximation (VCA). However, the average lattice parameter obtained from EXAFS data for the first and the second shells agrees well with the one calculated from XRD data. (authors)

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

    Science.gov (United States)

    Zheng, Jian; Yamada, Masatoshi

    2005-08-01

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

  19. Phase distribution studies in metallic alloy SIMFUEL

    International Nuclear Information System (INIS)

    Kolay, S.; Basu, M.; Kaity, S.; Das, D.

    2014-01-01

    Utilization of U-Pu based alloy fuel in the three stage nuclear power generation program in India is one of the important mandate due to shorter doubling time for breeding of the fissile isotopes ( 239 Pu and 233 U) to be used in Th based driver fuel in the 3rd stage. Reported information shows successful performance of fuel with porous alloy matrix in achieving 10-15 atom % burn-up. The porosity and microstructure of this alloy are strongly dependent on the composition and phases of the fission products incorporated in the matrix. The porosity influences the extent of fuel swelling and fission gas release, which affects the performance and integrity of the fuel. This study addresses to these issues taking the base alloy U-10wt% Zr

  20. Stress corrosion of the alloy U-7.5 Nb-2.5 Zr

    International Nuclear Information System (INIS)

    Lepoutre, D.; Nomine, A.M.; Miannay, D.

    1983-09-01

    Oxide formed on U-7.5 Nb-2.5 Zr at room temperature during stress corrosion cracking in oxygen is identical to the natural oxide of the alloy. It is formed by UO 2 with Nb and Zr and is associated with an increased Nb content at the interface. This oxide would be responsible for cracking [fr

  1. Performance of cladding on MOX fuel with low 240Pu/239Pu ratio

    International Nuclear Information System (INIS)

    McCoy, K.; Blanpain, P.; Morris, R.

    2015-01-01

    The U.S. Department of Energy has decided to dispose of a portion of its surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. As part of fuel qualification, four lead assemblies were manufactured and irradiated to a maximum fuel rod average burnup of 47.3 MWd/kg heavy metal. This was the world's first commercial irradiation of MOX fuel with a 240 Pu/ 239 Pu 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. (authors)

  2. Influence of air and vacuum environment on fatigue behavior of Zr-based bulk metallic glasses

    International Nuclear Information System (INIS)

    Wang, G.Y.; Liaw, P.K.; Yokoyama, Y.; Peter, W.H.; Yang, B.; Freels, M.; Buchanan, R.A.; Liu, C.T.; Brooks, C.R.

    2007-01-01

    High-cycle fatigue (HCF) experiments in air and vacuum at room temperature were conducted on zirconium (Zr)-based bulk-metallic glasses (BMGs): Zr 50 Cu 40 Al 10 , Zr 50 Cu 30 Al 10 Ni 10 , and Zr 50 Cu 37 Al 10 Pd 3 in atomic percent. The fatigue-endurance limit of Zr 50 Cu 37 Al 10 Pd 3 was found to be significantly greater than those of Zr 50 Cu 40 Al 10 and Zr 50 Cu 30 Al 10 Ni 10 , which indicates that the inclusions of Pd and the resulting nano structures improve the fatigue resistances of the Zr-based BMGs. The fatigue lives in vacuum and air were generally found to be comparable

  3. Is the beagle dog an appropriate experimental animal for extrapolating data to humans on organ distribution patterns of U, Th, and Pu

    International Nuclear Information System (INIS)

    Singh, N.P.; Wrenn, M.E.

    1989-01-01

    Concentrations and organ distribution patterns of alpha-emitting isotopes of U (238U and 234U), Th (232Th, 230Th, and 228Th), and Pu (239,240Pu) were determined for beagle dogs of our colony. The dogs were exposed to environmental levels of U and Th isotopes through ingestion (food and water) and inhalation to stimulate environmental exposures of the general human population. The organ distribution patterns of these radionuclides in beagles are compared to patterns in humans to determine if it is appropriate to extrapolate organ content data from beagles to humans. The results indicated that approximately 80% of the U and Th accumulated in bone in both species. The organ content percentages of these radionuclides in soft tissues such as liver, kidney, etc. of both species were comparable. The human lung contained higher percentages of U and Th than the beagle lung, perhaps because the longer life span of humans resulted in a longer exposure time. If the U and Th content of dog lung is normalized to an exposure time of 58 y and 63 y, median ages of the U and Th study populations, respectively, the lung content for both species is comparable. The organ content of 239,240Pu in humans and beagles differed slightly. In the beagle, the liver contained more than 60%, and the skeleton contained less than 40% of the Pu body content. In humans, the liver contained approximately 37%, and the skeleton contained approximately 58% of the body content. This difference may have been due to differences in the mode of intake of Pu in each species or to differences in the chemical form of Pu. In general, the results suggest that the beagle may be an appropriate experimental animal from which to extrapolate data to humans with reference to the percentage of U, Th, and Pu found in the organs

  4. A study on selective precipitation of U(VI) by hydrophilic cyclic urea derivatives for development of a reprocessing system based on precipitation method

    International Nuclear Information System (INIS)

    Suzuki, Tomoya; Takao, Koichiro; Kawasaki, Takeshi; Harada, Masayuki; Ikeda, Yasuhisa; Nogami, Masanobu

    2014-01-01

    Selective precipitation ability of 2-imidazolidone (EU) and tetrahydro-2-pyrimidinone (PU) for U(VI) species in HNO 3 solutions containing U(VI), U(IV) (simulant of Pu(IV)), and simulated fission products (FPs) was investigated. As a result, it was found that these compounds precipitate almost quantitatively U(VI) as UO 2 (NO 3 ) 2 L 2 (L = EU, PU) from 3.0 M HNO 3 solution. In contrast, these urea derivatives form neither solid precipitates nor oily products with U(IV) in HNO 3 solutions containing only U(IV) species and even in U(VI)-U(IV) admixture system. Therefore, the separation of U(VI) from U(IV) was demonstrated to be achieved in use of EU and PU. Furthermore, EU and PU are capable to remove most of simulated FPs[Sr(II), Ru(III), Rh(III), Re(VII) La(III), Ce(III), Pr(III), Nd(III), and Sm(III)] from U(VI) to give their decontamination factors (DFs) higher than 100, while those values of Zr(IV), Mo(VI), Pd(II), and Ba(II) are necessary to be improved in both systems. From these results, it is expected that EU and PU are the promising precipitants for selective separation of U(VI) from HNO 3 solutions dissolving spent FBR fuels. (author)

  5. An interatomic potential for studying CuZr bulk metallic glasses

    DEFF Research Database (Denmark)

    Paduraru, Anca; Kenoufi, Abdel; Bailey, Nicholas

    2007-01-01

    -scale deformation events and may furthermore involve localization through formation of shear bands. In this paper, an Effective Medium Theory (EMT) potential optimized for modeling the mechanical and thermodynamic properties of CuZr bulk metallic glass is studied. The late transition metals crystallizing in close......The mechanical properties of BMGs are remarkably different from the ones of ordinary metallic alloys due to the atomic level disorder in the glassy state. Unlike crystalline materials plastic deformation in metallic glasses cannot be caused by lattice defects but takes place through atomic...

  6. Response to letter “Electron correlation and relativity of the 5f electrons in the U−Zr alloy system”

    Energy Technology Data Exchange (ETDEWEB)

    Xie, Wei [Materials Science Program, University of Wisconsin-Madison, Madison, WI 53706 (United States); Marianetti, Chris A. [Department of Applied Physics and Applied Mathematics, Columbia University, New York, NY 10027 (United States); Morgan, Dane, E-mail: ddmorgan@wisc.edu [Materials Science Program, University of Wisconsin-Madison, Madison, WI 53706 (United States); Department of Materials Science and Engineering, University of Wisconsin-Madison, Madison, WI 53706 (United States)

    2016-08-01

    In the Letter [Söderlind et al., J. Nucl. Mater. 444, 356 (2014)], Söderlind et al. state their interpretation that 1) we view electron correlation to be strong and including spin-orbit coupling (SOC) to be necessary for U metal and U−Zr alloy in our article [Xiong et al., J. Nucl. Mater. 443, 331 (2013)]. Further, they argue that 2) density functional theory (DFT) without adding the Hubbard U potential, especially when solved using all electron methods, already models U and U−Zr accurately, and 3) adding the Hubbard U potential to DFT in DFT + U models U and U−Zr worse than DFT according to volume, bulk modulus, and magnetic moments predicted from their calculations of the γU phase of elemental U metal. With respect to Söderlind et al.’s interpretation 1), we clarify that our opinions are that U and U−Zr are not strongly, but weakly to moderately correlated and that including SOC is beneficial but not necessary for modeling most ground state properties of U and U−Zr. With respect to Söderlind et al.’s argument 2) we demonstrate that previously neglected and very recent experimental data suggest that DFT in Söderlind's full-potential linear muffin-tin orbital calculations [Söderlind, Phys. Rev. B 66, 085113 (2002)] in fact models the bulk modulus and elastic constants of αU with errors considerably larger than other related elements, e.g., most transition metals. With respect to Söderlind et al.’s argument 3) we argue that they have inappropriately focused on just one phase (the BCC γU phase of U metal), neglecting the other phases which represent the majority of our evidence, and made overgeneralizations based on results at only one U{sub eff} value of 2 eV. We therefore maintain our original conclusion that the accuracy of DFT for modeling U and U−Zr has room for improvement and DFT + U can be of value for this purpose on at least some ground state properties.

  7. Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydin, E-mail: karahan@mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge MA 24-204 (United States)

    2011-07-15

    Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

  8. Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın

    2011-07-01

    Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

  9. Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydin

    2011-01-01

    Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other

  10. Separation of technetium and rare earth metals for co-decontamination process

    International Nuclear Information System (INIS)

    Riddle, Catherine; Martin, Leigh

    2015-01-01

    Poster. In the US there are several technologies under consideration for the separation of the useful components in used nuclear fuel. One such process is the co-decontamination process to separate U, Np and Pu in a single step and produce a Np/ Pu and a U product stream. Although the behavior of the actinide elements is reasonably well defined in this system, the same is not true for the fission products, mainly Zr, Mo, Ru and Tc. As these elements are cationic and anionic they may interact with each other to extract in a manner not predicted by empirical models such as AMUSE. This poster presentation will discuss the initial results of batch contact testing under flowsheet conditions and as a function of varying acidity and flowsheet conditions to optimize recovery of Tc and minimize extraction of Mo, Zr and Ru with the goal of developing a better understanding of the behavior of these elements in the co-decontamination process.

  11. Brazing characteristics of a Zr-Ti-Cu-Fe eutectic alloy filler metal for Zircaloy-4

    Science.gov (United States)

    Lee, Jung G.; Lim, C. H.; Kim, K. H.; Park, S. S.; Lee, M. K.; Rhee, C. K.

    2013-10-01

    A Zr-Ti-Cu-Fe quaternary eutectic alloy was employed as a new Be-free brazing filler metal for Zircaloy-4 to supersede physically vapor-deposited Be coatings used conventionally with several disadvantages. The quaternary eutectic composition of Zr58Ti16Cu10Fe16 (at.%) showing a low melting temperature range from 832 °C to 853 °C was designed by a partial substitution of Zr with Ti based on a Zr-Cu-Fe ternary eutectic system. By applying an alloy ribbon with the determined composition, a highly reliable joint was obtained with a homogeneous formation of predominantly grown α-Zr phases owing to a complete isothermal solidification, exhibiting strength higher than that of Zircaloy-4. The homogenization of the joint was rate-controlled by the diffusion of the filler elements (Ti, Cu, and Fe) into the Zircaloy-4 base metal, and the detrimental segregation of the Zr2Fe phase in the central zone was completely eliminated by an isothermal holding at a brazing temperature of 920 °C for 10 min.

  12. Industrial prospects for the optimized use of U, Pu and Th for sustainable nuclear energy deployment

    International Nuclear Information System (INIS)

    Durpel, Luc Van Den; Guesdon, Bernard; Lecomte, Michel; Greneche, Dominique

    2010-01-01

    'Nuclear energy is part of a sustainable energy future' is a conclusion which is increasingly reached by a variety of energy scenario studies by world-renown institutions such as the IAEA, OECD/IEA and OECD/NEA, World Energy Council, and also reached by different national energy assessment reports. Nuclear does own various unique features that make this energy technology a prime candidate to fulfill a large part of our energy needs, beyond today's use of nuclear energy for our electricity needs. The previous 'wave' of nuclear energy deployment since the 1970's was, and still is, governed by the use of 235 U as main driver to spur this deployment of nuclear energy with gradually the introduction of the U/Pu -cycle in the thermal neutron spectrum reactor park (essentially LWR-technology). Technological progress and good economics of the U/Pu - cycle and especially the economic competitiveness of the LWR's have made this U/Pu-cycle essentially the standard worldwide. Fast spectrum reactors (FRs) haven't yet been developed at sufficient large and industrial scale to make full benefit of the U/Pu-cycle and there are no prospects that the world would massively turn to such FRs in the immediate future. On the verge of a second wave of nuclear deployment, increasing interest is and has to be given to synergies between various nuclear reactor technologies and especially the global nuclear fuel cycle as enabler towards sustainable nuclear energy deployment. Those synergies aim at a reduced reliance on natural uranium resources, continued good economic competitiveness of nuclear energy in local markets, safe and nonproliferant use of nuclear energy, and a reduction of ultimate wastes to be disposed of. This paper provides an overview of various avenues towards sustainable nuclear energy deployment and perspectives from the nuclear industry leader AREVA. (author)

  13. Coulomb effects in isobaric cold fission from reactions 233U(nth,f), 235U(nth,f),239Pu(nth,f) and 252Cf(sf)

    International Nuclear Information System (INIS)

    Montoya, Modesto

    2013-01-01

    The Coulomb effect hypothesis, formerly used to interpret fluctuations in the curve of maximal total kinetic energy as a function of light fragment mass in reactions 233 U(n th ,f), 235 U(n th ,f) and 239 Pu(n th ,f), is confirmed in high kinetic energy as well as in low excitation energy windows, respectively. Data from reactions 233 U(n th ,f), 235 U(n th ,f), 239 Pu(n th ,f) and 252 Cf(sf) show that, between two isobaric fragmentations with similar Q-values, the more asymmetric charge split reaches the higher value of total kinetic energy. Moreover, in isobaric charge splits with different Q-values, similar preference for asymmetrical fragmentations is observed in low excitation energy windows. (author).

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

    Science.gov (United States)

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

    2007-03-28

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

  15. Studies on adsorptions of metallic ions in water by zirconium glyphosate (ZrGP): Behaviors and mechanisms

    International Nuclear Information System (INIS)

    Jia Yunjie; Zhang Yuejuan; Wang Runwei; Fan Faying; Xu Qinghong

    2012-01-01

    A new adsorbent named zirconium glyphosate [Zr(O 3 PCH 2 NHCH 2 COOH) 2 ·0.5H 2 O, denoted as ZrGP] and its selective adsorptions to Pb 2+ , Cd 2+ , Mg 2+ and Ca 2+ ions in water were reported in this paper. Compared to other zirconium adsorbents, such as zirconium phosphate [Zr(HPO 4 ) 2 ], ZrGP exhibited highly selective adsorption to Pb 2+ in solution which contained Pb 2+ , Cd 2+ , Mg 2+ and Ca 2+ ions. The loaded ZrGP with metallic ions can be efficaciously regenerated by aqueous solution of HCl (1.0 M) without any noticeable capacity loss, and almost all of it can be reused and recycled. The memory effect on structural regeneration of ZrGP was also found when Mg 2+ and Ca 2+ were adsorbed. To be specific, the structure of ZrGP was destroyed due to adsorbing these two ions, but it could be regenerated after the loaded materials were dipped in HCl solution (1.0 M) for several minutes to remove metallic ions.

  16. The data-base of properties of actinides for metal fuels

    International Nuclear Information System (INIS)

    Inoue, Tadashi; Kurata, Masateru

    1989-01-01

    It is developed the technology that transuranium elements (TRUs) to be recovered from high active wastes transmute into relatively short lived nuclides by burning them within metallic fuel alloys. In this paper, we collect published data of properties of TRUs and U-Pu(-Zr) alloys and make up the data base for the design study of alloys with TRUs. In addition, the data base possesses a function of statistic analysis in order to facilitate the comparison of data and can afford to estimate properties. This data base collects (1) properties affecting fuel temperature and microstructure, (2) mechanical properties and (3) fundamental properties such as hardness and density, and furthermore, (1) fission gas release, (2) swelling and (3) fuel-cladding interaction and eutectic property as irradiation behavior. (author)

  17. Transient safety performance of the PRISM innovative liquid metal reactor

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Rhow, S.K.; Wu, T.

    1988-01-01

    The PRISM sodium-cooled reactor concept utilizes passive safety characteristics and modularity to increase performance margins, improve licensability, reduce owner's risk and reduce costs. The relatively small size of each reactor module (471 MWt) facilitates the use of passive self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. Key to the transient performance is the inherent negative reactivity feedback characteristics of the core design resulting from the use of metal (U-Pu-Zr) swing, and very low control rod runout worth. Selected beyond design basis events relying only on these core design features are analyzed and the design margins summarized to demonstrate the advancement in reactor safety achieved with the PRISM design concept

  18. Core concepts for 'zero-sodium-void-worth core' in metal fuelled fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.; Hill, R.N.; Fujita, E.K.; Wade, D.C.; Kumaoka, Y.; Suzuki, M.; Kawashima, M.; Nakagawa, H.

    1991-01-01

    Core design options to reduce the sodium void worth in metal fuelled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a 'pancaked' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket-zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. (author)

  19. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States)

    2010-01-31

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium

  20. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydin; Buongiorno, Jacopo

    2010-01-01

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors

  1. Study on direct dissolution of U-10Zr alloy and distribution of uranium and zirconium in liquid cadmium

    International Nuclear Information System (INIS)

    Ye Yuxing; Gao Yuan

    1997-09-01

    The effect of dissolution time, temperature, total surface area of U-10Zr alloy pellets and stirring on the dissolution and dissolution rate of uranium in liquid cadmium were studied. Cadmium containing U and Zr dissolved from U-10Zr alloy at 475 degree C and 500 degree C respectively was analyzed with electron microanalyzer. The experimental results show that at 400 degree and 500 degree C with the stirring rate of some 150 r/min, the solubilities of uranium in liquid cadmium are 0.4% and 2.2%, respectively. At the first 30 min, the dissolution rates of U-10Zr alloy pellets are 0.05 g/(cm 2 ·h) and 0.32 g/(cm 2 ·h), respectively. The suitable dissolution conditions for U-10Zr alloy pellets in liquid cadmium (the ratio of the mass of liquid cadmium to that of the pellets ≅7) are: temperature, about 480 degree C; stirring rate, about 150 r/min; dissolution time, 4 h. The distribution of uranium and zirconium in cadmium is homogeneous

  2. Mono- and di-n-butyl phosphates of some metals in spent nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Solovkin, A.S.

    1982-01-01

    Results of investigations which have been carried out in the Soviet Union for the last 10 years on the determination of the composition, structure, conditions of the formation and solubility of mono- and di-n-butyl phosphates of metals (U/sup 6 +/, Pu/sup 4 +/, Pu/sup 3 +/, Th, Zr, Fe/sup 3 +/, Am, Al, rare-earth elements), which are important for the processes of irradiated nuclear fuel reprocessing, are presented. A conclusion is made that zirconium mono- and di-n-butyl phosphates are the least soluble in aqueous and organic solvents of all investigated compounds. FeA/sub 3/ and AmA/sub 3/ are weakly soluble in aqueous solutions. The other compounds are sufficiently soluble in moderately acidic aqueous solutions or in DBP and TBP with dilutents. The obtained results indicate at the similarity of zirconium and plutonium (4) chemical properties; thorium, in this respect, is not an analogue of plutonium (4). Possible structural formulas of the investigated compounds are considered.

  3. Measurement test on creep strain rate of uranium-zirconium solid solutions

    International Nuclear Information System (INIS)

    Ogata, Takanari; Akabori, Mitsuo; Ogawa, Toru

    1996-11-01

    In order to measure creep strain rate of a small specimen of U-Zr solid solution, authors proposed an estimation method which was based upon the stress relaxation after compression. It was applied to measurement test on creep strain rate of the U-10wt%Zr specimen in the temperature range of 757 to 911degC. It may be concluded that the proposed method is valid, provided that the strain is within the appropriate range and that sufficient amount of the load decrement is observed. The obtained creep rate of U-10wt%Zr alloy indicated significantly smaller value, compared to the experimental data for pure U metal and evaluated data for U-Pu-Zr alloy. However, more careful measurement is desired in future since the present data are thought to be influenced by the precipitations included in the specimen. (author)

  4. Synthesis and evaluation of complexing and extracting properties of diamides for the U(VI)/Pu(IV) partitioning

    International Nuclear Information System (INIS)

    Wahu, S.

    2012-01-01

    This work concerns the synthesis and evaluation of new diamides for co-extraction of U(VI) and Pu(IV) from spent nuclear fuels in concentrated nitric acid medium and the subsequent selective back-extraction of Pu(IV) at lower nitric acidity. Thus, three diamides series, namely R 2 NOC(CH 2 ) n CONR 2 , RR'NOC(CH 2 ) n CONRR' and R 2 NOC(CH 2 CHR')CONR 2 (series 1, 2 and 3, respectively) were synthesized and studied in the expectation to establish a structure-activity relationship regarding the extraction properties of such compounds towards actinides. Extraction tests showed that these diamides can effectively and selectively extract U(VI) and Pu(IV) from Am(III) at high acidity. In addition, it was found that the selective back-extraction of Pu(IV) from U(VI) can be achieved, as expected, merely by changing the acidity, when the nitrogen atoms of these diamides are substituted by branched alkyl chains. The mode of coordination of actinides by the above diamides has also been studied through the use of short chain diamides model compounds, allowing the crystallization of the actinide complexes and, thus, their characterization by X-ray diffraction. EXAFS investigation suggested that the uranium coordination sphere in the complexes formed with long chain diamides and extracted into the organic phase is identical to the one observed in U(VI) complexes obtained with short chain diamides model compounds. However, the stoichiometry of model complexes is 1:1, UO 2 (NO 3 ) 2 TEDA, whereas extracted U(VI) complexes exhibit 1:2 stoichiometry, UO 2 (NO 3 ) 2 (TEHDA) 2 . Finally, Th (IV) complexes were also prepared and characterized and further compared to U(VI) complexes. (author) [fr

  5. Effects of Al addition on atomic structure of Cu-Zr metallic glass

    Science.gov (United States)

    Li, Feng; Zhang, Huajian; Liu, Xiongjun; Dong, Yuecheng; Yu, Chunyan; Lu, Zhaoping

    2018-02-01

    The atomic structures of Cu52Zr48 and Cu45Zr48Al7 metallic glasses (MGs) have been studied by molecular dynamic simulations. The results reveal that the molar volume of the Cu45Zr48Al7 MG is smaller than that of the Cu52Zr48 MG, although the size of the Al atom is larger than that of the Cu atom, implying an enhanced atomic packing density achieved by introducing Al into the ternary MG. Bond shortening in unlike atomic pairs Zr-Al and Cu-Al is observed in the Cu45Zr48Al7 MG, which is attributed to strong interactions between Al and (Zr, Cu) atoms. Meanwhile, the atomic packing efficiency is enhanced by the minor addition of Al. Compared with the Cu52Zr48 binary MG, the potential energy of the ternary MG decreases and the glass transition temperature increases. Structural analyses indicate that more Cu- and Al-centered full icosahedral clusters emerge in the Cu45Zr48Al7 MG as some Cu atoms are substituted by Al. Furthermore, the addition of Al leads to more icosahedral medium-range orders in the ternary MG. The increase of full icosahedral clusters and the enhancement of the packing density are responsible for the improved glass-forming ability of Cu45Zr48Al7.

  6. Synthesis of (U,Zr)C solid solutions under exothermic conditions

    International Nuclear Information System (INIS)

    Wang, L.L.; Moore, H.G.; Gladson, J.W.

    1993-01-01

    The reactions of forming (U,Zr)C solid solutions from their elemental components or similarly less stable reactants such as UC 2 are strongly exothermic due to the high stability of these solid solutions. A simple approach of utilizing this heat of formation energy to assist the solid solution reaction process is to intimately mix the less stable reactant powders and then pressed them into a compact. The compact is then heated to the ignition temperature of the reaction. The feasibility of this reaction method to synthesize (U,Zr)C solid solutions has been demonstrated in this study. The preliminary results also show that both the initial composition and the heating rate have a significant effect on the nature of the reaction process. As expected the degree of powder mixing was also found to affect the completeness of the reaction

  7. Fast reactor of 1.000 MWt started with U-Zr

    International Nuclear Information System (INIS)

    Nascimento, J.A.; Ishiguro, Y.

    1990-01-01

    A U-Zr fueled 1000 MWt liquid metal reactor (LMR) to be used in a second step of the fast breeder reactor development program that we propose for Brazil is studied. Initially, principal technological aspects and cost trends are reviewed in order to place this type of reactors in a proper perspective regarding their application to electric power generation. Then two models are compared and one is selected for cycle-by-cycle analysis isotopic evolution and parameters of interest such as the Doppler effect, sodium void reactivity, control requirement and availability, resources consumption, and enrichment requirement. The analysed model is quite adequate for the phase for which it is considered due to its high degree of inherent safety, which should contribute to a better public acceptance of nuclear energy. In addition, its introduction with enriched uranium, available in the country, allows an autonomous development of LMR which is a better alternative to the PWR meeting for future power demand. (author)

  8. Synthesis and evaluation of N,N-di-alkyl-2-methoxyacetamides for the separation of U(VI) and Pu(IV) from nitric acid medium

    Energy Technology Data Exchange (ETDEWEB)

    Kumaresan, R.; Prathibha, T.; Selvan, B. Robert; Venkatesan, K.A.; Antony, M.P. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India). Fuel Chemistry Div.

    2017-07-01

    The homologs of N,N-di-alkyl-2-methoxyacetamides (DAMeOA) having three different alkyl chains varying from hexyl to decyl (C{sub 6}, C{sub 8} and C{sub 10}) were synthesized and characterized by NMR and IR spectral analyses. Extraction behavior of U(VI) and Pu(IV) from nitric acid medium in a solution of 0.5 M of DAMeOA in n-dodecane (n-DD) was studied and the results were compared with those obtained using N,N-di-hexyloctanamide (DHOA) in n-dodecane. The effect of various parameters on the distribution ratio of U(VI) and Pu(IV) in DAMeOA was studied. The extraction of nitric acid increased with decrease in chain length of alkyl group attached to amidic nitrogen atom of DAMeOA and the conditional nitric acid extraction constant was determined. The extraction of nitric acid in DAMeOA/n-DD resulted in the formation of third phase in organic phase and the third phase occurred early with DAMeOA having smaller alkyl chain length. In contrast to this, the distribution ratio (D) of U(VI) and Pu(IV) in DAMeOA/n-DD increased with increase in the concentration of nitric acid and with increase in the chain length of alkyl group attached to amidic nitrogen atom of DAMeOA. The stoichiometry of the metal - solvate was determined from the slope of extraction data. Quantitative recovery of uranium and plutonium from the loaded organic phase was achieved using dilute nitric acid.

  9. New micrographic examination methods for nuclear metals and alloys; Methodes micrographiques originales pour l'examen de differents metaux et alliages nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Monti, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Compagnie Saint Gobain, Services Nucleaires (France)

    1959-07-01

    The report describes several methods of preparation of specimens (polishing and pickling) and the results obtained on metals: Pu, U, Th, Nb, Ta, V and alloys: Al-Pu, U-Pu and Th-Nb. (author) [French] On decrit diverses methodes de preparation d'echantillons metallographiques (polissage et attaque) ainsi que les resultats obtenus sur les metaux: Pu, U, Th, Nb, Ta, V et les alliages: Al-Pu, U-Pu et Th-Nb. (auteur)

  10. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Kramer, J.M.

    1992-01-01

    The next step in the development of metal fuels for the integral fast reactor (IFR) is the conversion of the Experimental Breeder Reactor II (EBR-II) core to one containing the ternary U-20 Pu-10 Zr alloy clad with HT-9 cladding, i.e., the Mk-V core. This paper presents results of three hot-cell furnace simulation tests on irradiated Mk-V-type fuel elements (U-19 Pu-10 Zr/HT-9), which were performed to support the safety case for the Mk-V core. These tests were designed to envelop an umbrella (bounding) unlikely loss-of-flow (LOF) event in EBR-II during which the calculated peak cladding temperature would reach 776 degree C for < 2 min. The principal objectives of these tests were (a) demonstration of the safety margin of the fuel element, (b) investigation of cladding breaching behavior, and (c) provision of data for validation of the FPIN2 and LIFE-METAL codes

  11. Oxygen diffusion model of the mixed (U,Pu)O{sub 2±x}: Assessment and application

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Emily, E-mail: eemevans@gmail.com [Den-Service de la Corrosion et du Comportement des Matériaux dans leur Environment (SCCME), CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette, Cedex (France); Guéneau, Christine [Den-Service de la Corrosion et du Comportement des Matériaux dans leur Environment (SCCME), CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette, Cedex (France); Crocombette, Jean-Paul [Den-Service de Recherche de Métallurgie Physique (SRMP), CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette, Cedex (France)

    2017-03-15

    The uranium-plutonium (U,Pu)O{sub 2±x} mixed oxide (MOX) is used as a nuclear fuel in some light water reactors and considered for future reactor generations. To gain insight into fuel restructuring, which occurs during the fuel lifetime as well as possible accident scenarios understanding of the thermodynamic and kinetic behavior is crucial. A comprehensive evaluation of thermo-kinetic properties is incorporated in a computational CALPHAD type model. The present DICTRA based model describes oxygen diffusion across the whole range of plutonium, uranium and oxygen compositions and temperatures by incorporating vacancy and interstitial migration pathways for oxygen. The self and chemical diffusion coefficients are assessed for the binary UO{sub 2±x} and PuO{sub 2−x} systems and the description is extended to the ternary mixed oxide (U,Pu)O{sub 2±x} by extrapolation. A simulation to validate the applicability of this model is considered.

  12. Microstructure of bonding interface for resistance welding of Zr-based metallic glass sheets

    International Nuclear Information System (INIS)

    Kuroda, Toshio; Ikeuchi, Kenji; Shimada, Masahiro; Kobayashi, Akira; Kimura, Hisamichi; Inoue, Akihisa

    2009-01-01

    Resistance welding of Zr 55 Cu 30 Al 10 Ni 5 metallic glass sheets was investigated at 723 K in a supercooled liquid region. The welding time was changed from 5 s to 20 s at 723 K. The joint interface of the metallic glass was no defect and no crack. X-ray diffraction technique of the bonding interface of specimens was performed. The specimens showed halo patterns showing existence of only glassy phase, when the welding time was 5 s and 10 s. X-ray diffraction patterns of specimen bonded for 20 s showed crystalline peaks with halo patterns for the welding for 20 s. The crystalline phase at the bonding interface was small. Transmission electron micrograph at the bonding interface showed nanostructures of NiZr 2 and Al 5 Ni 3 Zr 2 . (author)

  13. Investigation of tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2

    International Nuclear Information System (INIS)

    Yildiz, K.; Sahin, S.; Sahin, H. M.; Acir, A.; Yalcin, S.; Altinok, T.; Bayrak, M.; Alkan, M.; Durukan, O.

    2007-01-01

    In the world, thorium reserves are three times more than natural Uranium reserves. It is planned in the near future that nuclear reactors will use thorium as a fuel. Thorium is not a fissile isotope because it doesn't make fission with thermal neutrons so it could be converted to 2 33U isotope which has very high quality fission cross-section with thermal neutrons. 2 33U isotope can be used in present LWRs as an enrichment fuel. In the fusion reactors, tritium is the most important fossil fuel. Because tritium is not natural isotope, it has to be produced in the reactor. The purpose of this work is to investigate the tritium and 2 33U breeding in a fission-fusion hybrid reactor fuelling with ThO 2 for Δt=10 days during a reactor operation period in five years. The neutronic analysis is performed on an experimental hybrid blanket geometry. In the center of the hybrid blanket, there is a line neutron source in a cylindrical cavity, which simulates the fusion plasma chamber where high energy neutrons (14.1 MeV) are produced. The conventional fusion reaction delivers the external neutron source for blankets following, 2 D + 3 T →? 4 He (3.5 MeV) + n (14.1 MeV). (1) The fuel zone made up of natural-ThO 2 fuel and it is cooled with different coolants. In this work, five different moderator materials, which are Li 2 BeF 4 , LiF-NaF-BeF 2 , Li 2 0Sn 8 0, natural Lithium and Li 1 7Pb 8 3, are used as coolants. The radial reflector, called tritium breeding zones, is made of different Lithium compounds and graphite in sandwich structure. In the work, eight different Lithium compounds were used as tritium breeders in the tritium breeding zones, which are Li 3 N, Li 2 O, Li 2 O 2 , Li 2 TiO 3 , Li 4 SiO 3 , Li 2 ZrO 3 , LiBr and LiH. Neutron transport calculations are conducted in spherical geometry with the help of SCALE4.4A SYSTEM by solving the Boltzmann transport equation with code CSAS and XSDRNPM, under consideration of unresolved and resolved resonances, in S 8 -P 3

  14. The uptake of Cs, Cl, I, Co, U and Pu by C-S-H, portlandite and ettringite

    International Nuclear Information System (INIS)

    Le Callonnec, C.; Richet, C.; Ayache, R.

    2000-01-01

    The aim of this work is to study the retention of Pu, Cs, Cl, U, Co and I onto the three phases of a cement paste: calcium silicate hydrates or C-S-H (Ca/Si=1.65), portlandite (Ca(OH) 2 ) and ettringite (Ca 6 Al 2 S 3 O 21 .32H 2 O). C-S-H of Ca/Si molar ratio of 1.65 and ettringite were prepared in suspension, while portlandite was provided by an industrial laboratory. The sorptive power for each ion was determined by a batch model at 20 degrees Celsius. The uptake of ions onto the phases has been obtained by introducing a very small volume of radioactive Cs, Co, Cl, I, U and Pu solution in each batch. Aliquots of solution were periodically removed for analysis. The results show that there are two kinds of ions whatever the solid phase: those which are poorly sorbed: Cs, Cl, I and Co, U and Pu which are strongly sorbed (by a factor 100 to 10000). Elsewhere, the comparison between the sorptive power of each phase reveals that C-S-H develops a significant sorption potential for pluri-valent ions (Pu, U, Co) as well as for Cl. The retention capacity of portlandite mainly towards pluri-valent cations is one or two orders of magnitude weaker than C-S-H capacity. Ettringite only plays a sorptive role relatively to Cs. However, iodine does not have a strong affinity for any solid at all. The data allow us to calculate a global sorption coefficient Kd of the cement paste for each ion. Assuming that the proportion of C-S-H, portlandite, ettringite is respectively 65%, 15% and 5% in a Portland cement, the global sorption coefficient is in good agreement with literature regarding on U, Cs, I and Cl. Cementitious materials appear to be an interesting chemical barrier that can limit the diffusion of pluri-valent cations. As C-S-H represent from 60% to 75% of the cement paste, this study leads to underline that C-S-H is mainly responsible for the retention capacity of cementitious materials specially towards U, Pu, Co, Cl although its sorption power is limited concerning

  15. Pu'u 'Ō'ō-Kūpaianaha eruption of Kilauea, November 1991-February 1994; field data and flow maps

    Science.gov (United States)

    Heliker, C. Christina; Mangan, Margaret T.; Mattox, Tari N.; Kauahikaua, James P.

    1998-01-01

    The Pu'u 'Ō'ō-Kūpaianaha eruption on the east rift zone of Kīlauea, which began in January 1983, is the longest-lived rift zone eruption of the last two centuries. By 1994, a broad field of lava, nearly 1 km3 in volume and 12 km wide at the coast, had buried 87 km2 of the volcano's south flank. The initial six months of fissure eruptions (episodes 1-3) were followed by three years of episodic lava fountaining from the Pu'u 'Ō'ō vent (episodes 4–47). In July 1986, after two days of fissure eruptions up- and downrift from Pu'u 'Ō'ō (episodes 48a and 48b), the eruption shifted to a new vent, Kūpaianaha, 3.5 km downrift. For the next five-and-a-half years (episode 48), Kūpaianaha was the site of nearly continuous low-level effusion. The 49th episode occurred in November 1991, when several fissures opened between Pu'u 'Ō'ō and Kūpaianaha (see Mangan and others, 1995, Bulletin of Volcanology, v. 57, p. 127-135). This three-week-long outburst was the result of the waning output of the Kūpaianaha vent, which finally died in February 1992 (see Kauahikaua and others, 1996, Bulletin of Volcanology, v. 57, p. 641-648). The third epoch of the eruption began ten days later, when vents opened on the uprift slope of the Pu'u 'Ō'ō cone. Several flank vents erupted over the next two years (episodes 50-53). In the first year, from February 1992 through February 1993, the low-level effusion was interrupted by 21 brief pauses. These ended with the beginning of episode 53 in February 1993, and for the next year, lava effusion was continuous. Episode 53 was ongoing at the end of the interval covered by this report. During the years that Kūpaianaha was active, the Pu'u 'Ō'ō conduit gradually evolved into a crater 300 m in diameter as the conduit walls collapsed. Beginning in 1987, an active lava pond was intermittently visible in the bottom of the crater; from 1990 on, the pond was almost continuously present. The Pu'u 'Ō‘ō pond drained at the beginning of episode

  16. Life-cycle impacts from novel thorium–uranium-fuelled nuclear energy systems

    International Nuclear Information System (INIS)

    Ashley, S.F.; Fenner, R.A.; Nuttall, W.J.; Parks, G.T.

    2015-01-01

    Highlights: • LCA performed for three open cycle Th–U-fuelled nuclear energy systems. • LCA for open cycle U-fuelled nuclear energy system (Areva’s EPR) used as benchmark. • U-fuelled EPR had lowest emissions per kWh over all systems studied in this work. • LCA model developed for thorium recovered from monazitic beach sands. • LCA model developed for the production of heavy water. - Abstract: Electricity generated from nuclear power plants is generally associated with low emissions per kWh generated, an aspect that feeds into the wider debate surrounding nuclear power. This paper seeks to investigate how life-cycle emissions would be affected by including thorium in the nuclear fuel cycle, and in particular its inclusion in technologies that could prospectively operate open Th–U-based nuclear fuel cycles. Three potential Th–U-based systems operating with open nuclear fuel cycles are considered: AREVA’s European Pressurised Reactor; India’s Advanced Heavy Water Reactor; and General Atomics’ Gas-Turbine Modular Helium Reactor. These technologies are compared to a reference U-fuelled European Pressurised Reactor. A life-cycle analysis is performed that considers the construction, operation, and decommissioning of each of the reactor technologies and all of the other associated facilities in the open nuclear fuel cycle. This includes the development of life-cycle analysis models to describe the extraction of thorium from monazitic beach sands and for the production of heavy water. The results of the life-cycle impact analysis highlight that the reference U-fuelled system has the lowest overall emissions per kWh generated, predominantly due to having the second-lowest uranium ore requirement per kWh generated. The results highlight that the requirement for mined or recovered uranium (and thorium) ore is the greatest overall contributor to emissions, with the possible exception of nuclear energy systems that require heavy water. In terms of like

  17. Superconductivity and spin fluctuations in M-Zr metallic glasses (M = Cu, Ni, Co, and Fe)

    International Nuclear Information System (INIS)

    Altounian, Z.; Strom-Olsen, J.O.

    1983-01-01

    The superconducting transition temperature, upper critical field, and magnetic susceptibility have been measured in four binary metallic glass systems: Cu-Zr, Ni-Zr, Co-Zr, and Fe-Zr. For each alloy system, a full and continuous range of Zr-rich compositions accessible by melt spinning has been examined. For Cu-Zr, the range is 0.75>x>0.30; for Ni-Zr, 0.80>x>0.30; for Co-Zr, 0.80>x>0.48, and for Fe-Zr, 0.80>x>0.55 (x being the concentration of Zr in at. %). The results show clearly the influence of spin fluctuations in reducing the superconducting transition temperature. The data have been successfully analyzed using a modified form of the McMillan equation together with expressions for the Stoner enhanced magnetic susceptibility and the Ginsburg-Landau-Abrikosov-Gor'kov expression for the upper critical field

  18. Metallic fuel development

    International Nuclear Information System (INIS)

    Walters, L.C.

    1987-01-01

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

  19. Concept of the new generation high safety liquid metal reactor (LMFR)

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Zverkov, Y.A.; Morozov, A.G.; Orlov, V.V.; Ponomarev-Stepnoi, N.N.; Proshkin, A.A.; Slesarev, I.S.; Subbotin, S.A.

    1988-01-01

    The comparative analysis of the inner stability of the liquid metal reactors to severe accidents was made using the asymptotic reactivity balance. The group of the BN-reactors, Superphenix, IFR, LMFR were considered. This paper lists the characteristics of the reactors, used in the self-protectiveness analysis. The authors present the maximum coolant temperatures in post-accident asymptotic state for IFRs as on of the possible designs of a high safety fast reactor with metal fuel, U-Pu-Zr and LMFR. As is known, these values are very important for assessment of the ATWS accidence consequences. The authors consider the following situations and their combinations: loss of reactor coolant flow-LOFWS, loss of heat sink-LOHSWS, uncontrolled reactor sodium overcooling (down to the freezing point)-OVCWS, uncontrolled excess reactivity insertion-TOPWS. The calculation results demonstrate a high stability of the IFR and LMFR reactors to the most severe accidence sequences

  20. New Fuel Alloys Seeking Optimal Solidus and Phase Behavior for High Burnup and TRU Burning

    International Nuclear Information System (INIS)

    Blackwood, V.S.; Jones, Z.S.; Olson, D.L.; Mishra, B.; Mariani, R.D.; Porter, D.L.; Kennedy, J.R.; Hayes, S.L.

    2013-01-01

    Summary: • Pd will bind lanthanide fission products. • 2 wt% Pd in alloy is expected to allow 20 at% Heavy Metal burnup, 4 wt% Pd possibly 30-40 at% HM burnup. • For recycled fuel with some lanthanide carryover, palladium additive will also prevent premature FCCI. • Novel uranium alloy systems suitable for burning transuranics were identified. • U-Mo-Ti-Zr and U-W-Mo irradiations may perform comparably to U-10Zr, but the real tests needed must include Pu and Np for TRU burning. – Diffusion couples with alloys and Fe or cladding; – Irradiations

  1. Raman spectrometric determination of Pu(VI) and Pu(V) in nitric acid solutions

    International Nuclear Information System (INIS)

    Gantner, E.; Freudenberger, M.; Steinert, D.; Ache, H.J.

    1987-03-01

    The determination of Pu(VI) in nitric acid solutions by spontaneous Laser Raman Spectrometry (LRS) was investigated and a calibration curve was established using U(VI) as internal standard. In addition, the concentrations of Pu(VI) and Pu(V) as a function of time were measured by this method in Pu(VI) solutions of different acidity containing H 2 O 2 as the reducing agent. In solutions which are intensely coloured by the presence of Ru(NO) complexes Pu(VI) can also be determined by LRS using a Kr + laser as excitation source. In future experiments, the study of the Pu(IV)-interaction with Ru using LRS and spectrophotometry as analytical techniques is therefore intended. (orig.) [de

  2. Static electric dipole polarizabilities of tri- and tetravalent U, Np, and Pu ions.

    Science.gov (United States)

    Parmar, Payal; Peterson, Kirk A; Clark, Aurora E

    2013-11-21

    High-quality static electric dipole polarizabilities have been determined for the ground states of the hard-sphere cations of U, Np, and Pu in the III and IV oxidation states. The polarizabilities have been calculated using the numerical finite field technique in a four-component relativistic framework. Methods including Fock-space coupled cluster (FSCC) and Kramers-restricted configuration interaction (KRCI) have been performed in order to account for electron correlation effects. Comparisons between polarizabilities calculated using Dirac-Hartree-Fock (DHF), FSCC, and KRCI methods have been made using both triple- and quadruple-ζ basis sets for U(4+). In addition to the ground state, this study also reports the polarizability data for the first two excited states of U(3+/4+), Np(3+/4+), and Pu(3+/4+) ions at different levels of theory. The values reported in this work are the most accurate to date calculations for the dipole polarizabilities of the hard-sphere tri- and tetravalent actinide ions and may serve as reference values, aiding in the calculation of various electronic and response properties (for example, intermolecular forces, optical properties, etc.) relevant to the nuclear fuel cycle and material science applications.

  3. Electronic structure and magnetic properties of dilute U impurities in metals

    Science.gov (United States)

    Mohanta, S. K.; Cottenier, S.; Mishra, S. N.

    2016-05-01

    The electronic structure and magnetic moment of dilute U impurity in metallic hosts have been calculated from first principles. The calculations have been performed within local density approximation of the density functional theory using Augmented plane wave+local orbital (APW+lo) technique, taking account of spin-orbit coupling and Coulomb correlation through LDA+U approach. We present here our results for the local density of states, magnetic moment and hyperfine field calculated for an isolated U impurity embedded in hosts with sp-, d- and f-type conduction electrons. The results of our systematic study provide a comprehensive insight on the pressure dependence of 5f local magnetism in metallic systems. The unpolarized local density of states (LDOS), analyzed within the frame work of Stoner model suggest the occurrence of local moment for U in sp-elements, noble metals and f-block hosts like La, Ce, Lu and Th. In contrast, U is predicted to be nonmagnetic in most transition metal hosts except in Sc, Ti, Y, Zr, and Hf consistent with the results obtained from spin polarized calculation. The spin and orbital magnetic moments of U computed within the frame of LDA+U formalism show a scaling behavior with lattice compression. We have also computed the spin and orbital hyperfine fields and a detail analysis has been carried out. The host dependent trends for the magnetic moment, hyperfine field and 5f occupation reflect pressure induced change of electronic structure with U valency changing from 3+ to 4+ under lattice compression. In addition, we have made a detailed analysis of the impurity induced host spin polarization suggesting qualitatively different roles of f-band electrons on moment stability. The results presented in this work would be helpful towards understanding magnetism and spin fluctuation in U based alloys.

  4. Studies on adsorptions of metallic ions in water by zirconium glyphosate (ZrGP): Behaviors and mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Jia Yunjie; Zhang Yuejuan [State Key Laboratory of Chemical Resource Engineering, Beijing University of Chemical Technology, Box. 98, No.15, Beisanhuan donglu, Beijing 100029 (China); Wang Runwei [State Key Laboratory of Inorganic Synthesis and Preparative Chemistry, Jilin University, Changchun 130012 (China); Fan Faying [State Key Laboratory of Chemical Resource Engineering, Beijing University of Chemical Technology, Box. 98, No.15, Beisanhuan donglu, Beijing 100029 (China); Xu Qinghong, E-mail: xuqh@mail.buct.edu.cn [State Key Laboratory of Chemical Resource Engineering, Beijing University of Chemical Technology, Box. 98, No.15, Beisanhuan donglu, Beijing 100029 (China)

    2012-01-15

    A new adsorbent named zirconium glyphosate [Zr(O{sub 3}PCH{sub 2}NHCH{sub 2}COOH){sub 2}{center_dot}0.5H{sub 2}O, denoted as ZrGP] and its selective adsorptions to Pb{sup 2+}, Cd{sup 2+}, Mg{sup 2+} and Ca{sup 2+} ions in water were reported in this paper. Compared to other zirconium adsorbents, such as zirconium phosphate [Zr(HPO{sub 4}){sub 2}], ZrGP exhibited highly selective adsorption to Pb{sup 2+} in solution which contained Pb{sup 2+}, Cd{sup 2+}, Mg{sup 2+} and Ca{sup 2+} ions. The loaded ZrGP with metallic ions can be efficaciously regenerated by aqueous solution of HCl (1.0 M) without any noticeable capacity loss, and almost all of it can be reused and recycled. The memory effect on structural regeneration of ZrGP was also found when Mg{sup 2+} and Ca{sup 2+} were adsorbed. To be specific, the structure of ZrGP was destroyed due to adsorbing these two ions, but it could be regenerated after the loaded materials were dipped in HCl solution (1.0 M) for several minutes to remove metallic ions.

  5. Local atomic structure of α-Pu

    International Nuclear Information System (INIS)

    Espinosa, F. J.; Villella, P.; Lashley, J. C.; Conradson, S. D.; Cox, L. E.; Martinez, R.; Martinez, B.; Morales, L.; Terry, J.; Pereyra, R. A.

    2001-01-01

    The local atomic structure of α-Pu was investigated using x-ray absorption fine structure (XAFS) spectroscopy. XAFS spectra were obtained for a zone-refined α-Pu and the results were compared to 32-year-old and Ce-doped (0.34 at.%) samples. X-ray diffraction (XRD) patterns were also measured for the zone-refined and 32-year-old materials. The extent of the Bragg peaks showed that amorphization of the 32-year-old sample had not occurred despite the prolonged exposure to self-radiation. Analogous to metastable δ-Pu alloys, the local atomic structure around Pu for the zone-refined material shows the possible presence of noncrystallographic Pu-Pu distances. Conversely, the Ce and the 32-year-old sample show no evidence for such noncrystallographic distances. Disorder in the Pu local environment was found to be impurity dependent. The Ce-doped sample presented a larger Pu-Pu nearest neighbor disorder than the aged sample, although the total amount of Am, U, and He impurities was actually higher in the aged sample. The local environment around U and Ce impurities is consistent with these elements being in substitutional lattice sites. In addition, U and Ce do not introduce significant lattice distortion to their nearest neighbors. This is consistent with disorder being more related to the perturbation of the coupling between the electronic and crystal structure, or the Peierls--Jahn-Teller distortion that generates the monoclinic α-Pu structure, and less to strain fields produced in the vicinity of the impurities

  6. Formation of Ti--Zr--Cu--Ni bulk metallic glasses

    International Nuclear Information System (INIS)

    Lin, X.H.; Johnson, W.L.

    1995-01-01

    Formation of bulk metallic glass in quaternary Ti--Zr--Cu--Ni alloys by relatively slow cooling from the melt is reported. Thick strips of metallic glass were obtained by the method of metal mold casting. The glass forming ability of the quaternary alloys exceeds that of binary or ternary alloys containing the same elements due to the complexity of the system. The best glass forming alloys such as Ti 34 Zr 11 Cu 47 Ni 8 can be cast to at least 4-mm-thick amorphous strips. The critical cooling rate for glass formation is of the order of 250 K/s or less, at least two orders of magnitude lower than that of the best ternary alloys. The glass transition, crystallization, and melting behavior of the alloys were studied by differential scanning calorimetry. The amorphous alloys exhibit a significant undercooled liquid region between the glass transition and first crystallization event. The glass forming ability of these alloys, as determined by the critical cooling rate, exceeds what is expected based on the reduced glass transition temperature. It is also found that the glass forming ability for alloys of similar reduced glass transition temperature can differ by two orders of magnitude as defined by critical cooling rates. The origins of the difference in glass forming ability of the alloys are discussed. It is found that when large composition redistribution accompanies crystallization, glass formation is enhanced. The excellent glass forming ability of alloys such as Ti 34 Zr 11 Cu 47 Ni 8 is a result of simultaneously minimizing the nucleation rate of the competing crystalline phases. The ternary/quaternary Laves phase (MgZn 2 type) shows the greatest ease of nucleation and plays a key role in determining the optimum compositions for glass formation. copyright 1995 American Institute of Physics

  7. Phonon dispersion of metallic glass CuZr{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Nakashima, S [Department of Condensed Matter Chemistry and Physics, Graduate School of Sciences, Kyushu University, 4-2-1 Ropponmatsu, Chuo-ku, Fukuoka 810-8560 (Japan); Kawakita, Y [Department of Physics, Faculty of Sciences, Kyushu University, 4-2-1 Ropponmatsu, Chuo-ku, Fukuoka 810-8560 (Japan); Otomo, T [Japan Proton Accelerator Research Complex, High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba 305-0801, Japan (Japan); Suenaga, R [Department of Condensed Matter Chemistry and Physics, Graduate School of Sciences, Kyushu University, 4-2-1 Ropponmatsu, Chuo-ku, Fukuoka 810-8560 (Japan); Baron, A Q R [Materials Dynamics Laboratory, Harima RIKEN, 1-1-1 Kouto, Sayo-cho, Sayo-gun, Hyogo 679-5198 (Japan); Tsutsui, S [Japan Synchrotron Radiation Research Institute (JASRI, SPring-8), 1-1-1 Kouto, Sayo-cho, Sayo-gun, Hyogo 679-5198, Japan (Japan); Kohara, S [Japan Synchrotron Radiation Research Institute (JASRI, SPring-8), 1-1-1 Kouto, Sayo-cho, Sayo-gun, Hyogo 679-5198, Japan (Japan); Takeda, S [Department of Physics, Faculty of Sciences, Kyushu University, 4-2-1 Ropponmatsu, Chuo-ku, Fukuoka 810-8560 (Japan); Itoh, K [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennnan-gun, Osaka 590-0494 (Japan); Kato, H [Institute for Material Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai 980-8577 (Japan); Fukunaga, T [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennnan-gun, Osaka 590-0494 (Japan); Hasegawa, M [Department of Materials, Physics and Energy Engineering, Graduate School of Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya, 464-8603 (Japan)

    2007-12-15

    Collective dynamics of metallic glass CuZr{sub 2} has been studied in the first pseudo Brillouin zone using high-resolution inelastic X-ray scattering. Acoustic-like longitudinal propagating excitations were observed and the dispersion relation was determined. In addition of longitudinal mode, transverse mode with half excitation energy contributes to medium energy-transfer region.

  8. Kinetics and thermodynamics of the dissolution of Th1-xMxO2 solid solutions (M = U, Pu)

    International Nuclear Information System (INIS)

    Hubert, S.; Heisbourg, G.; Dacheux, N.; Moisy, Ph.; Purans, J.

    2004-01-01

    Kinetics of the dissolution of Th 1-x M x O 2 (M = U, Pu) solid solutions was investigated as a function of several chemical parameters such as pH, substitution ratio, temperature, ionic strength, and electrolyte. Several compositions of Th 1-x U x O 2 and Th 1-x Pu x O 2 were synthesized and characterized before and after leaching by using several methods such as XRD, EXAFS, BET, PIXE, SEM, and XPS. Leaching tests were performed in nitric, hydrochloric or sulfuric media and groundwater. The normalized dissolution rates were evaluated for Th 1-x U x O 2 , and Th 0.88 Pu 0.12 O 2 leading to the determination of the partial order related to the proton concentration, n, and to the corresponding normalized dissolution rate constant at pH = 0, k'T. While for Th enriched solids, the solid solutions Th 1-x U x O 2 have the same dissolution behaviour than ThO 2 with a partial order n ∼ 0.3, in the case of uranium enriched solids, Th 1-x U x O 2 has the same dissolution behaviour than UO 2 with a partial order of n = 1, indicating that uranium oxidation rate becomes the limiting step of the dissolution process. The stoichiometry of the release of both actinides (U or Pu, Th) was verified until the precipitation of thorium occurred in the leachate for pH > 2, while uranium was released in the solution as an uranyl form. For uranium enriched solid solutions, thermodynamic equilibrium was reached after 100 days, and solubility constant of secondary phase was determined. In the case of Th 1-x Pu x O 2 , the dissolution behaviour is similar to that of ThO 2 , but only kinetic aspect of the dissolution can be studied. From the analysis of XPS and EXAFS data on leached and un-leached Th 1-x U x O 2 samples, the dissolution mechanism of solid solutions was explained and will be discussed. The role of the electrolytes on the dissolution of the solid solutions is discussed. Kinetics parameters of dissolution are also given in groundwater and in neutral media

  9. Migration of plutonium from freshwater ecosystem at Hanford. [/sup 238/Pu, /sup 239/Pu, /sup 240/Pu

    Energy Technology Data Exchange (ETDEWEB)

    Emery, R. M.; Klopfer, D. C.; McShane, M. C.

    1977-09-01

    A reprocessing waste pond at Hanford has been inventoried to determine quantities of plutonium (Pu) that have been accumulated since its formation in 1944. Expressions of export were developed from these inventory data and from informed assumptions about the vectors which act to mobilize material containing Pu. This 14-acre pond provides a realistic illustration of the mobility of Pu in a lentic ecosystem. The ecological behavior of Pu in this pond is similar to that in other contaminated aquatic systems having widely differing limnological characteristics. Since its creation, this pond has received about one Ci of /sup 239/,/sup 240/Pu and /sup 238/Pu, most of which has been retained by its sediments. Submerged plants, mainly diatoms and Potamogeton, accumulate >95% of the Pu contained in biota. Emergent insects are the only direct biological route of export, mobilizing about 5 x 10/sup 3/ nCi of Pu annually, which is also the estimated maximum quantity of the Pu exported by waterfowl, birds and mammals collectively. There is no apparent significant export by wind, and it is not likely that Pu has migrated to the ground water below U-Pond via percolation. Although this pond has a rapid flushing rate, a eutrophic nutrient supply with a diverse biotic profile, and interacts with an active terrestrial environment, it appears to effectively bind Pu and prevent it from entering pathways to man and other life.

  10. Site preference of metal atoms in Gd_5_-_xM_xTt_4 (M = Zr, Hf; Tt = Si, Ge)

    International Nuclear Information System (INIS)

    Yao, Jinlei; Mozharivskyj, Yurij

    2011-01-01

    Zirconium and hafnium were incorporated into the Gd_5Ge_4 and Gd_5Si_4 parent compounds in order to study the metal-site occupation in the M_5X_4 magnetocaloric phases (M = metals; X = p elements) family. The Gd_5_-_xZr_xGe_4 phases adopt the orthorhombic Sm_5Ge_4-type (space group Pnma) structure for x ≤ 1.49 and the tetragonal Zr_5Si_4-type (P4_12_12) structure for x ≥ 1.77. The Gd_5_-_xHf_xSi_4 compounds crystallize in the orthorhombic Gd_5Si_4-type (Pnma) structure for x ≤ 0.41 and the Zr_5Si_4-type structure for x ≥ 0.7. In both systems, single-crystal X-ray diffraction reveals that the Zr/Hf atoms preferentially occupy the slab-surface M2 and slab-center M3 sites, both of which have a significantly larger Zr/Hf population than the slab-surface M1 site. The metal-site preference, i.e. the coloring problem on the three metal sites, is discussed considering geometric and electronic effects of the local coordination environments. The analysis of the metal-site occupation in Gd_5_-_xZr_xGe_4 and Gd_5_-_xHf_xSi_4 as well as other metal-substituted M_5X_4 systems suggests that both geometric and electronic effects can be used to explain the metal-site occupation. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  11. Fact reactor fuel alloys: Retrospective and prospective views

    International Nuclear Information System (INIS)

    Nevitt, M.V.

    1989-01-01

    The relationship between the physical metallurgy of the EBR-II metallic fuel, U-5% Fs, and its performance in the reactor are described. An understanding of these relationships, along with the optimal matching of fuel properties to fuel-element design, have been essential in the 23 year successful utilization of the fuel. The knowledge and experience gained are being employed in the current development of a new U-Pu-Zr metallic fuel for a proposed advanced reactor (orig./MM)

  12. Chlorine determination in (U, Pu)C fuel by total reflection X-ray fluorescence spectrometry

    International Nuclear Information System (INIS)

    Misra, Nand Lal; Dhara, Sangita; Mudher, Khush Dev Singh; Aggarwal, Suresh K.; Thakur, Uday Kumar; Shah, Dipti; Sawant, R.M.; Ramakumar, K.L.

    2007-01-01

    A Total Reflection X-ray Fluorescence (TXRF) method for the determination of chlorine in (U,Pu)C has been developed. The method involves calibration of the instrument with standard solutions and validation of TXRF determination of chlorine using synthetic standard solutions. Cl K α line excited with W L α source was used for TXRF determinations of chlorine. Chlorine present in trace amounts in (U,Pu)C samples was first separated by pyro hydrolysis. The evolved chlorine, in form of HCl, was collected in 5 mM NaOH solution. This solution was analyzed for chlorine by Total Reflection X-ray Fluorescence Spectrometry using cobalt as an internal standard. The TXRF detection limit of chlorine was found to be 3.6 pg with sample size of 30 μL. In order to assess the applicability of TXRF method for chlorine determinations in other nuclear materials, one U 3 O 8 trace element standard was also analyzed for chlorine in similar way. The precision of the method was found to be 25% (1 σ) at ng level in most of the cases. (author)

  13. Neutronic analysis concerning the utilization of mixed U N-Pu N nitride fuel for fast reactors

    International Nuclear Information System (INIS)

    Renke, C.A.C.; Batista, J.L.; Waintraub, M.; Santos Bastos, W. dos; Brito Aghina, L.O. de.

    1991-08-01

    Neutronic behavior of mixed UN-PuN nitride fuel in substitution of the mixed oxide U O 2 - Pu O 2 for fast reactors is discussed with focus on Super Phenix I. Characteristics parameters of both cores are calculated and compared and the results presented show a great advantage for the nitride fuel, pointing out a larger performance of fuel elements in the core and an effective reduction of reactivity loss during the cycle. (author)

  14. A combined XAFS, ESI TOF-MS and LIBD study on the formation of polynuclear Zr(IV), Th(IV) and Pu(IV) species

    Science.gov (United States)

    Rothe, J.; Walther, C.; Brendebach, B.; Büchner, S.; Fuss, M.; Denecke, M. A.; Geckeis, H.

    2009-11-01

    The long term radiotoxicity of spent nuclear fuel disposed of in deep underground repositories after discharge from nuclear power reactors is determined by actinide elements, mainly plutonium. Water intrusion into the repository might cause container corrosion and leaching of the waste matrices, leading to the release of Pu and other actinides into the geological environment. Performance assessment for a future nuclear waste repository requires detailed knowledge on actinide aqueous chemistry in the aquifer surrounding the disposal site. Tetravalent actinides exhibit a strong tendency towards hydrolysis and subsequent polymerization and/or colloid formation. These species provide a potential pathway for migration of actinides away from the repository. Therefore, it is of fundamental interest to study their generation and properties in-situ. To this end, X-ray Absorption Fine Structure Spectroscopy (XAFS) at the INE-Beamline for actinide research at ANKA, Electrospray Mass-Spectrometry (ESI TOF-MS) and Laser Induced Breakdown Detection (LIBD) are combined at FZK-INE in a comprehensive attempt to characterize Zr(IV) (An(IV) analogue), Th(IV) and Pu(IV) polymerization and colloid formation.

  15. Studies of the Fission Integrals of U-{sup 235} and Pu-{sup 239} with Cadmium and Filters

    Energy Technology Data Exchange (ETDEWEB)

    Hellstrand, E

    1965-04-15

    The resonance fissions in U{sup 235} and Pu{sup 239} have been studied using cadmium and boron filters. Fission chambers were used as detectors and the experiments were performed in beam geometry. The neutron energy distribution in the beams transmitted through the different filters was determined with a fast chopper. From the cadmium filter, measurements the fission resonance integrals were determined. The values obtained were 278{+-}9 b for U{sup 235} and 301{+-}10 b for Pu{sup 239}; 0.5 eV < E < 1 MeV. Complementary Pu{sup 239} measurements were made in which the fission events were detected from the fission product activity in irradiated foils. Contrary to what has been reported elsewhere the value of the Pu{sup 239} resonance integral, found in this way, agreed well with that obtained from the fission chamber measurement. The experiments with the boron filters yielded results which, for the thin filter, agreed well with those calculated from the cross section data given in the Karlsruhe compilation. The discrepancy was larger for the thick filter but the values did not disagree outside the common limits of error.

  16. Fabrication of preliminary fuel rods for SFR

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan

    2012-01-01

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  17. A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium; Contribution a l'etude du monocarbure d'uranium et de plutonium avec de faibles additions de zirconium

    Energy Technology Data Exchange (ETDEWEB)

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

    1970-03-01

    We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author) [French] On a etudie un combustible de type carbure (U,Pu)C pour les reacteurs a neutrons rapides. Les recherches preliminaires ont porte sur le carbure (UZr)C (rapport CEA-R-3765(1)). L'addition de faibles quantites de zirconium (3 at. pour cent) au monocarbure (U,Pu)C, ameliore certaines proprietes, commee la tenue a la corrosion atmospherique, la durete et surtout la compatibilite avec l'acier inoxydable X-18 M, Par contre le coefficient de dilatation et la densite sont peu changes. Le rapport Pu/Pu+U etait fixe a 20 pour cent. Deux procedes de fabrication ont ete etudies: l'un par fusion a l'arc, l'autre par frittage a partir de metaux hydrures. Au vu des resultats metallurgiques obtenus le carbure (U,Pu,Zr)C semble presenter un interet certain. (auteur)

  18. A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium; Contribution a l'etude du monocarbure d'uranium et de plutonium avec de faibles additions de zirconium

    Energy Technology Data Exchange (ETDEWEB)

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

    1970-03-01

    We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author) [French] On a etudie un combustible de type carbure (U,Pu)C pour les reacteurs a neutrons rapides. Les recherches preliminaires ont porte sur le carbure (UZr)C (rapport CEA-R-3765(1)). L'addition de faibles quantites de zirconium (3 at. pour cent) au monocarbure (U,Pu)C, ameliore certaines proprietes, commee la tenue a la corrosion atmospherique, la durete et surtout la compatibilite avec l'acier inoxydable X-18 M, Par contre le coefficient de dilatation et la densite sont peu changes. Le rapport Pu/Pu+U etait fixe a 20 pour cent. Deux procedes de fabrication ont ete etudies: l'un par fusion a l'arc, l'autre par frittage a partir de metaux hydrures. Au vu des resultats metallurgiques obtenus le carbure (U,Pu,Zr)C semble presenter un interet certain. (auteur)

  19. Effect of alloying on elastic properties of ZrN based transition metal nitride alloys

    KAUST Repository

    Kanoun, Mohammed; Goumri-Said, Souraya

    2014-01-01

    We report the effect of composition and metal sublattice substitutional element on the structural, elastic and electronic properties of ternary transition metal nitrides Zr1-xMxN with M=Al, Ti, Hf, V, Nb, W and Mo. The analysis of the elastic constants, bulk modulus, shear modulus, Young's modulus, and Poisson's ratio provides insights regarding the mechanical behavior of Zr1-xMxN. We predict that ternary alloys are more ductile compared to their parent binary compounds. The revealed trend in the mechanical behavior might help for experimentalists on the ability of tuning the mechanical properties during the alloying process by varying the concentration of the transition metal. © 2014 Elsevier B.V.

  20. Effect of alloying on elastic properties of ZrN based transition metal nitride alloys

    KAUST Repository

    Kanoun, Mohammed

    2014-09-01

    We report the effect of composition and metal sublattice substitutional element on the structural, elastic and electronic properties of ternary transition metal nitrides Zr1-xMxN with M=Al, Ti, Hf, V, Nb, W and Mo. The analysis of the elastic constants, bulk modulus, shear modulus, Young\\'s modulus, and Poisson\\'s ratio provides insights regarding the mechanical behavior of Zr1-xMxN. We predict that ternary alloys are more ductile compared to their parent binary compounds. The revealed trend in the mechanical behavior might help for experimentalists on the ability of tuning the mechanical properties during the alloying process by varying the concentration of the transition metal. © 2014 Elsevier B.V.

  1. Computer simulations of nanoindentation in Mg-Cu and Cu-Zr metallic glasses

    DEFF Research Database (Denmark)

    Paduraru, Anca; Andersen, Ulrik Grønbjerg; Thyssen, Anders

    2010-01-01

    The formation of shear bands during plastic deformation of Cu0.50Zr0.50 and Mg0.85Cu0.15 metallic glasses is studied using atomic-scale computer simulations. The atomic interactions are described using realistic many-body potentials within the effective medium theory, and are compared with similar...... simulations using a Lennard-Jones description of the material. The metallic glasses are deformed both in simple shear and in a simulated nanoindentation experiment. Plastic shear localizes into shear bands with a width of approximately 5 nm in CuZr and 8 nm in MgCu. In simple shear, the shear band formation...... is very clear, whereas only incipient shear bands are seen in nanoindentation. The shear band formation during nanoindentation is sensitive to the indentation velocity, indenter radius and the cooling rate during the formation of the metallic glass. For comparison, a similar nanoindentation simulation...

  2. A First-Principles Study on the Vibrational and Electronic Properties of Zr-C MXenes

    Science.gov (United States)

    Wang, Chang-Ying; Guo, Yong-Liang; Zhao, Yuan-Yuan; Zeng, Guang-Li; Zhang, Wei; Ren, Cui-Lan; Han, Han; Huai, Ping

    2018-03-01

    Within the framework of density functional theory calculations, the structural, vibrational, and electronic properties of Zr n C n - 1 (n = 2, 3, and 4) and their functionalized MXenes have been investigated. We find that the most stable configurations for Zr-C MXene are the ones that the terminal groups F, O, and OH locate on the common hollow site of the superficial Zr layer and its adjacent C layer. F and OH-terminated Zr 3 C 2 and Zr 4 C 3 have small imaginary acoustic phonon branches around Γ point while the others have no negative phonon modes. The pristine MXenes (Zr 2 C, Zr 3 C 2 and Zr 4 C 3 ) are all metallic with large DOS contributed by the Zr atom at the Fermi energy. When functionalized by F, O and OH, new hybridization states appear and the DOS at the Fermi level are reduced. Moreover, we find that their metallic characteristic increases with an increase in n. For (Zr n C n - 1 )O 2, Zr 2 CO 2 is a semiconductor, Zr 3C2O2 is a semimetal, and Zr 4 C 3O2 becomes a metal. Supported by the National Natural Science Foundation of China under Grant Nos. 11605273, 21571185, U1404111, 11504089, 21501189, 21676291, the Shanghai Municipal Science and Technology Commission 16ZR1443100, the Strategic Priority Research Program of the Chinese Academy of Sciences (XDA02040104)

  3. Anomalous diffusion in body-centred and face-centred cubic metals

    International Nuclear Information System (INIS)

    Zanghi, J.-P.

    1975-10-01

    The initial rates of contraction due to self-irradiation damage at 4.2K in three PuSc alloys (5, 12, 18 at % Sc) stabilized in f.c.c. delta-phase were measured. The high negative value of the formation volume of a Frenkel pair which is deduced by extrapolating for pure Pu, can only be explained by assuming that the interstitial Pu may partly recover its distortion energy by creating bonds with its neighbours, by a localized enhancement of the d.f. hybridization and especially by provoking the formation of bonds between its very neighbours. It is shown that about twenty atoms around the interstitial Pu are affected by these bonds. The self-irradiation at 4.2K of a b.c.c. UPuMo alloy was also studied. The activation volume for self-diffusion of Pu in b.c.c. PuZr alloys (10 and 40 at % Zr) was determined. So the validity of Nachtrieb's melting-diffusion correlation could be checked. Indeed, in the Pu 40 at % Zr alloy, which has a pressure temperature diagram the liquidus of which has a positive slope, a positive activation volume was found, whereas in pure epsilon Pu which as a negative slope, the activation volume is negative. A self-diffusion mechanism in PuZr alloys is proposed. A study of the diffusion of Am in these alloys showed that Am and Pu likely diffuse by the same mechanism [fr

  4. Improved capability of U-ZrH fuels

    Energy Technology Data Exchange (ETDEWEB)

    Gietzen, A J [General Atomic Company, San Diego, CA (United States)

    1983-08-01

    This paper provides a brief background on TRIGA fuels and a summary of the development of U-ZrH fuels utilizing higher uranium loadings in the alloy. Most of the development work was reported two years ago; however, in this paper some emphasis will be placed upon applications that General Atomic Company is now making with these higher alloy fuels. They are referred to as higher uranium content alloys because their development was not really LEU development; the TRIGA fuel has used low enrichment since its inception.

  5. Episode 49 of the Pu'u 'Ō'ō-Kūpaianaha eruption of Kilauea volcano-breakdown of a steady-state eruptive era

    Science.gov (United States)

    Mangan, M.T.; Heliker, C.C.; Mattox, T.N.; Kauahikaua, J.P.; Helz, R.T.

    1995-01-01

    The Pu'u 'O'o-Kupaianaha eruption (1983-present) is the longest lived rift eruption of either Kilauea or neighboring Mauna Loa in recorded history. The initial fissure opening in January 1983 was followed by three years of episodic fire fountaining at the Pu'u 'O'o vent on Kilauea's east rift zone ∼19km from the summit (episodes 4–47). These spectacular events gave way in July 1986 to five and a half years of near-continuous, low-level effusion from the Kupaianaha vent, ∼ 3km to the cast (episode 48). A 49th episode began in November 1991 with the opening of a new fissure between Pu'u 'O'o and Kupaianaha. This three week long outburst heralded an era of more erratic eruptive behavior characterized by the shut down of Kupaianaha in February 1992 and subsequent intermittent eruption from vents on the west flank of Pu'u 'O'o (episodes 50 and 51). The events occurring over this period are due to progressive shrinkage of the rift-zone reservoir beneath the eruption site, and had limited impact on eruption temperatures and lava composition.

  6. Modelling the radial distribution of the inventory of U and Pu isotopes in PWR fuel rods at high burnup

    International Nuclear Information System (INIS)

    Furlano, L.; Marino, A.C.

    2015-01-01

    Nowadays trends on nuclear reactors are to take advantage of a more efficient use of the fuels. In this way, goals like decreasing the volume of radioactive wastes, increasing life time in the reactor, extended burnup and accident-tolerance fuels are taken as the principal guidelines for the design and construction for actual nuclear fuels. The development of tools and the capability of prediction of the behaviour of nuclear fuels under irradiation let us design and improve the new demanding conditions without neglecting the security and economics. In this way, the study of the evolution and details of the fissile material of the fuel in particular the 235 U inventory evolution and the 239 Pu production and evolution are demanding conditions. We implemented the PILR code, based on the RAPID model, for the radial distribution of U and Pu radioisotopes. The aim of this work is to determine production and disappearance of Pu and U isotopes during the irradiation of commercial and experimental fuel rods. (author)

  7. Neutron induced fission cross sections for 232Th, 235,238U, 237Np, and 239Pu

    International Nuclear Information System (INIS)

    Lisowski, P.W.; Ullmann, J.L.; Balestrini, S.J.; Hill, N.W.; Carlson, A.D.; Wasson, O.A.

    1989-01-01

    Neutron-induced fission cross section ratios for samples of 232 Th, 235,238 U, 237 Np and 239 Pu have been measured from 1 to 400 MeV. The fission reaction rate was determined for all samples simultaneously using a fast parallel plate ionization chamber at a 20-m flight path. A well characterized annular proton recoil telescope was used to measure the neutron fluence from 3 to 30 MeV. Those data provided the shape of the 235 U(n,f) cross section relative to the hydrogen scattering cross section. That shape was then normalized to the very accurately known value for 235 U(n,f) at 14.178 MeV. From 30 to 400 MeV cross section values were determined using the neutron fluence measured with a plastic scintillator. Cross section values of 232 Th, 235,238 U, 237 Np and 239 Pu were computed from the ratio data using the authors' values for 235 U(n,f). In addition to providing new results at high neutron energies, these data highlight several areas of deficiency in the evaluated nuclear data files and provide new information for the 235 U(n,f) standard

  8. The growth of the metallic ZrNx thin films on P-GaN substrate by pulsed laser deposition

    Science.gov (United States)

    Gu, Chengyan; Sui, Zhanpeng; Li, Yuxiong; Chu, Haoyu; Ding, Sunan; Zhao, Yanfei; Jiang, Chunping

    2018-03-01

    Although metal nitride thin films have attractive prospects in plasmonic applications due to its stable properties in harsh environments containing high temperatures, shock, and contaminants, the effect of deposition parameters on the properties of the metallic ZrN grown on III-N semiconductors by pulse laser deposition still lacks of detailed exploration. Here we have successfully prepared metallic ZrNx films on p-GaN substrate by pulsed laser deposition in N2 ambient of various pressures at a fixed substrate temperature (475 °C). It is found that the films exhibit quite smooth surfaces and (111) preferred orientation. The X-ray photoelectron spectroscopy measurements indicate that carbon contamination can be completely removed and oxygen contamination is significantly reduced on the film surfaces after cleaning using Ar+ sputtering. The N/Zr ratio increases from 0.64 to 0.75 when the N2 pressure increases from 0.5 Pa to 3 Pa. The optical reflectivity spectra measured by the UV-vis-NIR spectrophotometer show that the ZrNx is a typical and good metallic-like material and its metallic properties can be tuned with changing the film compositions.

  9. Exploratory study of fission product yields of neutron-induced fission of 235U , 238U , and 239Pu at 8.9 MeV

    Science.gov (United States)

    Bhatia, C.; Fallin, B. F.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.; Arnold, C. W.; Bond, E.; Bredeweg, T. A.; Fowler, M. M.; Moody, W.; Rundberg, R. S.; Rusev, G. Y.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Macri, R.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.

    2015-06-01

    Using dual-fission chambers each loaded with a thick (200 -400 -mg /c m2) actinide target of 235 ,238U or 239Pu and two thin (˜10 -100 -μ g /c m2) reference foils of the same actinide, the cumulative yields of fission products ranging from 92Sr to 147Nd have been measured at En= 8.9 MeV . The 2H(d ,n ) 3He reaction provided the quasimonoenergetic neutron beam. The experimental setup and methods used to determine the fission product yield (FPY) are described, and results for typically eight high-yield fission products are presented. Our FPYs for 235U(n ,f ) , 238U(n ,f ) , and 239Pu(n ,f ) at 8.9 MeV are compared with the existing data below 8 MeV from Glendenin et al. [Phys. Rev. C 24, 2600 (1981), 10.1103/PhysRevC.24.2600], Nagy et al. [Phys. Rev. C 17, 163 (1978), 10.1103/PhysRevC.17.163], Gindler et al. [Phys. Rev. C 27, 2058 (1983), 10.1103/PhysRevC.27.2058], and those of Mac Innes et al. [Nucl. Data Sheets 112, 3135 (2011), 10.1016/j.nds.2011.11.009] and Laurec et al. [Nucl. Data Sheets 111, 2965 (2010), 10.1016/j.nds.2010.11.004] at 14.5 and 14.7 MeV, respectively. This comparison indicates a negative slope for the energy dependence of most fission product yields obtained from 235U and 239Pu , whereas for 238U the slope issue remains unsettled.

  10. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel

  11. Assembly-level analysis of heterogeneous Th–Pu PWR fuel

    International Nuclear Information System (INIS)

    Zainuddin, Nurjuanis Zara; Parks, Geoffrey T.; Shwageraus, Eugene

    2017-01-01

    Highlights: • We directly compare homogeneous and heterogeneous Th–Pu fuel. • Examine whether there is an increase in Pu incineration in the latter. • Homogeneous fuel was able to achieve much higher Pu incineration. • In the heterogeneous case, U-233 breeding is faster (larger power fraction), thus decreasing incineration of Pu. - Abstract: This study compares homogeneous and heterogeneous thorium–plutonium (Th–Pu) fuel assemblies (with high Pu content – 20 wt%), and examines whether there is an increase in Pu incineration in the latter. A seed-blanket configuration based on the Radkowsky thorium reactor concept is used for the heterogeneous assembly. This separates the thorium blanket from the uranium seed, or in this case a plutonium seed. The seed supplies neutrons to the subcritical thorium blanket which encourages the in situ breeding and burning of "2"3"3U, allowing the fuel to stay critical for longer, extending burnup of the fuel. While past work on Th–Pu seed-blanket units shows superior Pu incineration compared to conventional U–Pu mixed oxide fuel, there is no literature to date that directly compares the performance of homogeneous and heterogeneous Th–Pu assembly configurations. Use of exactly the same fuel loading for both configurations allows the effects of spatial separation to be fully understood. It was found that the homogeneous fuel with and without burnable poisons was able to achieve much higher Pu incinerations than the heterogeneous fuel configurations, while still attaining a reasonably high discharge burnup. This is because in the heterogeneous cases, "2"3"3U breeding is faster, thereby contributing to a much larger fraction of total power produced by the assembly. In contrast, "2"3"3U build-up is slower in the homogeneous case and therefore Pu burning is greater. This "2"3"3U begins to contribute a significant fraction of power produced only towards the end of life, thus extending criticality, allowing more Pu to

  12. Welding of CuZr-based metallic glasses on air

    International Nuclear Information System (INIS)

    Batalha, W.; Gargarella, P.; Kiminami, C.S.

    2016-01-01

    Metallic glass alloys have been studied aiming at its exceptional mechanical properties. This alloys processing's requires high cooling rates, which diminishes the sample's size. There by welding these samples without the loss of amorphous structure is a good alternative. The DEMa group has developed a technique based on Joule effect heating. By applying pressure and electric current, reaching temperatures of super cold liquids (the temperature between crystallizing and vitric transition), the vitric metal has it’s viscosity reduced and sample binding occur. The objective of this paper was to weld samples of cylindrical geometry of 2 and 3 mm in diameter and 4 mm in length of the compositions Cu46Zr42Al7Y5 and (Cu47Zr45Al8)98Y2. The process was done using 2 copper electrodes under(over) argon flux. The samples were later analysed by microscopy, differential scanning calorimetry and X ray diffraction. The results showed that this kind of welding process is possible since crystal formation on the welding region did not occur and there were no faults like cracks or porosity. (author)

  13. α spectrometry on thick sources. Application to Pu dosing in U-Pu alloys with a low Pu content

    International Nuclear Information System (INIS)

    Guery, Marcel

    1969-06-01

    After having noticed that alpha spectrometry of thick sources has been not much studied, the author reports the application of this type of alpha spectrometry to a problem of plutonium dosing in uranium-plutonium alloys with low plutonium content. Four measurement and test campaigns, each being few month long, have been performed. During the first three ones, the best measurement conditions have been determined, the necessary apparatus has been acquired, and data compilation has been elaborated. This report addresses the last campaign during which exploitable results have been obtained. After some generalities about alpha rays from thick sources, the author reports a detailed examination of measurements made on U-Pu alloys, and of corrections to be made, notably those due to the Am-241 content, to stacking and to oxidation. The method appears to be very sensitive, with a precision better than 1 per cent. Improvements seem possible [fr

  14. Oxidation behaviour of Zr-based bulk metallic glasses

    International Nuclear Information System (INIS)

    Wang, Bin

    2011-01-01

    The Zr-based bulk metallic glasses, developed since the late 1980's, have very interesting mechanical properties, which can be considered for many applications including working under oxidizing atmosphere conditions at high temperatures. It is therefore interesting to study their oxidation resistance and to characterize the oxide scale formed on alloys surface. The fundamental objective of this thesis is to enhance the understanding of the role of various thermodynamic and chemistry parameters on the oxidation behaviour of the Zr-based bulk metallic glasses at high temperature under dry air, to determine the residual stresses in the oxide layer, in comparison with their crystalline alloys with the same chemical composition after an annealing treatment. The oxidation kinetics of these glasses and the crystalline structure of oxide scale ZrO 2 depend on the temperature and the oxidation duration: for short periods of oxidation or at a temperature below Tg, the kinetics follows a parabolic law, whereas, if the sample is oxidized at T ≥ Tg, the kinetics can be divided into two parts. The crystalline counterparts are oxidized by a parabolic rule whatever the temperature; for long oxidation duration at a temperature close to Tg, the kinetics becomes more complex because of the crystallisation of the glasses during the oxidation tests. Also the crystalline structure of the oxide layers depends on the oxidation temperature: the oxide layer consists only in tetragonal Zirconia at T ≤ Tg, while monoclinic Zirconia was formed at higher temperature. The mechanism of the formation of the oxide scale is due to both the interior diffusion of Oxygen ions and the external diffusion of Zirconium ions. However the diffusion of Zirconium ions slows gradually during the crystallisation process of the glass matrix. When the crystallisation is completed, the formation of Zirconia is controlled by only the internal diffusion of oxygen ions. The corresponding residual stresses

  15. Influence of in situ formed ZrB2 particles on microstructure and mechanical properties of AA6061 metal matrix composites

    International Nuclear Information System (INIS)

    Dinaharan, I.; Murugan, N.; Parameswaran, Siva

    2011-01-01

    Highlights: → In situ fabrication of aluminium metal matrix composite reinforced ZrB 2 particles. → Colour metallography of composites. → Improvement of matrix properties by ZrB 2 particles. → Sliding wear behaviour of in situ composites. - Abstract: Particulate reinforced metal matrix composites (PMMCs) have gained considerable amount of research emphasis and attention in the present era. Research is being carried out across the globe to produce new combination of PMMCs. PMMCs are prepared by adding a variety of ceramic particles with monolithic alloys using several techniques. An attempt has been made to produce aluminium metal matrix composites reinforced with zirconium boride (ZrB 2 ) particles by the in situ reaction of K 2 ZrF 6 and KBF 4 salts with molten aluminium. The influence of in situ formed ZrB 2 particles on the microstructure and mechanical properties of AA6061 alloy was studied in this work. The in situ formed ZrB 2 particles significantly refined the microstructure and enhanced the mechanical properties of AA6061 alloy. The weight percentage of ZrB 2 was varied from 0 to 10 in steps of 2.5. Improvement of hardness, ultimate tensile strength and wear resistance of AA6061 alloy was observed with the increase in ZrB 2 content.

  16. Microwave based oxidation process for recycling the off-specification (U,Pu)O{sub 2} fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Singh, G., E-mail: gitendars@barctara.gov.in [Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, 401 502 (India); Khot, P.M. [Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, 401 502 (India); Kumar, Pradeep [Integrated Fuel Fabrication Facility (IFFF), Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Bhatt, R.B.; Behere, P.G.; Afzal, Mohd [Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, 401 502 (India)

    2017-02-15

    This paper reports development of a process named MicroWave Direct Oxidation (MWDO) for recycling the off-specification (U,Pu)O{sub 2} mixed oxide (MOX) fuel pellets generated during fabrication of typical fast reactor fuels. MWDO is a two-stage, single-cycle process based on oxidative pulverisation of pellets using 2450 MHz microwave. The powder sinterability was evaluated by bulk density and BET specific surface area. The oxidised powders were analyzed for phases using XRD and stoichiometry by thermogravimetry. The sinterability was significantly enhanced by carrying out oxidation in higher oxygen partial pressure and by subjecting MOX to multiple micronisation-oxidation cycles. After three cycles, the recycled powder from (U,28%Pu)O{sub 2} resulted surface area >3 m{sup 2}/g and 100% re-used for MOX fabrication. The flow sheet was developed for maximum utilization of recycled powder describable by a parameter called Scrap Recycling Ratio (SRR). The process demonstrates smaller processing cycle, better powder properties and higher oxidative pulverisation over conventional method. - Highlights: • A process for recycling the off-specification (U,Pu)O{sub 2} sintered fuel pellets of fast reactors was demonstrated. • The method is a two-stage, single cycle process based on oxidative pulverization of MOX pellets using 2450 MHz microwave. • The process demonstrated utilization of recycled powder with SRR of 1.

  17. Molecular dynamics simulations of the structure evolutions of Cu-Zr metallic glasses under irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lang, Lin [College of Materials Science and Engineering, Hunan University, Changsha 410082 (China); Department of Applied Physics, School of Physics and Electronics, Hunan University, Changsha 410082 (China); Tian, Zean; Xiao, Shifang [Department of Applied Physics, School of Physics and Electronics, Hunan University, Changsha 410082 (China); Deng, Huiqiu, E-mail: hqdeng@hnu.edu.cn [Department of Applied Physics, School of Physics and Electronics, Hunan University, Changsha 410082 (China); Ao, Bingyun [Science and Technology on Surface Physics and Chemistry Laboratory, Mianyang 621907 (China); Chen, Piheng, E-mail: chenpiheng@caep.cn [Science and Technology on Surface Physics and Chemistry Laboratory, Mianyang 621907 (China); Hu, Wangyu [College of Materials Science and Engineering, Hunan University, Changsha 410082 (China)

    2017-02-15

    Highlights: • The structural evolution of Cu{sub 64.5}Zr{sub 35.5} MG under irradiation was studied. • The structure clusters were analyzed using the LSCA method. • Most of these radiation damages have been self-recovered quickly. - Abstract: Molecular dynamics simulations have been performed to investigate the structural evolution of Cu{sub 64.5}Zr{sub 35.5} metallic glasses under irradiation. The largest standard cluster analysis (LSCA) method was used to quantify the microstructure within the collision cascade regions. It is found that the majority of clusters within the collision cascade regions are full and defective icosahedrons. Not only the smaller structures (common neighbor subcluster) but also primary clusters greatly changed during the collision cascades; while most of these radiation damages self-recover quickly in the following quench states. These findings indicate the Cu-Zr metallic glasses have excellent irradiation-resistance properties.

  18. Fuelling with flow at Bruce A

    Energy Technology Data Exchange (ETDEWEB)

    Gray, M G [Canadian General Electric Co. Ltd., Peterborough, ON (Canada)

    1997-12-31

    Fuelling with flow is the solution chosen by Bruce A to overcome the potential power pulse caused by a major inlet header failure. Fuelling with flow solves the problem by rearranging the core to place new fuel at the channel inlet and irradiated fuel at the channel outlet. The change has a significant impact on the Bruce A fuel handling system which was designed primarily to do on power fuelling in the against flow direction. Mechanical changes to the fuelling machine include a modification to the existing ram head and the replacement of standard fuel carriers with new fuelling with flow fuel carriers having the capability of opening the channel latch. Changes to the control system are more involved. A new set of operational sequences are required for both the upstream and downstream fuelling machines to achieve the fuel change. Steps based on sensitive ram push are added to reduce the risk of failing to close the latch at the correct position to properly support the fuel string. Changes are also required to the protective interlocks to allow fuelling with flow and reduce risk. A new fuel string supporting shield plug was designed and tested to reduce the risk of endplate cracking that could occur on the irradiated bundle that would have been supported directly by the channel latch. Some operational changes have been incorporated to accommodate this new shield plug. Considerable testing has been carried out on all aspects of fuel handling where fuelling with flow differs from the reference fuelling against flow. (author). 3 figs.

  19. Effects of Nb addition on icosahedral quasicrystalline phase formation and glass-forming ability of Zr--Ni--Cu--Al metallic glasses

    International Nuclear Information System (INIS)

    Fan, Cang; Li, Chunfei; Inoue, Akihisa; Haas, Volker

    2001-01-01

    This work shows that the crystallization process of Zr--Ni--Cu--Al metallic glass is greatly influenced by adding Nb as an alloying element. Based on the results of the differential scanning calorimetry experiments for metallic glasses Zr 69-x Nb x Ni 10 Cu 12 Al 9 (x=0--15at.%), the crystallization process takes place through two individual stages. For Zr 69 Ni 10 Cu 12 Al 9 (x=0), metastable hexagonal ω-Zr and a small fraction of tetragonal Zr 2 Cu are precipitated upon completion of the first exothermic reaction. Contrary to this alloy, the precipitation of a nanoquasicrystalline phase is detected when 5--10 at.% Nb is added. Furthermore, the crystallization temperature T x , supercooled liquid region ΔT x and reduced temperature T g /T L (T g is the glass transition temperature, T L the liquidus temperature) increase with increasing Nb content. These results indicate that adding Nb content to Zr--Ni--Cu--Al metallic glasses not only induces quasicrystalline phase formation, but also enhances glass-forming ability. Copyright 2001 American Institute of Physics

  20. Redox reactions of U(IV) and Pu(IV) with H2O2 generated in nitric acid media by power ultrasound

    International Nuclear Information System (INIS)

    Moisy, P.; Venault, L.; Madic, C.; Nikitenko, S.

    1998-01-01

    Power ultrasound causes water molecule dissociation on H o and OH o radicals due to high local temperatures and pressures generated in the cavitation threshold. In nitric acid media scavenging of OH o radicals with NO 3 - followed by NO 3 o radicals hydrolysis leads to H 2 O 2 formation. It was shown that H 2 O 2 generated under the effect of ultrasound with the frequency 20 kHz and intensity 1-3 Wcm -2 (Ar atmosphere) oxidizes U(IV) to U(VI) or reduces Pu(IV) to Pu(III) in 1-4 M HNO 3 in the presence of antinitrous reagents ( N 2 H 5 NO 3 or NH 2 SO 3 H). The effect of HNO 3 concentration and ultrasonic intensity on the kinetics of U(IV) oxidation and Pu(IV) reduction was studied. (author)

  1. Thermodynamics and phase diagrams of the plutonium-uranium, uranium-zirconium, plutonium-zirconium and plutonium-uranium-zirconium systems

    International Nuclear Information System (INIS)

    Agarwal, R.; Venugopal, V.

    2004-05-01

    Thermodynamic and phase diagram data reported in literature for the binaries, Pu-U, Pu-Zr and U-Zr , were compiled and optimised to calculate Gibbs energies of all the binary phases of these systems. Lukas program was used to carry out these optimisations, where, thermodynamic and phase diagram data of all the binary phases of a binary system were optimised simultaneously. Gibbs energy sets thus calculated were used to compare our results with the experimental and calculated phase diagram and thermodynamic data reponed in the literature. The Gibbs energies of the binary systems were then compiled together to define Pu-U-Zr ternary system. (author)

  2. The factors affecting MTC of thorium–plutonium-fuelled PWRs

    International Nuclear Information System (INIS)

    Zainuddin, Nurjuanis Zara; Parks, Geoffrey T.; Shwageraus, Eugene

    2016-01-01

    Plutonium loading in a plutonium–thorium (Pu–Th) mixed oxide (MOX) fuelled pressurized water reactor (PWR) core is typically constrained by large maximum radial form factors (RFF) and positive moderator temperature coefficient (MTC). The large form factors in higher Pu content fuels stem from the large differences in burnup, and thus reactivity, between fresh and burnt fuel, while positive MTC can potentially be the result of the high soluble boron concentrations needed to maintain criticality for such reactive fuel. The conventional solution to these problems is the use of burnable poisons (BPs). While BPs are able to reduce RFF, the positive MTC is not entirely due to a large critical boron concentration (CBC) requirement. In fact, analysis shows a positive MTC in Th–Pu fuel is mainly caused by fissioning in the epithermal–fast energy range. A reduction in epithermal–fast fissioning through the use of certain BPs and the strategic employment of loading patterns that encourage leakage are more effective in attaining negative MTC, as a reduction in CBC has a negligible effect on MTC. This paper examines the contributions to positive MTC by isotope and energy and identifies characteristics of BPs that are able to mitigate positive MTC in a Pu–Th MOX PWR core.

  3. Mixed U/Pu oxide fabrication facility for gel-sphere-pac fuel

    International Nuclear Information System (INIS)

    1978-09-01

    This paper describes a conceptual plant which uses the gel-sphere-pac process to fabricate mixed oxide (MOX) fuel and covers (1) fabrication of co-processed MOX fuel and (2) fabrication of co-processed spiked MOX fuel, using 60 Co. The report describes: the fuel fabrication process and plant layout, including scrap and waste processing; and maintenance safety and ventilation measures. A description of the conversion of U and Pu nitrate using a gel sphere process is given in Appendix A

  4. Time-resolved record of 236U and 239,240Pu isotopes from a coral growing during the nuclear testing program at Enewetak Atoll (Marshall Islands).

    Science.gov (United States)

    Froehlich, M B; Chan, W Y; Tims, S G; Fallon, S J; Fifield, L K

    2016-12-01

    A comprehensive series of nuclear tests were carried out by the United States at Enewetak Atoll in the Marshall Islands, especially between 1952 and 1958. A Porites Lutea coral that was growing in the Enewetak lagoon within a few km of all of the high-yield tests contains a continuous record of isotopes, which are of interest (e.g. 14 C, 236 U, 239,240 Pu) through the testing period. Prior to the present work, 14 C measurements at ∼2-month resolution had shown pronounced peaks in the Δ 14 C data that coincided with the times at which tests were conducted. Here we report measurements of 236 U and 239,240 Pu on the same coral using accelerator mass spectrometry, and again find prominent peaks in the concentrations of these isotopes that closely follow those in 14 C. Consistent with the 14 C data, the magnitudes of these peaks do not, however, correlate well with the explosive yields of the corresponding tests, indicating that smaller tests probably contributed disproportionately to the debris that fell in the lagoon. Additional information about the different tests can also be obtained from the 236 U/ 239 Pu and 240 Pu/ 239 Pu ratios, which are found to vary dramatically over the testing period. In particular, the first thermonuclear test, Ivy-Mike, has characteristic 236 U/ 239 Pu and 240 Pu/ 239 Pu signatures which are diagnostic of the first arrival of nuclear test material in various archives. Copyright © 2016 Elsevier Ltd. All rights reserved.

  5. Thermodynamic analysis for molten stratification test MASCA with ionic liquid U-Zr-Fe-O-B-C-FPs database

    International Nuclear Information System (INIS)

    Fukasawa, Masanori; Tamura, Shigeyuki

    2007-01-01

    The molten corium stratification tested in the OECD MASCA project was analyzed with our thermo-dynamic database and the database was verified to be effective for the stratification analysis. The MASCA test shows that the molten corium can be stratified with the metal layer under the oxide when sub-oxidized corium including iron was retained in the lower head of the reactor vessel. This stratification is caused by the increased density of the metal layer attributed to a transfer of uranium metal that was reduced from uranium oxide by zirconium. Thermodynamic equilibrium calculations with the database, which was developed for the corium U-Zr-Fe-O-B-C-FPs system using the ionic two-sublattice model for liquid, show quantitative agreements with the MASCA test, such as the composition of each layer, fission product (FP) partitioning between the layers and B 4 C effect on the stratification. (author)

  6. Fission Product Yield Study of 235U, 238U and 239Pu Using Dual-Fission Ionization Chambers

    Science.gov (United States)

    Bhatia, C.; Fallin, B.; Howell, C.; Tornow, W.; Gooden, M.; Kelley, J.; Arnold, C.; Bond, E.; Bredeweg, T.; Fowler, M.; Moody, W.; Rundberg, R.; Rusev, G.; Vieira, D.; Wilhelmy, J.; Becker, J.; Macri, R.; Ryan, C.; Sheets, S.; Stoyer, M.; Tonchev, A.

    2014-05-01

    To resolve long-standing differences between LANL and LLNL regarding the correct fission basis for analysis of nuclear test data [M.B. Chadwick et al., Nucl. Data Sheets 111, 2891 (2010); H. Selby et al., Nucl. Data Sheets 111, 2891 (2010)], a collaboration between TUNL/LANL/LLNL has been established to perform high-precision measurements of neutron induced fission product yields. The main goal is to make a definitive statement about the energy dependence of the fission yields to an accuracy better than 2-3% between 1 and 15 MeV, where experimental data are very scarce. At TUNL, we have completed the design, fabrication and testing of three dual-fission chambers dedicated to 235U, 238U, and 239Pu. The dual-fission chambers were used to make measurements of the fission product activity relative to the total fission rate, as well as for high-precision absolute fission yield measurements. The activation method was employed, utilizing the mono-energetic neutron beams available at TUNL. Neutrons of 4.6, 9.0, and 14.5 MeV were produced via the 2H(d,n)3He reaction, and for neutrons at 14.8 MeV, the 3H(d,n)4He reaction was used. After activation, the induced γ-ray activity of the fission products was measured for two months using high-resolution HPGe detectors in a low-background environment. Results for the yield of seven fission fragments of 235U, 238U, and 239Pu and a comparison to available data at other energies are reported. For the first time results are available for neutron energies between 2 and 14 MeV.

  7. Dimensional, microstructural and compositional stability of metal fuels

    International Nuclear Information System (INIS)

    Solomon, A.A.; Dayananda, M.A.

    1993-01-01

    The projects undertaken were to address two areas of concern for metal-fueled fast reactors: metallurgical compatibility of fuel and its fission products with the stainless steel cladding, and effects of porosity development in the fuel on fuel/cladding interactions and on sodium penetration in fuel. The following studies are reported on extensively in appendices: hot isostatic pressing of U-10Zr by coupled boundary diffusion/power law creep cavitation, liquid Na intrusion into porous U-10Zr fuel alloy by differential capillarity, interdiffusion between U-Zr fuel and selected Fe-Ni-Cr alloys, interdiffusion between U-Zr fuel vs selected cladding steels, and interdiffusion of Ce in Fe-base alloys with Ni or Cr

  8. Inflow of atomic gas fuelling star formation

    DEFF Research Database (Denmark)

    Michałowski, M. J.; Gentile, G.; Hjorth, Jeppe

    2016-01-01

    Gamma-ray burst host galaxies are deficient in molecular gas, and show anomalous metal-poor regions close to GRB positions. Using recent Australia Telescope Compact Array (ATCA) Hi observations we show that they have substantial atomic gas reservoirs. This suggests that star formation in these ga......Gamma-ray burst host galaxies are deficient in molecular gas, and show anomalous metal-poor regions close to GRB positions. Using recent Australia Telescope Compact Array (ATCA) Hi observations we show that they have substantial atomic gas reservoirs. This suggests that star formation...... in these galaxies may be fuelled by recent inflow of metal-poor atomic gas. While this process is debated, it can happen in low-metallicity gas near the onset of star formation because gas cooling (necessary for star formation) is faster than the Hi-to-H2 conversion....

  9. Swelling behavior detection of irradiated U-10Zr alloy fuel using indirect neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Yong; Huo, He-yong; Wu, Yang [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Key Laboratory of Neutron Physics, Chinese Academy of Engineering Physics, Mianyang (China); Li, Jiangbo [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Zhou, Wei; Guo, Hai-bing [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Key Laboratory of Neutron Physics, Chinese Academy of Engineering Physics, Mianyang (China); Li, Hang, E-mail: lihang32@gmail.com [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Key Laboratory of Neutron Physics, Chinese Academy of Engineering Physics, Mianyang (China); Cao, Chao; Yin, Wei; Wang, Sheng; Liu, Bin; Feng, Qi-jie; Tang, Bin [Institute of Nuclear Physics and Chemistry, Chinese Academy of Engineering Physics, Mianyang (China); Key Laboratory of Neutron Physics, Chinese Academy of Engineering Physics, Mianyang (China)

    2016-11-21

    It is hopeful that fusion-fission hybrid energy system will become an effective approach to achieve long-term sustainable development of fission energy. U-10Zr alloy (which means the mass ratio of Zr is 10%) fuel is the key material of subcritical blanket for fusion-fission hybrid energy system which the irradiation performance need to be considered. Indirect neutron radiography is used to detect the irradiated U-10Zr alloy because of the high residual dose in this paper. Different burnup samples (0.1%, 0.3%, 0.5% and 0.7%) have been tested with a special indirect neutron radiography device at CMRR (China Mianyang Research Reactor). The resolution of the device is better than 50 µm and the quantitative analysis of swelling behaviors was carried out. The results show that the swelling behaviors relate well to burnup character which can be detected accurately by indirect neutron radiography.

  10. Pb, U, Ti, Hf and Zr distributions in zircons determined by proton microprobe and fission track techniques

    International Nuclear Information System (INIS)

    Clark, G.J.; Gulson, B.L.; Cookson, J.A.

    1979-01-01

    A proton microprobe has been used to determine Pb, Tl, Hf and Zr distributions across four single zircon crystals separated from a 'rapakivi' granite. The Pb and Zr data are quantitative: Pb and Tl concentrations were below the measurable limits for determinations in situ by most other techniques. The distribution of U in the same crystals was determined by the fission track technique. Limits on precision of U allow only a qualitative correlation of U and Pb, whereas the Tl and Pb correlation is more exactly determined. Zircons with distinct cores and overgrowths exhibited uniform Zr and Hf concentrations across the crystals, whereas the high U rims and 'inclusions' (domains) also had high Tl and Pb contents. Since almost all the Pb in these zircons is derived by radioactive decay of U, the Tl substitution has paralleled that of U. The results indicate that the high U domains are 'hot spots' rather than a separate mineral phase. The strong positive correlation of U and Pb indicates that there is little U daughter product migration relative to U, within the crystal. However, for the zircon population investigated here, the data are equivocal on the question of whether U addition to zircon crystals is associated with new zircon growth or not. In either case, the heterogeneous U and Pb distributions complicate any interpretations of U-Pb isotopic analysis for such zircon populations. (author)

  11. The integral fast reactor (IFR) concept: Physics of operation and safety

    International Nuclear Information System (INIS)

    Wade, D.C.; Chang, Y.I.

    1987-01-01

    The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed

  12. The integral fast reactor (IFR) concept: physics of operation and safety

    International Nuclear Information System (INIS)

    Wade, D.C.; Chang, Y.I.

    1987-01-01

    The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed

  13. Biodegradation of PuEDTA and Impacts on Pu Mobility

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  14. Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.; Mariani, R.D.

    1999-01-01

    Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr alloy fuel elements irradiated in the experimental breeder reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining. (orig.)

  15. Development of metallic fuel fabrication - A study on the interdiffusion behavior between ternary metallic fuel and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Soo; Seol, Kyung Won; Shon, In Jin [Chonbuk National University, Chonju (Korea)

    1999-04-01

    To study a new ternary metallic fuel for liquid metal reactor, various U-Zr-X alloys have been made by induction melting. The specimens were prepared for thermal stability tests at 630 deg. C upto 5000 hours in order to estimate the decomposition of the lamellar structure. Interdiffusion studies were carried out at 700 deg. C for 200 hours for the diffusion couples assembled with U-Zr-X ternary fuel versus austenitic stainless steel D9 and martensitic stainless steel HT9, respectively, to investigate the fuel-cladding compatibility. The ternary alloy, especially U-Zr-Mo and U-Zr-Nb alloys showed relatively good thermal stability as long as 5000hrs at 630 deg. C. From the composition profiles of the interdiffusion study, Fe penetrated deeper to the fuel side than other cladding elements such as Ni and Cr, whereas U did to the cladding side of fuel elements in the fuel/D9 couples. On the contrary, the reaction layers of Fuel/HT9 couple were thinner than that of Fuel/D9 couples and were less affected by cladding element, which was believed to be due to Zr rich layer between the fuel-cladding interface. HT9 is considered to be superior to D9 and a favorable choice as a cladding material in terms of fuel-cladding compatibility. 21 refs., 24 figs., 7 tabs. (Author)

  16. Structures and properties of (U,Pu)O2 containing non-active fission products. A simulation of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Schmitz, F.

    1969-01-01

    We have made oxides with the same uranium and plutonium content, the same stoichiometry and the same fission product content as an oxide fuel (U 0,8 PuO 2 )O 1,96 after 2 per cent burn up. We have calculated the stoichiometry changes due to irradiation and checked the calculation by X rays parameters measurements. We have calculated and measured the contraction of the oxide lattice due to fission products in solid solution. Microprobe analysis of precipitates have been made and have lead to the identification of non metallic barium containing compounds and have shown the particular behaviour of molybdenum. Some physical properties have been measured especially the electrical resistivity, the thermal diffusivity and the vapour pressure of zirconium in solid solution. (author) [fr

  17. Sorption properties of bentonite clays towards Pu(IV), U(VI), Np(V) and Cs: experimental and surface complexation study

    Energy Technology Data Exchange (ETDEWEB)

    Sabodina, M.N. [Institute of Physical Chemistry of Russian Academy of Science, Moscow 119192 (Russian Federation); Kalmykov, St.N.; Sapozhnikov, Yu.A. [Radiochemistry div., Chemistry dept., Lomonosov Moscow State University, Moscow 119992, (Russian Federation); Gupalo, T.A.; Beigul, V.P. [VNIPI Promtechnology, Moscow (Russian Federation)

    2005-07-01

    Full text of publication follows: Sorption of radionuclides, their diffusion in bentonite as well as its solubility are the major factors that define bentonite as a geochemical barrier. Sorption of cations by bentonite could be governed by two mechanisms including ion exchange with interlayer cations and formation of surface complexes with either silanol or aluminol groups. The aim of this work was to study mechanisms of {sup 137}Cs, Pu(IV), Np(V) and U(VI) sorption by bentonite and their solubility in bentonite pore waters. Bentonite (Khakassiya deposit) used in the experiments was taken in Na-form and characterized by powder X-ray diffraction, scanning electron microscopy, potentiometric titration. The cation exchange capacities of bentonite at pH=6 were measured by isotopic exchange with {sup 22}Na{sup +} and Cs{sup +} saturation. Sorption experiments were performed in N{sub 2} atmosphere in plastic vials. Bentonite samples were left in the working solutions to swell for few days before sorption experiments were performed. After the desired concentration of radionuclide ({sup 137}Cs, {sup 238}Pu, {sup 239}Pu, {sup 237}Np, {sup 239}Np, {sup 238}U) was added to the suspension, the required pH values are established and samples were left until the equilibrium was reached. Separation of solution after the sorption was performed using micro- and ultrafiltration techniques. The sorption of Pu(IV), U(VI) and Np(V) was highly pH dependent that indicates predominant surface complexation mechanism of sorption. For {sup 137}Cs the pH dependence of sorption was less pronounced and significant decrease of sorption occurs at pH<1.7 that indicate the ion exchange as the major mechanism. The equilibrium constant of Na{sup +}/Cs{sup +} exchange was calculated form sorption isotherms and pH dependence of sorption. It is established using micro- and ultra-filtrations, that sorption of radionuclides onto bentonite nano colloids is essential. Surface complexation modeling exercises

  18. Proposal for Analysis of the Safeguarded Nuclear Materials 235U and 239Pu by Delayed Neutrons Technique

    International Nuclear Information System (INIS)

    El-Mongy, S.A.

    2000-01-01

    This paper introduces, describes and initiates a very sensitive and rapid non-destructive technique to be used for analysis of the safeguarded nuclear materials 235 U and 239 Pu. The technique is based on fission of the nuclear material by neutrons and then measuring the delayed neutrons produced from the neutron rich fission products. By this technique, fissile isotope content ( 235 U) can be determined in the presence of the other fissile (e.g. 239 Pu) or fertile isotopes (e.g. 238 U) in fresh and spent fuel. The time consumed for analysis of bulk materials by this technique is only 4 minutes. The method is also used for analysis of uranium in rock, sediment, soil, meteorites, lunar, biological, urine, archaeological, zircon sand and seawater samples. The method enables uranium in a sample to be measured without respect to its oxidation state, organic and inorganic elements

  19. Investigation of electrorefining of metallic alloy fuel onto solid Al cathodes

    International Nuclear Information System (INIS)

    Cassayre, L.; Malmbeck, R.; Masset, P.; Rebizant, J.; Serp, J.; Soucek, P.; Glatz, J.-P.

    2007-01-01

    This work concerned the electrorefining of UZr and UPuZr alloys on a solid aluminium cathode, in the LiCl-KCl eutectic melt containing U 3+ , Pu 3+ , Np 3+ , Zr 2+ or Zr 4+ , Am 3+ , Nd 3+ , Y 3+ , Ce 3+ and Gd 3+ chlorides. During constant current electrolyses, the use of a cathodic cut-off potential (-1.25 V versus Ag/AgCl) allowed to selectively deposit actinides (mainly U), while lanthanides remained in the salt. The aim was to determine the maximal load achievable on a single aluminium electrode. The total exchange charge was 4300 C, which represents the deposition of 3.72 g of actinides in 4.17 g Al, yielding a composition of 44.6 wt% An in Al. It was shown that the melting of the cathode contributed to increase the total amount of actinides deposited on the aluminium

  20. Evaluation of covariances for resolved resonance parameters of 235U, 238U, and 239Pu in JENDL-3.2

    International Nuclear Information System (INIS)

    Kawano, Toshihiko; Shibata, Keiichi

    2003-02-01

    Evaluation of covariances for resolved resonance parameters of 235 U, 238 U, and 239 Pu was carried out. Although a large number of resolved resonances are observed for major actinides, uncertainties in averaged cross sections are more important than those in resonance parameters in reactor calculations. We developed a simple method which derives a covariance matrix for the resolved resonance parameters from uncertainties in the averaged cross sections. The method was adopted to evaluate the covariance data for some important actinides, and the results were compiled in the JENDL-3.2 covariance file. (author)

  1. Growth kinetics and microstructural evolution during hot isostatic pressing of U-10 wt.% Mo monolithic fuel plate in AA6061 cladding with Zr diffusion barrier

    Science.gov (United States)

    Park, Y.; Yoo, J.; Huang, K.; Keiser, D. D.; Jue, J. F.; Rabin, B.; Moore, G.; Sohn, Y. H.

    2014-04-01

    Phase constituents and microstructure changes in RERTR fuel plate assemblies as functions of temperature and duration of hot-isostatic pressing (HIP) during fabrication were examined. The HIP process was carried out as functions of temperature (520, 540, 560 and 580 °C for 90 min) and time (45-345 min at 560 °C) to bond 6061 Al-alloy to the Zr diffusion barrier that had been co-rolled with U-10 wt.% Mo (U10Mo) fuel monolith prior to the HIP process. Scanning and transmission electron microscopies were employed to examine the phase constituents, microstructure and layer thickness of interaction products from interdiffusion. At the interface between the U10Mo and Zr, following the co-rolling, the UZr2 phase was observed to develop adjacent to Zr, and the α-U phase was found between the UZr2 and U10Mo, while the Mo2Zr was found as precipitates mostly within the α-U phase. The phase constituents and thickness of the interaction layer at the U10Mo-Zr interface remained unchanged regardless of HIP processing variation. Observable growth due to HIP was only observed for the (Al,Si)3Zr phase found at the Zr/AA6061 interface, however, with a large activation energy of 457 ± 28 kJ/mole. Thus, HIP can be carried to improve the adhesion quality of fuel plate without concern for the excessive growth of the interaction layer, particularly at the U10Mo-Zr interface with the α-U, Mo2Zr, and UZr2 phases.

  2. Growth kinetics and microstructural evolution during hot isostatic pressing of U-10 wt.% Mo monolithic fuel plate in AA6061 cladding with Zr diffusion barrier

    International Nuclear Information System (INIS)

    Park, Y.; Yoo, J.; Huang, K.; Keiser, D.D.; Jue, J.F.; Rabin, B.; Moore, G.; Sohn, Y.H.

    2014-01-01

    Phase constituents and microstructure changes in RERTR fuel plate assemblies as functions of temperature and duration of hot-isostatic pressing (HIP) during fabrication were examined. The HIP process was carried out as functions of temperature (520, 540, 560 and 580 °C for 90 min) and time (45–345 min at 560 °C) to bond 6061 Al-alloy to the Zr diffusion barrier that had been co-rolled with U-10 wt.% Mo (U10Mo) fuel monolith prior to the HIP process. Scanning and transmission electron microscopies were employed to examine the phase constituents, microstructure and layer thickness of interaction products from interdiffusion. At the interface between the U10Mo and Zr, following the co-rolling, the UZr 2 phase was observed to develop adjacent to Zr, and the α-U phase was found between the UZr 2 and U10Mo, while the Mo 2 Zr was found as precipitates mostly within the α-U phase. The phase constituents and thickness of the interaction layer at the U10Mo-Zr interface remained unchanged regardless of HIP processing variation. Observable growth due to HIP was only observed for the (Al,Si) 3 Zr phase found at the Zr/AA6061 interface, however, with a large activation energy of 457 ± 28 kJ/mole. Thus, HIP can be carried to improve the adhesion quality of fuel plate without concern for the excessive growth of the interaction layer, particularly at the U10Mo-Zr interface with the α-U, Mo 2 Zr, and UZr 2 phases

  3. Bidirectional threshold switching characteristics in Ag/ZrO2/Pt electrochemical metallization cells

    Directory of Open Access Journals (Sweden)

    Gang Du

    2016-08-01

    Full Text Available A bidirectional threshold switching (TS characteristic was demonstrated in Ag/ZrO2/Pt electrochemical metallization cells by using the electrochemical active Ag electrode and appropriate programming operation strategies The volatile TS was stable and reproducible and the rectify ratio could be tuned to ∼107 by engineering the compliance current. We infer that the volatile behavior is essentially due to the moisture absorption in the electron beam evaporated films, which remarkably improved the anodic oxidation as well as the migration of Ag+ ions. The resultant electromotive force would act as a driving force for the metal filaments dissolution, leading to the spontaneous volatile characteristics. Moreover, conductance quantization behaviors were also achieved owing to formation and annihilation of atomic scale metal filaments in the film matrix. Our results illustrate that the Ag/ZrO2/Pt device with superior TS performances is a promising candidate for selector applications in passive crossbar arrays.

  4. Maps showing the development of the Pu'u 'O'o-Kupaianaha flow field, June 1984-February 1987, Kilauea Volcano, Hawaii

    Science.gov (United States)

    Heliker, Christina; Ulrich, George E.; Margriter, Sandy C.; Hoffmann, John P.

    2001-01-01

    The Pu'u 'O'o - Kupaianaha eruption on the middle east rift zone of Kilauea began in January 1983 with intermittent activity along several fissures. By June 1983, the eruption had localized at the Pu'u 'O'o vent, and the activity settled into an increasingly regular pattern of brief eruptive episodes characterized by high lava fountains. The first 18 months of this eruption are chronicled in Wolfe and others (1988), which includes maps of the flows erupted in episodes 1-20. The maps presented here extend this series through the beginning of episode 48.

  5. Direct observation of shear–induced nanocrystal attachment and coalescence in CuZr-based metallic glasses: TEM investigation

    International Nuclear Information System (INIS)

    Hajlaoui, K.; Alrasheedi, Nashmi H.; Yavari, A.R.

    2016-01-01

    In-situ tensile straining tests were performed in a transmission electron microscope (TEM) to analyse the deformation processes in CuZr-based metallic glasses and to directly observe the phase transformation occurrence. We report evidence of shear induced coalescence of nanocrystals in the vicinity of deformed regions. Nanocrystals grow in shear bands, come into contact, being attached and progressively coalesce under applied shear stress. - Highlights: • In-situ tensile straining test in TEM was investigated on CuZr-Based metallic glass. • Strain induces nanocrystallization and subsequent attachment and coalescence of nanocrystals. • The coalescence of nanocrystals compensates strain softening in metallic glasses.

  6. Effect of Ge addition on mechanical properties and fracture behavior of Cu-Zr-Al bulk metallic glass

    International Nuclear Information System (INIS)

    Malekan, M.; Shabestari, S.G.; Gholamipour, R.; Seyedein, S.H.

    2009-01-01

    Effect of the addition of a small amount of Ge on mechanical properties and fracture behavior of Cu 50 Zr 43 Al 7 (at.%) bulk metallic glass were studied. The Cu 50 Zr 43 Al 7 alloy has a surprising glass-forming ability (GFA), and the glassy rods up to 4 mm in diameter can be formed. Partial addition of Ge causes the crystalline phases precipitate in the glassy matrix of (Cu 50 Zr 43 Al 7 ) 100-x Ge x (x = 0, 1, 2) rods with a diameter of 4 mm. In uniaxial compression, Cu 50 Zr 43 Al 7 bulk metallic glass exhibit high strength of 1692 MPa and very limited plasticity of 0.05%. When Ge increases from 0 to 2 at.%, the strength decreases, but plastic strain increases about 2.5%. Fracture surface and shear bands of samples were investigated by scanning electron microscopy (SEM).

  7. Combinatorial development of antibacterial Zr-Cu-Al-Ag thin film metallic glasses.

    Science.gov (United States)

    Liu, Yanhui; Padmanabhan, Jagannath; Cheung, Bettina; Liu, Jingbei; Chen, Zheng; Scanley, B Ellen; Wesolowski, Donna; Pressley, Mariyah; Broadbridge, Christine C; Altman, Sidney; Schwarz, Udo D; Kyriakides, Themis R; Schroers, Jan

    2016-05-27

    Metallic alloys are normally composed of multiple constituent elements in order to achieve integration of a plurality of properties required in technological applications. However, conventional alloy development paradigm, by sequential trial-and-error approach, requires completely unrelated strategies to optimize compositions out of a vast phase space, making alloy development time consuming and labor intensive. Here, we challenge the conventional paradigm by proposing a combinatorial strategy that enables parallel screening of a multitude of alloys. Utilizing a typical metallic glass forming alloy system Zr-Cu-Al-Ag as an example, we demonstrate how glass formation and antibacterial activity, two unrelated properties, can be simultaneously characterized and the optimal composition can be efficiently identified. We found that in the Zr-Cu-Al-Ag alloy system fully glassy phase can be obtained in a wide compositional range by co-sputtering, and antibacterial activity is strongly dependent on alloy compositions. Our results indicate that antibacterial activity is sensitive to Cu and Ag while essentially remains unchanged within a wide range of Zr and Al. The proposed strategy not only facilitates development of high-performing alloys, but also provides a tool to unveil the composition dependence of properties in a highly parallel fashion, which helps the development of new materials by design.

  8. Devitrification kinetics and phase selection mechanisms in Cu-Zr metallic glasses

    Energy Technology Data Exchange (ETDEWEB)

    Kalay, Ilkay [Iowa State Univ., Ames, IA (United States)

    2010-01-01

    Metallic glasses have been a promising class of materials since their discovery in the 1960s. Indeed, remarkable chemical, mechanical and physical properties have attracted considerable attention, and several excellent reviews are available. Moreover, the special group of glass forming alloys known as the bulk metallic glasses (BMG) become amorphous solids even at relatively low cooling rates, allowing them to be cast in large cross sections, opening the scope of potential applications to include bulk forms and net shape structural applications. Recent studies have been reported for new bulk metallic glasses produced with lower cooling rates, from 0.1 to several hundred K/s. Some of the application products of BMGs include sporting goods, high performance springs and medical devices. Several rapid solidification techniques, including melt-spinning, atomization and surface melting have been developed to produce amorphous alloys. The aim of all these methods is to solidify the liquid phase rapidly enough to suppress the nucleation and growth of crystalline phases. Furthermore, the production of amorphous/crystalline composite (ACC) materials by partial crystallization of amorphous precursor has recently given rise to materials that provide better mechanical and magnetic properties than the monolithic amorphous or crystalline alloys. In addition, these advances illustrate the broad untapped potential of using the glassy state as an intermediate stage in the processing of new materials and nanostructures. These advances underlie the necessity of investigations on prediction and control of phase stability and microstructural dynamics during both solidification and devitrification processes. This research presented in this dissertation is mainly focused on Cu-Zr and Cu-Zr-Al alloy systems. The Cu-Zr binary system has high glass forming ability in a wide compositional range (35-70 at.% Cu). Thereby, Cu-Zr based alloys have attracted much attention according to fundamental

  9. Growth kinetics and microstructural evolution during hot isostatic pressing of U-10 wt.% Mo monolithic fuel plate in AA6061 cladding with Zr diffusion barrier

    Energy Technology Data Exchange (ETDEWEB)

    Park, Y.; Yoo, J.; Huang, K. [Advanced Materials Processing and Analysis Center, Department of Materials Science and Engineering, University of Central Florida, Orlando, FL 32816 (United States); Keiser, D.D.; Jue, J.F.; Rabin, B.; Moore, G. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83401 (United States); Sohn, Y.H., E-mail: Yongho.sohn@ucf.edu [Advanced Materials Processing and Analysis Center, Department of Materials Science and Engineering, University of Central Florida, Orlando, FL 32816 (United States)

    2014-04-01

    Phase constituents and microstructure changes in RERTR fuel plate assemblies as functions of temperature and duration of hot-isostatic pressing (HIP) during fabrication were examined. The HIP process was carried out as functions of temperature (520, 540, 560 and 580 °C for 90 min) and time (45–345 min at 560 °C) to bond 6061 Al-alloy to the Zr diffusion barrier that had been co-rolled with U-10 wt.% Mo (U10Mo) fuel monolith prior to the HIP process. Scanning and transmission electron microscopies were employed to examine the phase constituents, microstructure and layer thickness of interaction products from interdiffusion. At the interface between the U10Mo and Zr, following the co-rolling, the UZr{sub 2} phase was observed to develop adjacent to Zr, and the α-U phase was found between the UZr{sub 2} and U10Mo, while the Mo{sub 2}Zr was found as precipitates mostly within the α-U phase. The phase constituents and thickness of the interaction layer at the U10Mo-Zr interface remained unchanged regardless of HIP processing variation. Observable growth due to HIP was only observed for the (Al,Si){sub 3}Zr phase found at the Zr/AA6061 interface, however, with a large activation energy of 457 ± 28 kJ/mole. Thus, HIP can be carried to improve the adhesion quality of fuel plate without concern for the excessive growth of the interaction layer, particularly at the U10Mo-Zr interface with the α-U, Mo{sub 2}Zr, and UZr{sub 2} phases.

  10. The origin of magnetism in transition metal-doped ZrO2 thin films: Experiment and theory

    KAUST Repository

    Hong, Nguyenhoa

    2013-10-04

    We have investigated the magnetic properties of Fe/Co/Ni-doped ZrO 2 laser ablated thin films in comparison with the known results of Mn-doped ZrO2, which is thought to be a promising material for spintronics applications. It is found that doping with a transition metal can induce room temperature ferromagnetism in \\'fake\\' diamond. Theoretical analysis based on density functional theory confirms the experimental measurements, by revealing that the magnetic moments of Mn- and Ni-doped ZrO2 thin films are much larger than that of Fe- or Co-doped ZrO2 thin films. Most importantly, our calculations confirm that Mn- and Ni-doped ZrO2 show a ferromagnetic ground state in comparison to Co- and Fe-doped ZrO 2, which favor an antiferromagnetic ground state. © 2013 IOP Publishing Ltd.

  11. Cost comparison of 4x500 MW coal-fuelled and 4x850 MW CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Costa, M.

    1981-01-01

    The lifetime costs for a 4x850 MW CANDU generating station are compared to those for 4x500 MW bituminous coal-fuelled generating stations. Two types of coal-fuelled stations are considered; one burning U.S. coal which includes flue gas desulfurization and one burning Western Canadian coal. Current estimates for the capital costs, operation and maintenance costs, fuel costs, decommissioning costs and irradiated fuel management costs are shown. The results show: (1) The accumulated discounted costs of nuclear generation, although initially higher, are lower than coal-fuelled generation after two or three years. (2) Fuel costs provide the major contribution to the total lifetime costs for coal-fuelled stations whereas capital costs are the major item for the nuclear station. (3) The break even lifetime capacity factor between nuclear and U.S. coal-fuelled generation is projected to be 5%; that for nuclear and Canadian coal-fuelled generation is projected to be 9%. (4) Large variations in the costs are required before the cost advantage of nuclear generation is lost. (5) Comparison with previous results shows that the nuclear alternative has a greater cost advantage in the current assessment. (6) The total unit energy cost remains approximately constant throughout the station life for nuclear generation while that for coal-fuelled generation increases significantly due to escalating fuel costs. The 1978 and 1979 actual total unit energy cost to the consumer for several Ontario Hydro stations are detailed, and projected total unit energy costs for several Ontario Hydro stations are shown in terms of escalated dollars and in 1980 constant dollars

  12. First-principles study of new quaternary Heusler compounds without 3d transition metal elements: ZrRhHfZ (Z = Al, Ga, In)

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiaotian [School of Material Sciences and Engineering, Hebei University of Technology, Tianjin 300130 (China); Institute for Superconducting & Electronic Materials (ISEM), University of Wollongong, Wollongong 2500 (Australia); Cheng, Zhenxiang, E-mail: cheng@uow.edu.au [Institute for Superconducting & Electronic Materials (ISEM), University of Wollongong, Wollongong 2500 (Australia); Guo, Ruikang [School of Material Sciences and Engineering, Hebei University of Technology, Tianjin 300130 (China); Wang, Jianli [Institute for Superconducting & Electronic Materials (ISEM), University of Wollongong, Wollongong 2500 (Australia); Rozale, Habib [Condensed Matter and Sustainable Development Laboratory, Physics Department, University of Sidi-Bel-Abbès, 22000 Sidi-Bel-Abbès (Algeria); Wang, Liying [Department of Physics, Tianjin University, Tianjin 300350 (China); Yu, Zheyin [Institute for Superconducting & Electronic Materials (ISEM), University of Wollongong, Wollongong 2500 (Australia); Liu, Guodong, E-mail: gdliu1978@126.com [School of Material Sciences and Engineering, Hebei University of Technology, Tianjin 300130 (China)

    2017-06-01

    Plane-wave pseudo-potential methods based on density functional theory are employed to investigate the electronic structures, and the magnetic and half-metallic properties of the newly designed quaternary Heusler compounds ZrRhHfZ (Z = Al, Ga, In) without 3d transition metal elements. The calculated results show that ZrRhHfZ (Z = Al, Ga, In) compounds are half-metallic, with 100% spin polarization around the Fermi level. The structural stability of these compounds has been tested from the aspects of their cohesion energy and formation. The spin-flip/half-metallic gaps of ZrRhHfZ (Z = Al, Ga, In) compounds are quite large, with values of 0.2548 eV, 0.3483 eV, and 0.2866 eV, respectively. These compounds show Slater-Pauling behavior, and the total spin magnetic moment per unit cell (M{sub t}) scales with the total number of valence electrons (Z{sub t}) following the rule: M{sub t} = Z{sub t} - 18. The magnetization of ZrRhHfZ (Z = Al, Ga, In) compounds mainly comes from the 4d electrons of the Zr atoms and the 5d electrons of the Hf atoms. Furthermore, the effects of uniform strain and tetragonal deformation on the half metallicity has been investigated in detail, which is important for practical application. Finally, we reveal that the half-metallicity can be maintained when the Coulomb interactions are considered. - Highlights: • New quaternary compounds without 3d transition metal elements have been designed. • The electronic structures and magnetism of the ZrRhHfZ compounds have been studied. • The effect of strain on the half-metallic behavior has been tested. • The effect of the Coulomb interactions on the half-metallicity has been investigated.

  13. Assessment of heavy metal contamination of dust at some selected fuel filling stations in Accra

    International Nuclear Information System (INIS)

    Afrifa, C. G.

    2011-07-01

    Heavy metal contaminated dust particles of fuel filling stations can be re-suspended into the ambient air and serve as a source of atmospheric pollution since the fine particles are aerodynamic and have longer life time in ambient air. This can cause ill-health effect on the fuel attendants and residents within the neighbourhood especially infants and the aged who are more vulnerable. In spite of this, not much research has been done on heavy metal contamination of dust at fuel filling stations. In this study, 55 dust samples were collected from six fuel filling stations in the Ga-East district and Accra Metropolitan assembly, both in Accra, in order to assess the levels of contamination of heavy metals; their possible sources and the human health risk associated with them. The dust samples were divided into two parts with one part sieved into four fractions using metric mesh sizes 500 µm, 200µm, 100µm and 45 µm, and pulverised. Total concentrations of heavy metals (Si, K, Ca, Ti, V, Cr, Mn, Fe, Ni, Cu, Zn, Ga, Br, Rb, Sr, Y, Zr and Pb) were determined in the dust samples using energy dispersive X-ray fluorescence analysis. The pollution indices; enrichment factors (EF), index of geoaccumulation (Igeo), contamination factor (CF) and pollution load index (PLI) were used to identify possible levels of pollution from anthropogenic sources. The possible sources of metals were also identified with principal component analysis. Noncancer effect of children and adults due to exposure to dust from these fuel filling stations were also estimated. For the three fuelling areas, the average concentrations of V, Cr, Ni and Cu exceeded the acceptable values in common soil in the <45 µm fraction. The average concentration of Zn however exceeded the acceptable value only at the mixed-fuel fuelling area whereas the average concentration of Pb was within the acceptable value for all three fuelling areas. The dust samples showed moderate to significant enrichments for V, Cu, Br

  14. A measurement of the density and compressibility of (U, Pu)-mixed oxide at 3432 kJ/kg (7356 K)

    International Nuclear Information System (INIS)

    Breitung, W.

    1987-05-01

    In a transient in-pile heating test the density of liquid (U, Pu)-mixed oxide at 3432±103 kJ/kg was measured to be 5027±132 kg/m 3 . The corresponding temperature is estimated to 7356±212 K. The isothermal compressibility of the mixed oxide was evaluated to 2.85 (±1.65) x 10 -4 /MPa at the same temperature. Based on these new data and previously existing measurements, new relations are proposed for the following properties of liquid UO 2 and (U, Pu)O 2 , as well: density - enthalpy, density - temperature, thermal expansion - temperature, and isothermal compressibility - temperature. (orig.) [de

  15. Assessment of solid/liquid equilibria in the (U, Zr)O2+y system

    Science.gov (United States)

    Mastromarino, S.; Seibert, A.; Hashem, E.; Ciccioli, A.; Prieur, D.; Scheinost, A.; Stohr, S.; Lajarge, P.; Boshoven, J.; Robba, D.; Ernstberger, M.; Bottomley, D.; Manara, D.

    2017-10-01

    Solid/liquid equilibria in the system UO2sbnd ZrO2 are revisited in this work by laser heating coupled with fast optical thermometry. Phase transition points newly measured under inert gas are in fair agreement with the early measurements performed by Wisnyi et al., in 1957, the only study available in the literature on the whole pseudo-binary system. In addition, a minimum melting point is identified here for compositions near (U0.6Zr0.4)O2+y, around 2800 K. The solidus line is rather flat on a broad range of compositions around the minimum. It increases for compositions closer to the pure end members, up to the melting point of pure UO2 (3130 K) on one side and pure ZrO2 (2970 K) on the other. Solid state phase transitions (cubic-tetragonal-monoclinic) have also been observed in the ZrO2-rich compositions X-ray diffraction. Investigations under 0.3 MPa air (0.063 MPa O2) revealed a significant decrease in the melting points down to 2500 K-2600 K for increasing uranium content (x(UO2)> 0.2). This was found to be related to further oxidation of uranium dioxide, confirmed by X-ray absorption spectroscopy. For example, a typical oxidised corium composition U0.6Zr0.4O2.13 was observed to solidify at a temperature as low as 2493 K. The current results are important for assessing the thermal stability of the system fuel - cladding in an oxide based nuclear reactor, and for simulating the system behaviour during a hypothetical severe accident.

  16. Some features of formation and dissolution of a series of Pu(IV) and Zr alkyl and butyl alkyl phosphates in the system TBP -n-dodecane - nitric acid - water

    International Nuclear Information System (INIS)

    Markov, G.S.; Moshkov, M.M.; Kokina, S.A.

    1990-01-01

    The formation and composition of salts produced on interaction of a series of alkyl- and butylalkylphosphoric acids having alkyl radical chain lengths from C 4 to C 1 0 with Pu(IV) and Zr in organic and aqueous phases of the system TBP - n-dodecane -nitric acid - water were studied. The composition of compounds was found to depend on the conditions of their formation, defined first of all by the HNO 3 concentration in aqueous and organic phases. (author) 12 refs.; 4 figs.; 1 tab

  17. Phase stability, magnetic, electronic, half-metallic and mechanical properties of a new equiatomic quaternary Heusler compound ZrRhTiIn: A first-principles investigation

    Science.gov (United States)

    Wang, Jia-Xing; Chen, Z. B.; Gao, Y. C.

    2018-05-01

    In this manuscript, we have studied the electronic, magnetic, half-metallic and mechanical properties of a new Zr-based equiatomic quaternary Heusler (EQH) compound, ZrRhTiIn using first-principles calculations. The generalized gradient approximation (GGA) calculation results imply that at its equilibrium lattice constant of 6.70 Å, ZrRhTiIn is a half-metallic material (HMM) with a considerable band gap (Ebg) of 0.530 eV and a spin-filter/half-metallic band-gap (EHM) of 0.080 eV in the minority-spin channel. For ZrRhTiIn, the formation energy of -2.738 eV and the cohesive energy of 21.38 eV indicate that it is a thermodynamically stable material according to theory. The minority-spin EHM arises from the hybridization among Zr-4d, Ti-3d and Rh-4d electrons. The calculated total magnetic moment of ZrRhTiIn is 2 μB, meeting the well-known Slater-Pauling rule Mt = Zt -18. Furthermore, uniform strain and tetragonal strain were applied in this work to examine the magneto-electronic and half-metallic behaviors of the ZrRhTiIn system. Finally, we show that ZrRhTiIn is mechanically stable, ductile and anisotropic.

  18. Studies and manufacture of plutonium fuel

    International Nuclear Information System (INIS)

    Bussy, P.; Mustelier, J.P.; Pascard, R.

    1964-01-01

    The studies carried out at the C.E.A. on the properties of fast neutron reactor fuels, the manufacture of fuel elements and their behaviour under irradiation are broadly outlined. The metal fuels studied are the ternary alloys U Pu Mo, U Pu Nb, U Pa Ti, U Pa Zr, the ceramic fuels being mixed uranium and plutonium oxides, carbides and nitrides obtained by sintering. Results are given on the manufacture of uranium fuel elements containing a small proportion of plutonium, used in a critical experiment, and on the first experiments in the manufacture of fuel elements for the reactor Rapsodie. Finally the results of irradiation tests carried out on the prototype fuel pins for Rapsodie are described. (authors) [fr

  19. Synthesis report of a reactor-physics programme devoted to the qualification of the project calculation methods applied to the mixed (U-Pu) fuelled reloadings

    International Nuclear Information System (INIS)

    Golinelli, C.

    1982-01-01

    The Commission of the European Communities has sponsored three sets of experiments with a contract in the framework of the LWR plutonium recycling programme. The aim of this experimental programme was to supplement the qualification of the computer codes used in connection with control and safety studies relating to plutonium recycling in LWR power stations. The experiments are performed at the Cadarache centre by CEA and EdF in the MINERVE and EOLE facilities. The power distributions in various kinds of assemblies with some heterogeneities (water holes, absorbers, UO 2 /UO 2 -PuO 2 interfaces) are studied in MINERVE as well as the DOPPLER effect. This second programme consisted in oscillating the UO 2 and UO 2 -PuO 2 samples which were heated between 20 0 C and 800 0 C. The CREOLE experiments in EOLE dealt with the moderator temperature effect from 20 0 C to 300 0 C. All the results show that the calculation schema for the hot point values must be sophisticated and that it is possible to reduce the uncertainties concerning the feedback factors

  20. Experimental study of the lifetime and phase transition in neutron-rich Zr 98 ,100 ,102

    Science.gov (United States)

    Ansari, S.; Régis, J.-M.; Jolie, J.; Saed-Samii, N.; Warr, N.; Korten, W.; Zielińska, M.; Salsac, M.-D.; Blanc, A.; Jentschel, M.; Köster, U.; Mutti, P.; Soldner, T.; Simpson, G. S.; Drouet, F.; Vancraeyenest, A.; de France, G.; Clément, E.; Stezowski, O.; Ur, C. A.; Urban, W.; Regan, P. H.; Podolyák, Zs.; Larijani, C.; Townsley, C.; Carroll, R.; Wilson, E.; Mach, H.; Fraile, L. M.; Paziy, V.; Olaizola, B.; Vedia, V.; Bruce, A. M.; Roberts, O. J.; Smith, J. F.; Scheck, M.; Kröll, T.; Hartig, A.-L.; Ignatov, A.; Ilieva, S.; Lalkovski, S.; Mǎrginean, N.; Otsuka, T.; Shimizu, N.; Togashi, T.; Tsunoda, Y.

    2017-11-01

    Rapid shape changes are observed for neutron-rich nuclei with A around 100. In particular, a sudden onset of ground-state deformation is observed in the Zr and Sr isotopic chains at N = 60: Low-lying states in N ≤58 nuclei are nearly spherical, while those with N ≥60 have a rotational character. Nuclear lifetimes as short as a few picoseconds can be measured using fast-timing techniques with LaBr3(Ce) scintillators, yielding a key ingredient in the systematic study of the shape evolution in this region. We used neutron-induced fission of 241Pu and 235U to study lifetimes of excited states in fission fragments in the A ˜100 region with the EXILL-FATIMA array located at the PF1B cold neutron beam line at the Institut Laue-Langevin. In particular, we applied the generalized centroid difference method to deduce lifetimes of low-lying states for the nuclei 98Zr (N = 58), 100Zr, and 102Zr (N ≥60 ). The results are discussed in the context of the presumed phase transition in the Zr chain by comparing the experimental transition strengths with the theoretical calculations using the interacting boson model and the Monte Carlo shell model.

  1. Transient feedback from fuel motion in metal IFR [Integral Fast Reactor] fuel

    International Nuclear Information System (INIS)

    Rhodes, E.A.; Stanford, G.S.; Regis, J.P.; Bauer, T.H.; Dickerman, C.E.

    1990-01-01

    Results from hodoscope data analyses are presented for TREAT transient-overpower tests M5 through M7 with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding branch and prefailure elongation of D9-clad ternary (U-Pu-Zr) IFR-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT9-clad binary (U-Zr) FFTF-driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure. 4 refs., 6 figs

  2. Comminution by hydriding-dehydriding process of the U-Zr-Nb alloys stabilized at different phases by aging heat treatment

    International Nuclear Information System (INIS)

    Cantagalli, Natalia Mattar; Pais, Rafael Witter Dias; Braga, Daniel Martins; Santos, Ana Maria Matildes dos; Ferraz, Wilmar Barbosa

    2011-01-01

    Powders of the U-Zr-Nb alloys are raw materials for obtaining plate-type dispersion fuel of high density and medium enrichment for research and test reactors as well as small power reactors. U-2.5Zr-7.5Nb and U-3Zr-9Nb (wt%) alloys, initially homogenized at high temperatures, were transformed at different phases by means aging heat treatments, and then comminuted by hydriding-dehydriding process to powder production. The phases transformations were obtained by the homogenization of the U-2.5Zr-7.5Nb and U-3Zr-9Nb alloys at high temperatures (1000 deg C for 1 and 16 h), followed by aging heat treatment at 600 deg C, in times of 0.5 h, 3.0 h and 24h, and subsequently quenched in water to stabilize the desired phase. The comminution process was performed at 200 deg C for different times ranging from 20 minutes to 4 hours. The powders were then characterized by scanning electron microscopy, X-ray diffraction and determination of particle size distribution by means of laser equipment CILAS. One of the main objectives of this study was to verify the influence of the different phases in the characteristics of the obtained powders. It was found that alloys stabilized in gamma phase produced powders with smaller particles sizes than those with cellular structure of the α and γ phases. Regardless of retained phases, the produced powders consist of agglomerates with irregular morphology. (author)

  3. Comminution by hydriding-dehydriding process of the U-Zr-Nb alloys stabilized at different phases by aging heat treatment

    Energy Technology Data Exchange (ETDEWEB)

    Cantagalli, Natalia Mattar; Pais, Rafael Witter Dias; Braga, Daniel Martins; Santos, Ana Maria Matildes dos; Ferraz, Wilmar Barbosa, E-mail: ferrazw@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG) Belo Horizonte, MG (Brazil)

    2011-07-01

    Powders of the U-Zr-Nb alloys are raw materials for obtaining plate-type dispersion fuel of high density and medium enrichment for research and test reactors as well as small power reactors. U-2.5Zr-7.5Nb and U-3Zr-9Nb (wt%) alloys, initially homogenized at high temperatures, were transformed at different phases by means aging heat treatments, and then comminuted by hydriding-dehydriding process to powder production. The phases transformations were obtained by the homogenization of the U-2.5Zr-7.5Nb and U-3Zr-9Nb alloys at high temperatures (1000 deg C for 1 and 16 h), followed by aging heat treatment at 600 deg C, in times of 0.5 h, 3.0 h and 24h, and subsequently quenched in water to stabilize the desired phase. The comminution process was performed at 200 deg C for different times ranging from 20 minutes to 4 hours. The powders were then characterized by scanning electron microscopy, X-ray diffraction and determination of particle size distribution by means of laser equipment CILAS. One of the main objectives of this study was to verify the influence of the different phases in the characteristics of the obtained powders. It was found that alloys stabilized in gamma phase produced powders with smaller particles sizes than those with cellular structure of the {alpha} and {gamma} phases. Regardless of retained phases, the produced powders consist of agglomerates with irregular morphology. (author)

  4. Thermodynamic stabilities of MO2+x(s) (M = U, Np, Pu and Am), pourbaix diagrams

    International Nuclear Information System (INIS)

    Vitorge, Pierre; Faure, Marie-Helene; Vercouter, Thomas; Capdevila, Helene; Maillard, Serge

    2002-01-01

    The experimental solubilities of the hydrated amorphous freshly precipitated M(OH) z (am) and MO 2 (OH) z (am) compounds are often used as an upper limit for the safety assessments of deep waste repositories, since these compounds slowly transform to less soluble ones, as typically M(OH) 4 (am) to MO 2 (cr). Solubility (vs. redox potential) at pH=8, and E-pH predominance diagrams are plotted in aqueous solutions at 25degC by using thermodynamic data recently selected by the NEA-TDB review, or estimated by using classical chemical analogies for the non-redox reactions. The solubilities and relative stabilities are also calculated for the MO 2+x (s) crystalline compounds of known stabilities: U 4 O 9 (s), U 3 O 7 (s), U 3 O 8 (s) and Np 2 O 5 (s) where 2+x = 2.25, 2.33, 2.67 and 2.5 respectively. The stabilities of the other MO 2+x (s) compounds are estimated by analogy: M 4 O 9 (s) (M=U, Np, Pu), M 3 O 7 (s) and M 3 O 8 (s) (M=U, Pu), and M 2 O 5 (s) (M=Np, Am) are predicted to be more stable (i.e. less soluble), than the amorphous hydroxides. However their precipitation have never been observed at room temperature possibly for kinetic reasons or difficulties in interpreting solubility experiments. (author)

  5. Lattice vibrational properties of transition metal carbides (TiC, ZrC

    Indian Academy of Sciences (India)

    Lattice vibrational properties of transition metal carbides (TiC, ZrC and HfC) have been presented by including the effects of free-carrier doping and three-body interactions in the rigid shell model. The short-range overlap repulsion is operative up to the second neighbour ions. An excellent agreement has been obtained ...

  6. Promising Ta-Ti-Zr-Si metallic glass coating without cytotoxic elements for bio-implant applications

    Science.gov (United States)

    Lai, J. J.; Lin, Y. S.; Chang, C. H.; Wei, T. Y.; Huang, J. C.; Liao, Z. X.; Lin, C. H.; Chen, C. H.

    2018-01-01

    Tantalum (Ta) is considered as one of the most promising metal due to its high corrosion resistance, excellent biocompatibility and cell adhesion/in-growth capabilities. Although there are some researches exploring the biomedical aspects of Ta and Ta based alloys, systematic characterizations of newly developed Ta-based metallic glasses in bio-implant applications is still lacking. This study employs sputtering approach to produced thin-film Ti-based metallic glasses due to the high melting temperature of Ta (3020 °C). Two fully amorphous Ta-based metallic glasses composed of Ta57Ti17Zr15Si11 and Ta75Ti10Zr8Si7 are produced and experimentally characterized in terms of their mechanical properties, bio-corrosion properties, surface hydrophilic characteristics, and in-vitro cell viability and cells attachment tests. Compare to conventional pure Ti and Ta metals, the developed Ta-based metallic glasses exhibit higher hardness and lower modulus which are better match to the mechanical properties of bone. MTS assay results show that Ta-based metallic glasses show comparable cell viability and cell attachment rate compared to that of pure Ti and Ta surface in a 72 h in-vitro test.

  7. The application of photoelectron spectroscopy in the study of corrosion and oxidation mechanisms of alloys: Inconel 182, Fe/Cu(100 and U-Zr-Nb

    International Nuclear Information System (INIS)

    Mendonca, Renato de

    2011-05-01

    In. this work a study of the oxidation/corrosion process of three systems of metallic materials by Photoemission Spectroscopy is presented. In the first system, it was investigated the corrosion of Ineonel 182 at simulated Pressurized Water Reactor (PWR) environment. Samples with and without surface chemical treatment were exposed to the simulated environment for until 18 weeks. The oxide layer formed on the surfaces of the samples at different conditions was characterized by Scanning Electron Microscopy and XPS coupled with argon ion sputtering. The comparison between the oxide films grown on the samples showed that the oxide layer formed on the chemically treated sample is thinner and relatively Cr-rich. In second system it was studied the initial oxidation at room temperature of epitaxial films of Fe evapored on Cu (100). The films were deposited with two different thicknesses in order to get tbe fcc Fe (100) and bcc Fe (110) surfaces. The results, obtained by photoemission spectroscopy at the TEMPO beamline of the Synchrotron Soleil, showed the formation of distinct oxides films. The surfaces also presented different kinetics of oxidation and the (110) Fe-bcc showed highest reactivity. The analysis of the data indicated the Fe 1-x O formation on fcc Fe (100) and suggested the Fe 1-x O and FC 3 0 4 formation on (110) Fe-bcc surface. In the last system, it was investigated the initial oxidation of U-Zr-Nb alloys at room temperature. For this experiment, the alloys were exposed to oxygen in ultra high vacuum. The analysis of the U 4f peak showed the fast formation of U) 2 on the surfaces and similar kinetics of oxidation between the U and the U-Zr-Nb alloy. The alloying elements showed slower oxidation. The Zr 3d peak suggested the zr0 2 formation while the Nb 3d peak showed a remarkable enlargement that became necessary a deconvolution which indicated the formation ofNhO , Nb0 2 and N 2 0 5 . (author)

  8. Decomposition of metal nitrate solutions

    International Nuclear Information System (INIS)

    Haas, P.A.; Stines, W.B.

    1982-01-01

    Oxides in powder form are obtained from aqueous solutions of one or more heavy metal nitrates (e.g. U, Pu, Th, Ce) by thermal decomposition at 300 to 800 deg C in the presence of about 50 to 500% molar concentration of ammonium nitrate to total metal. (author)

  9. Investigation of electrorefining of metallic alloy fuel onto solid Al cathodes

    Energy Technology Data Exchange (ETDEWEB)

    Cassayre, L. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany); Malmbeck, R. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany)]. E-mail: rikard.malmbeck@cec.eu.int; Masset, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany); Rebizant, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany); Serp, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany); Soucek, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany); Glatz, J.-P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. 2340, 76125 Karlsruhe (Germany)

    2007-01-15

    This work concerned the electrorefining of UZr and UPuZr alloys on a solid aluminium cathode, in the LiCl-KCl eutectic melt containing U{sup 3+}, Pu{sup 3+}, Np{sup 3+}, Zr{sup 2+} or Zr{sup 4+}, Am{sup 3+}, Nd{sup 3+}, Y{sup 3+}, Ce{sup 3+} and Gd{sup 3+} chlorides. During constant current electrolyses, the use of a cathodic cut-off potential (-1.25 V versus Ag/AgCl) allowed to selectively deposit actinides (mainly U), while lanthanides remained in the salt. The aim was to determine the maximal load achievable on a single aluminium electrode. The total exchange charge was 4300 C, which represents the deposition of 3.72 g of actinides in 4.17 g Al, yielding a composition of 44.6 wt% An in Al. It was shown that the melting of the cathode contributed to increase the total amount of actinides deposited on the aluminium.

  10. Influence of alkali metal hydroxides on corrosion of Zr-based alloys

    International Nuclear Information System (INIS)

    Jeong, Y.H.; Ruhmann, H.; Garzarolli, F.

    1997-01-01

    In this study the influence of group-1 alkali hydroxides on different zirconium based alloys has been evaluated. The experiments have been carried out in small stainless steel autoclaves at 350 deg. C in pressurized 17 MPa water, with in low (0.32 mmol), medium (4.3 mmol) and high (31.5 mmol) equimolar concentrations of Li-, Na-, K-, Rb- and Cs-Hydroxides. Two types of alloys have been investigated: Zr-Sn-(Transition metal) and Zr-Sn-Nb-(Transition metal). The corrosion behaviour was evaluated from weight gain measurements. From the experiments the cation could be identified as the responsible species for zirconium alloy corrosion in alkalized water. The radius of the cation governs the corrosion behaviour in the pre accelerated region of zircaloy corrosion. Incorporating of alkali cations into the zirconium oxide lattice is probably the mechanism which allows the corrosion enhancement for Li and Na and the significantly lower effect for the other bases. Nb containing alloys show lower corrosion resistance than alloys from the Zr-Sn-TRM system in all alkali solutions. Both types of alloys corrode significantly more in LiOH and NaOH than in the other alkali environments. Lowest corrosive aggressiveness has been found for CsOH followed by KOH. Concluding from the corrosion behaviour in the different alkali environments and taking into account the tendency to promote accelerate corrosion, CsOH and KOH are possible alternate alkalis for PWR application. (author). 17 refs, 15 figs, 5 tabs

  11. Influence of alkali metal hydroxides on corrosion of Zr-based alloys

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Y H [Korea Atomic Energy Research Inst., Dae Jun (Korea, Republic of); Ruhmann, H; Garzarolli, F [Siemens-KWU, Power Generation Group, Erlangen (Germany)

    1997-02-01

    In this study the influence of group-1 alkali hydroxides on different zirconium based alloys has been evaluated. The experiments have been carried out in small stainless steel autoclaves at 350 deg. C in pressurized 17 MPa water, with in low (0.32 mmol), medium (4.3 mmol) and high (31.5 mmol) equimolar concentrations of Li-, Na-, K-, Rb- and Cs-Hydroxides. Two types of alloys have been investigated: Zr-Sn-(Transition metal) and Zr-Sn-Nb-(Transition metal). The corrosion behaviour was evaluated from weight gain measurements. From the experiments the cation could be identified as the responsible species for zirconium alloy corrosion in alkalized water. The radius of the cation governs the corrosion behaviour in the pre accelerated region of zircaloy corrosion. Incorporating of alkali cations into the zirconium oxide lattice is probably the mechanism which allows the corrosion enhancement for Li and Na and the significantly lower effect for the other bases. Nb containing alloys show lower corrosion resistance than alloys from the Zr-Sn-TRM system in all alkali solutions. Both types of alloys corrode significantly more in LiOH and NaOH than in the other alkali environments. Lowest corrosive aggressiveness has been found for CsOH followed by KOH. Concluding from the corrosion behaviour in the different alkali environments and taking into account the tendency to promote accelerate corrosion, CsOH and KOH are possible alternate alkalis for PWR application. (author). 17 refs, 15 figs, 5 tabs.

  12. Study of the initial oxidation of the U4Zr2Nb alloy by X-ray photoelectron spectroscopy

    International Nuclear Information System (INIS)

    Mendonca, Renato de; Ferraz, Wilmar B.; Braga, Daniel M.; Macedo, Waldemar A.A.

    2009-01-01

    In this work, the initial stages of oxidation of the U4Zr2Nb alloy in O 2 atmosphere were studied in-situ, in ultrahigh vacuum, by X-ray photoelectron spectroscopy (XPS), an advanced surface-sensitive technique. After several hours of Ar + ion-sputtering to surface cleanness, the O 2 exposures was realized on the sample at room temperature. The evolution of oxide film formed on the sample surface was followed by XPS measures, by using Mg K α radiation of 1253.6 eV and a CLAM-2 (Vacuum Generator) electron energy analyzer. The changes of U 4f, Nb 3d, Zr 3d and O 1s photoemission peaks with O 2 exposure indicate that the adsorption of oxygen on the U4Zr2Nb alloy surface leads to fast formation of UO 2 . The alloying elements show slower oxidation and different compounds are observed in Nb 3d spectra analysis. This work shows an expressive enlargement of Nb 3d peak at 100 Langmuir exposure, indicating the formation of Nb 2 O 5 and NbO in the oxide. On the other hand, the binding energy of Zr 3d suggests that the ZrO 2 formation is stable as well as uranium dioxide. (author)

  13. CANDU 9 fuelling machine carriage

    Energy Technology Data Exchange (ETDEWEB)

    Ullrich, D J; Slavik, J F [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1997-12-31

    Continuous, on-power refuelling is a key feature of all CANDU reactor designs and is essential to maintaining high station capacity factors. The concept of a fuelling machine carriage can be traced to the early CANDU designs, such as the Douglas Point Nuclear Generating Station. In the CANDU 9 480NU unit, the combination of a mobile carriage and a proven fuelling machine head design comprises an effective means of transporting fuel between the reactor and the fuel transfer ports. It is a suitable alternative to the fuelling machine bridge system that has been utilized in the CANDU 6 reactor units. The CANDU 9 480NU fuel handling system successfully combines features that meet the project requirements with respect to fuelling performance, functionality, seismic qualification and the use of proven components. The design incorporates improvements based on experience and applicable current technologies. (author). 4 figs.

  14. CANDU 9 fuelling machine carriage

    International Nuclear Information System (INIS)

    Ullrich, D.J.; Slavik, J.F.

    1996-01-01

    Continuous, on-power refuelling is a key feature of all CANDU reactor designs and is essential to maintaining high station capacity factors. The concept of a fuelling machine carriage can be traced to the early CANDU designs, such as the Douglas Point Nuclear Generating Station. In the CANDU 9 480NU unit, the combination of a mobile carriage and a proven fuelling machine head design comprises an effective means of transporting fuel between the reactor and the fuel transfer ports. It is a suitable alternative to the fuelling machine bridge system that has been utilized in the CANDU 6 reactor units. The CANDU 9 480NU fuel handling system successfully combines features that meet the project requirements with respect to fuelling performance, functionality, seismic qualification and the use of proven components. The design incorporates improvements based on experience and applicable current technologies. (author). 4 figs

  15. (236)U and (239,)(240)Pu ratios from soils around an Australian nuclear weapons test site.

    Science.gov (United States)

    Tims, S G; Froehlich, M B; Fifield, L K; Wallner, A; De Cesare, M

    2016-01-01

    The isotopes (236)U, (239)Pu and (240)Pu are present in surface soils as a result of global fallout from nuclear weapons tests carried out in the 1950's and 1960's. These isotopes potentially constitute artificial tracers of recent soil erosion and sediment movement. Only Accelerator Mass Spectrometry has the requisite sensitivity to measure all three isotopes at these environmental levels. Coupled with its relatively high throughput capabilities, this makes it feasible to conduct studies of erosion across the geographical extent of the Australian continent. In the Australian context, however, global fallout is not the only source of these isotopes. As part of its weapons development program the United Kingdom carried out a series of atmospheric and surface nuclear weapons tests at Maralinga, South Australia in 1956 and 1957. The tests have made a significant contribution to the Pu isotopic abundances present in the region around Maralinga and out to distances ∼1000 km, and impact on the assessment techniques used in the soil and sediment tracer studies. Quantification of the relative fallout contribution derived from detonations at Maralinga is complicated owing to significant contamination around the test site from numerous nuclear weapons safety trials that were also carried out around the site. We show that (236)U can provide new information on the component of the fallout that is derived from the local nuclear weapons tests, and highlight the potential of (236)U as a new fallout tracer. Crown Copyright © 2015. Published by Elsevier Ltd. All rights reserved.

  16. Analysis of U and Pu resin bead samples with a single stage mass spectrometer

    International Nuclear Information System (INIS)

    Smith, D.H.; Walker, R.L.; Bertram, L.K.; Carter, J.A.

    1979-01-01

    Resin bead sampling enables the shipment of nanogram U and Pu quantities for analysis. Application of this sampling technique to safeguards was investigated with a single-stage mass spectrometer. Standards gave results in good agreement with NBS certified values. External precisions of +-0.5% were obtained on isotopic ratios of approx. 0.01; precisions on quantitative measurements are +-1.0%

  17. Evaluation of the barrier capability of Zr-Si films with different substrate temperature for Cu metallization

    International Nuclear Information System (INIS)

    Wang Ying; Cao Fei; Ding Minghui; Shao Lei

    2009-01-01

    Barrier capability of Zr-Si diffusion barriers in Cu metallization has been investigated. Amorphous Zr-Si diffusion barriers were deposited on the Si substrates by RF reactive magnetron sputtering under various substrate temperatures. An increase in substrate temperature results in a slightly decreased deposition rate together with an increase in mass density. An increase in substrate temperature also results in grain growth as deduced from field emission scanning electron microscopy (FE-SEM) micrographs. X-ray diffraction (XRD) spectra and Auger electron spectroscopy (AES) depth profiles for Cu/Zr-Si(RT)/Si and Cu/Zr-Si(300 deg. C)/Si samples subjected to anneal at various temperatures show that the thermal stability was strongly correlated with the deposition temperature (consequently different density and chemical composition etc.) of the Zr-Si barrier layers. ZrSi(300 deg. C) with higher mass density make the Cu/Zr-Si(300 deg. C)/Si sample more stable. The appearance of Cu 3 Si in the Cu/Zr-Si/Si sample is attributed to the failure mechanism which may be associated with the diffusion of Cu and Si via the grain boundaries of the Zr-Si barriers.

  18. Effects of combined inhalation exposure of rats to 239PuO2 and beryllium metal

    International Nuclear Information System (INIS)

    Finch, G.L.; Mewhinney, J.A.; Hoover, M.D.; Haley, P.J.; Cuddihy, R.G.; Griffith, W.C.; Boecker, B.B.

    1988-01-01

    We exposed rats acutely to achieve one of two initial lung burdens (ILBs) of 239 PuO 2 alone or in combination with one of three ILBs of beryllium metal. Additional control groups of rats were sham exposed to air. Currently, approximately 58% of all rats planned for inclusion have been exposed. This report describes procedures used for the exposure, maintenance, and evaluation of rats in this study. Most of the animals are to be held for their life span in order to quantitate cancer incidence, with other animals assigned to serial sacrifice groups for quantitation of Pu and Be retention and determination of translocation patterns. Exposure to beryllium at any of the three doses tested retarded clearance of plutonium from the lung by a factor of approximately six. Acute inflammatory responses were studied in a separate group of rats exposed to Be. Except for rats receiving the highest ILB of beryllium metal, no differences between exposed and sham-exposed control groups have yet been noted in terms of mortality, weight changes, and clinical signs. (author)

  19. High-240Pu fuel worth in the Fast Test Reactor Engineering Mockup

    International Nuclear Information System (INIS)

    Daughtry, J.W.; Dobbin, K.D.

    1975-01-01

    Reactivity effects associated with the replacement of low- 240 Pu fuel with high- 240 Pu fuel were calculated and compared to measurements made in the FTR Engineering Mockup Critical (EMC). When the Pu and U isotopic compositions were changed in a way that increased the amounts of 240 Pu and 241 Pu and reduced the amounts of 239 Pu and 238 U while conserving total fissile mass and total fertile mass, the reactivity effect was positive. Calculation-to-experiment bias factors were obtained for this type of change and for the replacement of Fe 2 O 3 with U 3 O 8 in subassembly-size zones of the EMC. The k/sub e/--k/sub c/ bias decreased when high- 240 Pu fuel was introduced and increased when Fe 2 O 3 was replaced with U 3 O 8 . When the two changes were combined, their effects on the k/sub e/ --k/sub c/ bias tended to cancel out. The work described is related to plans for the utilization of light water reactor discharge Pu in the FTR

  20. Cooling rate dependence of simulated Cu{sub 64.5}Zr{sub 35.5} metallic glass structure

    Energy Technology Data Exchange (ETDEWEB)

    Ryltsev, R. E. [Institute of Metallurgy, Ural Branch of Russian Academy of Sciences, 101 Amundsen Str., 620016 Ekaterinburg (Russian Federation); Ural Federal University, 19 Mira Str., 620002 Ekaterinburg (Russian Federation); L.D. Landau Institute for Theoretical Physics, Russian Academy of Sciences, 2 Kosygina Str., 119334 Moscow (Russian Federation); Klumov, B. A. [L.D. Landau Institute for Theoretical Physics, Russian Academy of Sciences, 2 Kosygina Str., 119334 Moscow (Russian Federation); Aix-Marseille-Université, CNRS, Laboratoire PIIM, UMR 7345, 13397 Marseille Cedex 20 (France); High Temperature Institute, Russian Academy of Sciences, 13/2 Izhorskaya Str., 125412 Moscow (Russian Federation); Chtchelkatchev, N. M. [Institute of Metallurgy, Ural Branch of Russian Academy of Sciences, 101 Amundsen Str., 620016 Ekaterinburg (Russian Federation); L.D. Landau Institute for Theoretical Physics, Russian Academy of Sciences, 2 Kosygina Str., 119334 Moscow (Russian Federation); Moscow Institute of Physics and Technology, 9 Institutskiy Per., Dolgoprudny, 141700 Moscow Region (Russian Federation); All-Russia Research Institute of Automatics, 22 Sushchevskaya, 127055 Moscow (Russian Federation); Shunyaev, K. Yu. [Institute of Metallurgy, Ural Branch of Russian Academy of Sciences, 101 Amundsen Str., 620016 Ekaterinburg (Russian Federation); Ural Federal University, 19 Mira Str., 620002 Ekaterinburg (Russian Federation)

    2016-07-21

    Using molecular dynamics simulations with embedded atom model potential, we study structural evolution of Cu{sub 64.5}Zr{sub 35.5} alloy during the cooling in a wide range of cooling rates γ ∈ (1.5 ⋅ 10{sup 9}, 10{sup 13}) K/s. Investigating short- and medium-range orders, we show that the structure of Cu{sub 64.5}Zr{sub 35.5} metallic glass essentially depends on cooling rate. In particular, a decrease of the cooling rate leads to an increase of abundances of both the icosahedral-like clusters and Frank-Kasper Z16 polyhedra. The amounts of these clusters in the glassy state drastically increase at the γ{sub min} = 1.5 ⋅ 10{sup 9} K/s. Analysing the structure of the glass at γ{sub min}, we observe the formation of nano-sized crystalline grain of Cu{sub 2}Zr intermetallic compound with the structure of Cu{sub 2}Mg Laves phase. The structure of this compound is isomorphous with that for Cu{sub 5}Zr intermetallic compound. Both crystal lattices consist of two types of clusters: Cu-centered 13-atom icosahedral-like cluster and Zr-centered 17-atom Frank-Kasper polyhedron Z16. That suggests the same structural motifs for the metallic glass and intermetallic compounds of Cu–Zr system and explains the drastic increase of the abundances of these clusters observed at γ{sub min}.

  1. Radiative capture on $^{242}$Pu for MOX fuel reactors

    CERN Multimedia

    The use of MOX fuel (mixed-oxide fuel made of UO$_{2}$ and PuO$_{2}$) in nuclear reactors allows substituting a large fraction of the enriched Uranium by Plutonium reprocessed from spent fuel. Indeed around 66% of the plutonium from spent fuel is made of $^{239}$Pu and $^{241}$Pu, which are fissile in thermal reactors. A typical reactor of this type uses a fuel with 7% reprocessed Pu and 93% depleted U, thus profiting from both the spent fuel and the remaining $^{238}$U following the $^{235}$U enrichment. With the use of such new fuel compositions rich in Pu the better knowledge of the capture and fission cross sections of the Pu isotopes becomes very important. This is clearly stated in the recent OECD NEA’s “High Priority Request List” and in the WPEC-26 “Uncertainty and target accuracy assessment for innovative systems using recent covariance data evaluations” report. In particular, a new series of cross section evaluations have been recently carried out jointly by the European (JEFF) and United ...

  2. Uptake of Radionuclide Metals by SPME Fibers

    International Nuclear Information System (INIS)

    Duff, M; S Crump, S; Robert Ray, R; Keisha Martin, K; Donna Beals, D

    2006-08-01

    The Federal Bureau of Investigation (FBI) Laboratory currently does not have on site facilities for handling radioactive evidentiary materials and there are no established FBI methods or procedures for decontaminating high explosive (HE) and fire debris (FD) evidence while maintaining evidentiary value. One experimental method for the isolation of HE and FD residue involves using solid phase microextraction or SPME fibers to remove residue of interest. Due to their high affinity for organics, SPME fibers should have little affinity for most metals. However, no studies have measured the affinity of radionuclides for SPME fibers. The focus of this research was to examine the affinity of dissolved radionuclide ( 239/240 Pu, 238 U, 237 Np, 85 Sr, 133 Ba, 137 Cs, 60 Co and 226 Ra) and stable radionuclide surrogate metals (Sr, Co, Ir, Re, Ni, Ba, Cs, Nb, Zr, Ru, and Nd) for SPME fibers at the exposure conditions that favor the uptake of HE and FD residues. Our results from radiochemical and mass spectrometric analyses indicate these metals have little measurable affinity for these SPME fibers during conditions that are conducive to HE and FD residue uptake with subsequent analysis by liquid or gas phase chromatography with mass spectrometric detection

  3. Glass-Forming Ability and Early Crystallization Kinetics of Novel Cu-Zr-Al-Co Bulk Metallic Glasses

    Directory of Open Access Journals (Sweden)

    Xiaoliang Han

    2016-09-01

    Full Text Available In recent years, CuZr-based bulk metallic glass (BMG composites ductilized by a shape memory B2 CuZr phase have attracted great attention owing to their outstanding mechanical properties. However, the B2 CuZr phase for most CuZr-based glass-forming compositions is only stable at very high temperatures, leading to the uncontrollable formation of B2 crystals during quenching. In this work, by introducing Co (i.e., 4, 5, and 6 at. % and 10 at. % Al into CuZr-based alloys, the relatively good glass-forming ability (GFA of CuZr-based alloys still can be achieved. Meanwhile, the B2 phase can be successfully stabilized to lower temperatures than the final temperatures of crystallization upon heating CuZr-based BMGs. Unlike previous reported CuZr-based BMGs, the primary crystallization products upon heating are mainly B2 CuZr crystals but not CuZr2 and Cu10Zr7 crystals. Furthermore, the primary precipitates during solidification are still dominated by B2 crystals, whose percolation threshold is detected to lie between 10 ± 2 vol. % and 31 ± 2 vol. %. The crystallization kinetics underlying the precipitation of B2 crystals was also investigated. Our results show that the present glass-forming composites are promising candidates for the fabrication of ductile CuZr-based BMG composites.

  4. Determination of U and Pu in highly radioactive solutions. Application to the analysis of fuel reprocessing solutions

    International Nuclear Information System (INIS)

    Denis, Alphonse

    Various devices intended for the X-ray fluorescence analysis of U and Pu in radioactive solutions were developed in a hot cell: a device for sample deposition and preparation; a measurement unit. The problems investigated are described [fr

  5. Calculated activities of some isotopes in the RA reactor highly enriched fuel significant for possible environmental contamination - Operational report; Radni izvestaj - Proracun aktivnosti nekih izotopa u visokoobogacenom uranskom gorivu reaktora RA, znacajnih sa gledista moguce kontaminacije okoline

    Energy Technology Data Exchange (ETDEWEB)

    Bulovic, V; Martinc, R; Cupac, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-12-15

    This report contains calculation basis and obtained results of activities for three groups of isotopes in the RA reactor 80% enriched fuel element. The following isotopes are included: 1) {sup 85m}Kr, {sup 87}Kr, {sup 88}Kr, {sup 131}J, {sup 132}J, {sup 133}J, {sup 134}J, {sup 135}J, {sup 133}Xe, {sup 138}Xe i {sup 138}Cs, 2) {sup 89}Sr, {sup 90}Sr, {sup 91}Sr, {sup 92}Sr, {sup 95}Zr, {sup 97}Zr, {sup 103}Ru, {sup 105}Ru, {sup 106}Ru, {sup 129m}Te, {sup 134}Cs, {sup 137}Cs, {sup 140}Ba, {sup 144}Ce, kao i 3) {sup 238}Pu, {sup 239}Pu i {sup 240}Pu. It was estimated that the fuel is exposed to mean neutron flux. The periodicity of reactor operation is taken into account. Calculation results are given dependent on the time of exposure. These results are to be used as source data for Ra reactor safety analyses. [Serbo-Croat] Izlozene su osnove i prikazani su rezultati izvedenog proracuna aktivnosti tri grupe izotopa u gorivnom elementu reaktora RA sa 80% obogacenim uranom - 235. Obuhvaceni su: 1) {sup 85m}Kr, {sup 87}Kr, {sup 88}Kr, {sup 131}J, {sup 132}J, {sup 133}J, {sup 134}J, {sup 135}J, {sup 133}Xe, {sup 138}Xe i {sup 138}Cs, zatim, 2) {sup 89}Sr, {sup 90}Sr, {sup 91}Sr, {sup 92}Sr, {sup 95}Zr, {sup 97}Zr, {sup 103}Ru, {sup 105}Ru, {sup 106}Ru, {sup 129m}Te, {sup 134}Cs, {sup 137}Cs, {sup 140}Ba, {sup 144}Ce, kao i 3) {sup 238}Pu, {sup 239}Pu i {sup 240}Pu. Pretpostavljeno je da se gorivo ozracuje na srednjem fluksu neutrona, a periodicnost rada reaktora je uvazavana. Rezultati proracuna, dati u numerickom obliku, sistematizovani su kao funkcija toka vremena ozracivanja goriva. Ovi rezultati bice korisceni kao izvorni podaci kod izrade sigurnosnih analiza za reaktor RA (author)

  6. Preparation and characterization of 238Pu-ceramics for radiation damage experiments

    International Nuclear Information System (INIS)

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

    2000-01-01

    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 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 2 O 7 ], (2) pyrochlore-rich baseline, (3) pyrochlore-rich baseline with impurities, (4) phase-pure zirconolite [ideally Ca(U, Pu)Ti 2 O 7 ], and (5) a zirconolite-rich baseline. These ceramics were prepared with either normal weapons-grade Pu, which is predominantly 239 Pu, or 238 Pu. The 238 Pu accelerates the radiation damage relative to the 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 ∼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 238 Pu specimens become metamict and the damage saturates. They will

  7. Bidirectional threshold switching characteristics in Ag/ZrO{sub 2}/Pt electrochemical metallization cells

    Energy Technology Data Exchange (ETDEWEB)

    Du, Gang, E-mail: dugang@hdu.edu.cn; Li, Hongxia; Mao, Qinan; Ji, Zhenguo [College of Materials and Environmental Engineering, Hangzhou Dianzi University, Hangzhou 310018 (China); Wang, Chao [Key Laboratory of Nanodevices and Applications, Suzhou Institute of Nano-tech and Nano-bionics, Chinese Academy of Sciences, Ruoshui Road 398, Suzhou 215123 (China)

    2016-08-15

    A bidirectional threshold switching (TS) characteristic was demonstrated in Ag/ZrO{sub 2}/Pt electrochemical metallization cells by using the electrochemical active Ag electrode and appropriate programming operation strategies The volatile TS was stable and reproducible and the rectify ratio could be tuned to ∼10{sup 7} by engineering the compliance current. We infer that the volatile behavior is essentially due to the moisture absorption in the electron beam evaporated films, which remarkably improved the anodic oxidation as well as the migration of Ag{sup +} ions. The resultant electromotive force would act as a driving force for the metal filaments dissolution, leading to the spontaneous volatile characteristics. Moreover, conductance quantization behaviors were also achieved owing to formation and annihilation of atomic scale metal filaments in the film matrix. Our results illustrate that the Ag/ZrO{sub 2}/Pt device with superior TS performances is a promising candidate for selector applications in passive crossbar arrays.

  8. SEM in situ MiniCantilever Beam Bending of U-10Mo/Zr/Al Fuel Elements

    Energy Technology Data Exchange (ETDEWEB)

    Mook, William [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Baldwin, Jon K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez, Ricardo M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mara, Nathan A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-06-16

    In this work, the fracture behavior of Al/Zr and Zr/dU-10Mo interfaces was measured via the minicantilever bend technique. The energy dissipation rates were found to be approximately 3.7-5 mj/mm2 and 5.9 mj/mm2 for each interface, respectively. It was found that in order to test the Zr/U-10Mo interface, location of the hinge of the cantilever was a key parameter. While this test could be adapted to hot cell use through careful alignment fixturing and measurement of crack lengths with an optical microscope (as opposed to SEM, which was used here out of convenience), machining of the cantilevers via MiniMill in such a way as to locate the interfaces at the cantilever hinge, as well as proper placement of a femtosecond laser notch will continue to be key challenges in a hot cell environment.

  9. Study of the variation with the energy of the fission cross-sections of {sup 233}U, {sup 235}U, {sup 239}Pu for the fast neutrons; Etude de la variation avec l'energie des sections efficaces de fission de {sup 233}U, {sup 235}U, {sup 239}Pu pour les neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Szteinsznaider, D; Naggiar, V; Netter, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    This measurements have been done while taking the value of the fission cross-sections of {sup 238}U as reference. The neutrons are produced by the reaction {sup 7}Li(p,n) in the Van de Graaff generator of Saclay. The explored domain spreads from some tenths to 2000 keV. We find: for {sup 239}Pu: {sigma}{sub f} = 2,04 {+-} 0,12 barns, cross-section constant between 150 and 2000 keV, for {sup 235}U: {sigma}{sub f} = 1,15 {+-} 0,15 barns, cross-section constant between 700 and 1000 keV, for {sup 233}U: {sigma}{sub f} = 1,92 {+-} 0,25 barns, for neutrons of 850 keV. (authors) [French] Ces mesures ont ete effectuees en prenant la valeur de la section efficace de fission de {sup 238}U comme reference. Les neutrons sont produits par la reaction {sup 7}Li(p,n) au generateur Van de Graaff de Saclay. Le domaine explore s'etend de quelques dizaines de kev a 2000 kev. Nous trouvons: pour {sup 239}Pu: {sigma}{sub f} = 2,04 {+-} 0,12 barns, section efficace constante entre 150 et 2000 kev. pour {sup 235}U: {sigma}{sub f} = 1,15 {+-} 0,15 barns, section efficace constante entre 700 et 1000 kev. pour {sup 233}U: {sigma}{sub f} = 1,92 {+-} 0,25 barns, pour des neutrons de 850 kev. (auteurs)

  10. Improvement of bio-corrosion resistance for Ti42Zr40Si15Ta3 metallic glasses in simulated body fluid by annealing within supercooled liquid region.

    Science.gov (United States)

    Huang, C H; Lai, J J; Wei, T Y; Chen, Y H; Wang, X; Kuan, S Y; Huang, J C

    2015-01-01

    The effects of the nanocrystalline phases on the bio-corrosion behavior of highly bio-friendly Ti42Zr40Si15Ta3 metallic glasses in simulated body fluid were investigated, and the findings are compared with our previous observations from the Zr53Cu30Ni9Al8 metallic glasses. The Ti42Zr40Si15Ta3 metallic glasses were annealed at temperatures above the glass transition temperature, Tg, with different time periods to result in different degrees of α-Ti nano-phases in the amorphous matrix. The nanocrystallized Ti42Zr40Si15Ta3 metallic glasses containing corrosion resistant α-Ti phases exhibited more promising bio-corrosion resistance, due to the superior pitting resistance. This is distinctly different from the previous case of the Zr53Cu30Ni9Al8 metallic glasses with the reactive Zr2Cu phases inducing serious galvanic corrosion and lower bio-corrosion resistance. Thus, whether the fully amorphous or partially crystallized metallic glass would exhibit better bio-corrosion resistance, the answer would depend on the crystallized phase nature. Copyright © 2015 Elsevier B.V. All rights reserved.

  11. Derivation of decay heat benchmarks for U235 and Pu239 by a least squares fit to measured data

    International Nuclear Information System (INIS)

    Tobias, A.

    1989-05-01

    A least squares technique used by previous authors has been applied to an extended set of available decay heat measurements for both U235 and Pu239 to yield simultaneous fits to the corresponding beta, gamma and total decay heat. The analysis takes account of both systematic and statistical uncertainties, including correlations, via calculations which use covariance matrices constructed for the measured data. The results of the analysis are given in the form of beta, gamma and total decay heat estimates following fission pulses and a range of irradiation times in both U235 and Pu239. These decay heat estimates are considered to form a consistent set of benchmarks for use in the assessment of summation calculations. (author)

  12. Separation of Np from U and Pu using a salt-free reductant for Np(VI) by continuous counter-current back-extraction

    International Nuclear Information System (INIS)

    Ban, Yasutoshi; Asakura, Toshihide; Morita, Yasuji

    2005-01-01

    Reduction properties of several salt-free reagents for Np(VI) and Pu(IV) were reviewed to choose selective reductants that reduce only Np(VI) to Np(V) for separating Np from U and Pu in TBP by reductive back-extraction. Allylhydrazine was proposed as a candidate for selective Np(VI) reductant, and it was confirmed by a batch experiment that allylhydrazine reduced almost all Np(VI) to Np(V) and back-extracted Np from organic phase (30 vol.% TBP diluted in n-dodecane) to aqueous phase (3 mol/dm 3 HNO 3 ) within 10 min. A continuous counter-current experiment using a miniature mixer-settler was carried out with allylhydrazine at room temperature. At least 91% of Np(VI) that fed to the mixer-settler was selectively reduced to Np(V) and separated from U and Pu. (author)

  13. Superhydrophobic Zr-based metallic glass surface with high adhesive force

    Science.gov (United States)

    Li, Ning; Xia, Ting; Heng, Liping; Liu, Lin

    2013-06-01

    Micro/nano hierarchical structures were constructed on Zr35Ti30Be26.75Cu8.25 metallic glass surface by silicon moulding and subsequently chemical etching. The as-formed surface exhibited both superhydrophobicity and high adhesive force towards water. The superhydrophobicity is rationalized based on the modified Cassie-Baxter model [A. B. D. Cassie and S. Baxter, Trans. Faraday Soc. 40, 546 (1944)]. The origin of the robust adhesion is described in terms of intermolecular capillary forces. The present results not only provide a method to fabricate superhydrophobic metallic glasses surface but also explore an important industrial application as dry adhesives and transport of liquid microdroplets.

  14. The radiolysis of solutions containing Pu(6)

    International Nuclear Information System (INIS)

    Rance, P.J.W.; Zilberman, B.Y.

    2000-01-01

    The reduction of Pu(VI) in nitric acid solutions containing uranium and various fission product elements as a result of both its inherent alpha radiation and also external gamma irradiation at dose rates similar to those experienced by dissolved fuel solutions has been investigated. The presence of the additional metals has been shown to eliminate the induction periods required prior to the reduction of Pu(VI) in nitric acid. G values for the auto-radiolytic reduction of Pu(VI) have been found to be between 0.6 and 1.1 for 3 g/1 Pu solutions containing between 0.12 and 9.2 % 238 Pu (balance 239 Pu). Uranium and palladium have been found to accelerate the reduction of Pu(VI) during gamma irradiation at dose rates of between 0.41 and 1.64 kGy/hour. (authors)

  15. The radiolysis of solutions containing Pu(6)

    Energy Technology Data Exchange (ETDEWEB)

    Rance, P.J.W. [BNFL British Nuclear Fuels, Sellafield, Seascale, Cumbria, Research and Technology (United Kingdom); Zilberman, B.Y. [V.G. Khlopin Radium Institute, St. Petersburg (Russian Federation)

    2000-07-01

    The reduction of Pu(VI) in nitric acid solutions containing uranium and various fission product elements as a result of both its inherent alpha radiation and also external gamma irradiation at dose rates similar to those experienced by dissolved fuel solutions has been investigated. The presence of the additional metals has been shown to eliminate the induction periods required prior to the reduction of Pu(VI) in nitric acid. G values for the auto-radiolytic reduction of Pu(VI) have been found to be between 0.6 and 1.1 for 3 g/1 Pu solutions containing between 0.12 and 9.2 % {sup 238}Pu (balance {sup 239}Pu). Uranium and palladium have been found to accelerate the reduction of Pu(VI) during gamma irradiation at dose rates of between 0.41 and 1.64 kGy/hour. (authors)

  16. Considerations on the DEMO pellet fuelling system

    Energy Technology Data Exchange (ETDEWEB)

    Lang, P.T., E-mail: peter.lang@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Day, Ch. [Karlsruhe Institute of Technology, 76021 Karlsruhe (Germany); Fable, E. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Igitkhanov, Y. [Karlsruhe Institute of Technology, 76021 Karlsruhe (Germany); Köchl, F. [Association EURATOM-Ö AW/ATI, Atominstitut, TU Wien, 1020 Vienna (Austria); Mooney, R. [Culham Centre for Fusion Energy, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Pegourie, B. [CEA, IRFM, 13108 Saint-Paul-lez-Durance (France); Ploeckl, B. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Wenninger, R. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); EFDA, Garching (Germany); Zohm, H. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-10-15

    Graphical abstract: - Highlights: • Considerations are made for a core particle fuelling system covering all DEMO requirements. • Particle deposition beyond the pedestal top is needed to achieve efficient fuelling. • Conventional pellet technology enabling launching from the torus inboard side can be used. • Efforts have been taken for integrating a suitable pellet guiding system into the EU DEMO model. • In addition, further techniques bearing potential for advanced fuelling performance are considered. - Abstract: The Demonstration Fusion Power Reactor DEMO is the step foreseen to bridge the gap between ITER and the first commercial fusion power plant. One key element in the European work plan for DEMO is the elaboration of a conceptual design for a suitable core particle fuelling system. First considerations for such a system are presented in this contribution. Following the well-considered ITER solution, most analysis performed in this study assumes conventional pellet technology will be used for the fuelling system. However, taking advantage of the less compressed time frame for the DEMO project, several other techniques thought to bear potential for advanced fuelling performance are considered as well. In a first, basic analysis all actuation parameters at hand and their implications on the fuelling performance were considered. Tentative transport modeling of a reference scenario strongly indicates only particles deposited inside the plasma pedestal allow for efficient fuelling. Shallow edge fuelling results in an unbearable burden on the fuel cycle. Sufficiently deep particle deposition seems technically achievable, provided pellets are launched from the torus inboard at sufficient speed. All components required for a DEMO pellet system capable for high speed inboard pellet launch are already available or can be developed in due time with reasonable efforts. Furthermore, steps to integrate this solution into the EU DEMO model are taken.

  17. Considerations on the DEMO pellet fuelling system

    International Nuclear Information System (INIS)

    Lang, P.T.; Day, Ch.; Fable, E.; Igitkhanov, Y.; Köchl, F.; Mooney, R.; Pegourie, B.; Ploeckl, B.; Wenninger, R.; Zohm, H.

    2015-01-01

    Graphical abstract: - Highlights: • Considerations are made for a core particle fuelling system covering all DEMO requirements. • Particle deposition beyond the pedestal top is needed to achieve efficient fuelling. • Conventional pellet technology enabling launching from the torus inboard side can be used. • Efforts have been taken for integrating a suitable pellet guiding system into the EU DEMO model. • In addition, further techniques bearing potential for advanced fuelling performance are considered. - Abstract: The Demonstration Fusion Power Reactor DEMO is the step foreseen to bridge the gap between ITER and the first commercial fusion power plant. One key element in the European work plan for DEMO is the elaboration of a conceptual design for a suitable core particle fuelling system. First considerations for such a system are presented in this contribution. Following the well-considered ITER solution, most analysis performed in this study assumes conventional pellet technology will be used for the fuelling system. However, taking advantage of the less compressed time frame for the DEMO project, several other techniques thought to bear potential for advanced fuelling performance are considered as well. In a first, basic analysis all actuation parameters at hand and their implications on the fuelling performance were considered. Tentative transport modeling of a reference scenario strongly indicates only particles deposited inside the plasma pedestal allow for efficient fuelling. Shallow edge fuelling results in an unbearable burden on the fuel cycle. Sufficiently deep particle deposition seems technically achievable, provided pellets are launched from the torus inboard at sufficient speed. All components required for a DEMO pellet system capable for high speed inboard pellet launch are already available or can be developed in due time with reasonable efforts. Furthermore, steps to integrate this solution into the EU DEMO model are taken.

  18. A modified Gaussian integration method for thermal reaction rate calculation in U- and Pu-isotopes

    International Nuclear Information System (INIS)

    Bosevski, T.; Fredin, B.

    1966-01-01

    An advanced multi-group cell calculations a lot of data information is very often necessary, and hence the data administration will be elaborate, and the spectrum calculation will be time consuming. We think it is possible to reduce the necessary data information by using an effective reaction rate integration method well suited for U- and Pu-absorptions (author)

  19. Lattice thermal expansions of NpN, PuN and AmN

    International Nuclear Information System (INIS)

    Takano, Masahide; Akabori, Mitsuo; Arai, Yasuo; Minato, Kazuo

    2008-01-01

    Lattice parameters of NpN, PuN and AmN were measured by a high temperature X-ray diffraction method from room temperature up to 1478 K. Linear thermal expansions of these TRU nitrides were determined as a function of temperature. The average coefficients of linear thermal expansion from 293 to 1273 K were 8.8, 11.1 and 11.2 x 10 -6 K -1 for NpN, PuN and AmN, respectively. The instantaneous coefficient of thermal expansion either at 293 or at 1273 K against the reciprocal decomposition temperature under 1 atm of nitrogen showed a linear relationship for TiN, ZrN, HfN, UN, NpN and PuN. Based on this relationship, the decomposition temperature of AmN was roughly predicted to be 2700 K

  20. Fuel design for the U.S. accelerator driven transmutation system

    International Nuclear Information System (INIS)

    Meyer, M. K.; Hayes, S. L.; Crawford, D. C.; Pahl, R. G.; Tsai, H.

    2002-01-01

    The U.S. concept for actinide transmutation is currently envisioned as a system to destroy plutonium as well as minor actinides in a single or two tier system. In order to maximize the actinide destruction rate, an inert matrix fuel is used. The effectiveness of transmutation in reducing the actinide inventory is linked to the development of a robust fuel system, capable of achieving very high burnup. Very little fuel performance data has been generated to date on inert matrix systems, and there are several issues specific to the behavior of higher actinides that do not allow extension of the existing uranium-plutonium fuel database to these new fuels. These issues include helium production, fuel-cladding-chemical-interaction, and americium migration. In the early 1990's, two U-Pu-Zr metal alloy fuel elements containing 1.2 wt.% Am and 1.3 wt.% Np were fabricated and irradiated to approximately 6 at.% burnup in the Experimental Breeder Reactor-II. Postirradiation examination results were not published; however the recent interest in fuel for actinide transmutation has prompted a reexamination of this data. The results of the postirradiation examination of this experiment, including gas sampling, metallography, and gamma scanning are discussed. Available data on inert matrix fuels and other fuels incorporating actinides are used to assess the implications of minor-actinide specific issues on transmuter fuel. Considerations for the design of nitride and oxide fuels, metallic fuels, and metal-matrix dispersion fuels are discussed

  1. Fuel Design for the U.S. Accelerator Driven Transmutation System

    International Nuclear Information System (INIS)

    Meyer, M.K.; Hayes, S.L.; Crawford, D.C.; Pahl, R.G.; Tsai, H.

    2002-01-01

    The U.S. concept for actinide transmutation is currently envisioned as a system to destroy plutonium as well as minor actinides in a single or two tier system. In order to maximize the actinide destruction rate, an inert matrix fuel is used. The effectiveness of transmutation in reducing the actinide inventory is linked to the development of a robust fuel system, capable of achieving very high burnup. Very little fuel performance data has been generated to date on inert matrix systems, and there are several issues specific to the behavior of higher actinides that do not allow extension of the existing uranium-plutonium fuel database to these new fuels. These issues include helium production, fuel-cladding-chemical-interaction, and americium migration. In the early 1990's, two U-Pu-Zr metal alloy fuel elements containing 1.2 wt.% Am and 1.3 wt.% Np were fabricated and irradiated to approximately 6 at.% burnup in the Experimental Breeder Reactor-II. Postirradiation examination results were not published; however the recent interest in fuel for actinide transmutation has prompted a reexamination of this data. The results of the postirradiation examination of this experiment, including gas sampling, metallography, and gamma scanning are discussed. Available data on inert matrix fuels and other fuels incorporating actinides are used to assess the implications of minor-actinide specific issues on transmuter fuel. Considerations for the design of nitride and oxide fuels, metallic fuels, and metal-matrix dispersion fuels are discussed. (authors)

  2. Determination of carbon content of UO2, (U, Gd)O2 and (U, Pu)O2 powders and sintered pellets - Combustion in a high-frequency induction furnace -Infrared absorption spectrometry

    International Nuclear Information System (INIS)

    2008-01-01

    This International Standard describes a method for determining the carbon content in UO 2 , (U,Gd)O 2 and (U,Pu)O 2 powder and sintered pellets by combustion in an induction furnace and infrared absorption spectroscopy measurement. It is applicable for determining 10 μg/g to 500 μg/g of carbon in UO 2 , (U,Gd)O 2 and (U,Pu)O 2 powder and pellets. The sample is heated to a temperature above 1500 deg. C in an induction furnace, under pure oxygen atmosphere, to convert any carbon compounds to carbon dioxide gas. The resulting carbon dioxide gas is filtered and dried before measurement using infrared spectroscopy to measure the carbon dioxide signal at 2350 cm -1 . The result is converted into the carbon content of the material analysed

  3. Welding of Zr-based bulk metallic glasses

    International Nuclear Information System (INIS)

    Elahi, M.

    2010-01-01

    Recently, many bulk metallic glass (BMG) materials with high specific strength, hardness and superior corrosion resistance have been developed and the maximum thickness of some Zr-based BMGs have reached several tenths of millimeters. Nevertheless, homogeneous glassy BMGs are not thick enough to be used for structural applications. In order to extend the engineering applications of BMG materials, BMG welding technologies needed to be developed. Specifically, the welding technologies of dissimilar materials such as BMG materials to crystalline alloys are to be developed. The functional use of the specific properties of each material in dissimilar material combination provides flexible design possibilities for products. In this project electron beam welding is employed to join BMG with BMG of different composition as well as with different crystalline materials (i.e. Hastealoy C-276, Inconel-625 and pure Ti metal). Defects free weld joint was achieved in BMG-BMG welding. Some cracks were produced in melt zone of BMG-Ti and BMG-Hastealoy C-276 welding while at joint they fuse properly with BMG. Inconel-625 could not properly weld with BMG. In all cases, hardness of melt zone was found to be higher than the base metals and the heat affected zone (HAZ). (author)

  4. Oxidation kinetics of simulated metallic spent fuel in air at 200∼300 .deg. C

    International Nuclear Information System (INIS)

    Joo, J. S.; Yoo, K. S.; Jo, I. J.; Kook, D. H.; Lee, E. P.; Lee, J. C.; Bang, K. S.; Kim, H. D.

    2003-01-01

    In order to evaluate the long term storage safety study of the metallic spent fuel, U-5Zr, U-5Ti, U-5Ni, U-5Nb, and U-5Hf simulated metallic uranium alloys, known as corrosion resistant alloys, were fabricated and oxidized in oxygen gas at 200 .deg. C ∼ 300 .deg. C. All simulated metallic uranium alloys were more corrosion resistant than pure uranium metal, and corrosion resistance increases Nb, Ni, Ti, Zr, Hf in that order. The oxidation rates of uranium alloys determined and activation energy was calculated for each alloy. The matrix microstructure of the test specimens were analyzed using OM, SEM, and EPMA. It was concluded that Nb was the best acceptable alloying elements for reducing corrosion of uranium metal, and Ni, Ti were also considered to suitable as candidate

  5. Roles of texture of Zr alloys in ZrO{sub 2} film formation and δ-hydride orientation near ZrO{sub 2}/Zr interface

    Energy Technology Data Exchange (ETDEWEB)

    Qin, W.; Szpunar, J.A., E-mail: weq565@mail.usask.ca, E-mail: jerzy.szpunar@usask.ca [Univ. of Saskatchewan, Dept. of Mechanical Engineering, Saskatoon, SK (Canada); Kozinski, J., E-mail: janusz.kozinski@lassonde.yorku.ca [York Univ., Faculty of Science and Engineering, Toronto, ON (Canada)

    2014-07-01

    Oxidation and hydrogen embrittlement are related to formation of cracks and failure of Zr alloys used in nuclear reactor applications. An in-depth understanding of the formation of ZrO{sub 2} film and the hydride precipitation and orientation is important for improving the corrosion resistance of zirconium alloys. In this work a theoretical model is developed to analyze the microstructure of ZrO{sub 2} film formed on Zr alloys and the effect of stress that results from ZrO{sub 2} formation on hydride reorientation in the region near oxide/metal interface. Our work shows that the macroscopic stress produced due to Pilling-Bedworth ratio for ZrO{sub 2}/Zr could lead to the hydride re-orientation in the region near ZrO{sub 2}/Zr interface. Whether or not this effect can occur is dependent on the texture of the zirconium alloys. Control of texture of zirconium alloys can affect the microstructure of ZrO{sub 2} film and can be responsible for change of hydride orientation. (author)

  6. Application of Ceramic Bond Coating for Reusable Melting Crucible of Metallic Fuel Slugs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki-Hwan; Song, Hoon; Ko, Young-Mo; Park, Jeong-Yong; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Hong, Ki-Won [Chungnam National University, Daejeon (Korea, Republic of)

    2016-10-15

    Metal fuel slugs of the driver fuel assembly have been fabricated by injection casting of the fuel alloys under a vacuum state or an inert atmosphere. Traditionally, metal fuel such as a U-Zr alloy system for SFR has been melted in slurry-coated graphite crucibles and cast in slurry-coated quartz tube molds to prevent melt/material interactions. Reactive coatings and porous coatings can be a source of melt contaminations, and fuel losses, respectively. Ceramic Y{sub 2}O{sub 3}, TiC, and TaC coating materials showed no penetration in the protective layer after a melt dipping test. However, the ceramic coating materials showed separations in the coating interface between the substrate and coating layer, or between the coating layer and fuel melt after the dipping test. All plasma-spray coated methods maintained a sound coating state after a dipping test with U-10wt.%Zr melt. A single coating Y{sub 2}O{sub 3}(150) layer and double coating layer of TaC(50)-Y{sub 2}O{sub 3}(100), showed a sound state or little penetration in the protective layer after a dipping test with U-10wt.%Zr-5wt.%RE melt. Injection casting experiments of U-10wt.%Zr and U-10wt.%Zr-5wt.%RE fuel slugs have been performed to investigate the feasibility of a reusable crucible of the metal fuel slugs. U–10wt.%Zr and U–10wt.%Zr–5wt.%RE fuel slugs have been soundly fabricated without significant interactions of the graphite crucibles. Thus, the ceramic plasma-spray coatings are thought to be promising candidate coating methods for a reusable graphite crucible to fabricate metal fuel slugs.

  7. Amorphous Ti-Zr

    International Nuclear Information System (INIS)

    Rabinkin, A.; Liebermann, H.; Pounds, S.; Taylor, T.

    1991-01-01

    This paper is the first report on processing, properties and potential application of amorphous titanium/zirconium-base alloys produced in the form of a good quality continuous and ductile ribbon having up to 12.5 mm width. To date, the majority of titanium brazing is accomplished using cooper and aluminum-base brazing filler metals. The brazements produced with these filler metals have rather low (∼300 degrees C) service temperature, thus impeding progress in aircraft and other technologies and industries. The attempt to develop a generation of high temperature brazing filler metals was made in the late sixties-early seventies studies in detail were a large number of Ti-, Zr-Ti-Zr, Ti-V and Zr-V-Ti based alloys. The majority of these alloys has copper and nickel as melting temperature depressants. The presence of nickel and copper converts them into eutectic alloys having [Ti(Zr)] [Cu(Ni)], intermetallic phases as major structural constituents. This, in turn, results in high alloy brittleness and poor, if any, processability by means of conventional, i.e. melting-ingot casting-deformation technology. In spite of good wettability and high joint strength achieved in dozens of promising alloys, only Ti-15Cu-15Ni is now widely used as a brazing filler metal for high service temperature. Up until now this material could not be produced as a homogeneous foil and is instead applied as a clad strip consisting of three separate metallic layers

  8. The case for wood-fuelled heating

    International Nuclear Information System (INIS)

    Bent, Ewan

    2001-01-01

    This article looks at the wood heating industry in the UK and examines the heat market and the growth potential in the domestic, public, agricultural and commercial sectors. The current status of wood-fueled heating technology is considered, along with log and chip boilers, and the use of pellet fuel. The economics of wood-fuelled heating, the higher level of utilisation of wood-fuelled heating by utilities in northern European countries compared with the UK, and the barriers to the exploitation of wood fuelled heating are examined

  9. FCRD Advanced Reactor (Transmutation) Fuels Handbook

    Energy Technology Data Exchange (ETDEWEB)

    Janney, Dawn Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Papesch, Cynthia Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

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

  10. A contribution to the study of the mixed uranium-plutonium mono-carbides containing small quantities of zirconium

    International Nuclear Information System (INIS)

    Bocker, S.

    1970-03-01

    We have studied a mixed monocarbide, type (U,Pu)C, containing small additions of zirconium for the application as a fast neutron reactor fuel. A preliminary study was conducted on the (U,Zr)C monocarbide (Report CEA-R-3765(1). It was found that small additions of zirconium to the uranium-plutonium monocarbide improve a number of properties such as atmospheric corrosion, the hardness, and particularly the compatibility with 316 stainless steel. However, properties such as the coefficient of expansion and the melting point are only slightly changed. The relative percentage of Pu/U+Pu in the monocarbide was fixed at 20 per cent. Two processes of fabrication were employed: casting in an arc furnace, sintering, carried out after having the hydrides of the metals carburized. The metallurgical results indicate, that the above mentioned fuel might be of interest for fast neutron reactor application. (author) [fr

  11. Extraction chromatography of U(VI) and Pu(IV) adsorbed on amberlite XAD-7/dibutyloctanamide

    International Nuclear Information System (INIS)

    Prabhu, D.R.; Mahajan, G.R.; Nair, G.M.; Subramanian, M.S.

    1992-01-01

    The adsorption of U(VI) and Pu(IV) into the neutral poly acrylic resin Amberlite XAD-7, impregnated with dibutyloctanamide was found to be maximum at around 6M HNO 3 . Both these ions were found to be adsorbed as their monosolvates. The thermodynamic parameters obtained from the data at different temperatures indicated that the adsorption reaction was enthalpy favoured and entropy disfavored. (author). 5 refs., 1 tab

  12. Establishment of Experimental Apparatus and Mechanical Test for SFR Metallic Fuel

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Lee, Chong Tak; Oh, Seok Jin; Ko, Young Mo; Kim, Ki Hwan; Woo, Yoon Myung; Lee, Chan Bock

    2010-12-01

    U-Zr binary alloys and U-Zr-Ce ternary alloys as SFR surrogate metallic fuels were fabricated by a casting process. Tensile tests were performed to evaluate the mechanical properties of the fuels. As a results, the mechanical properties such as yield strength, ultimate tensile strength, and elongation were measured. In this report, these experimental results are presented

  13. Surrogate measurement of the 238Pu(n,f) cross section

    International Nuclear Information System (INIS)

    Ressler, J. J.; Burke, J. T.; Escher, J. E.; Bernstein, L. A.; Bleuel, D. L.; Casperson, R. J.; Gostic, J.; Henderson, R.; Scielzo, N. D.; Thompson, I. J.; Wiedeking, M.; Angell, C. T.; Goldblum, B. L.; Munson, J.; Basunia, M. S.; Phair, L. W.; Beausang, C. W.; Hughes, R. O.; Hatarik, R.; Ross, T. J.

    2011-01-01

    The neutron-induced fission cross section of 238 Pu was determined using the surrogate ratio method. The (n,f) cross section over an equivalent neutron energy range 5-20 MeV was deduced from inelastic α-induced fission reactions on 239 Pu, with 235 U(α,α ' f) and 236 U(α,α ' f) used as references. These reference reactions reflect 234 U(n,f) and 235 U(n,f) yields, respectively. The deduced 238 Pu(n,f) cross section agrees well with standard data libraries up to ∼10 MeV, although larger values are seen at higher energies. The difference at higher energies is less than 20%.

  14. Use of combined alpha-spectrometry and fission track analysis for the determination of 240Pu/239Pu ratios in human tissue

    International Nuclear Information System (INIS)

    Love, S.F.; Filby, R.H.; Glover, S.E.; Stuit, D.B.; Kathren, R.L.

    1998-01-01

    Plutonium and other actinides were determined in human autopsy tissues of occupationally exposed workers who were registrants of the United States Transuranium and Uranium Registries (USTUR). In this study, Pu was purified and isolated from Am, U and Th, after drying and wet-ashing of the tissues, and the addition of 238 Pu as a radiotracer. After electrodeposition onto vanadium planchets, the 239+240 Pu activity was determined by alpha-spectrometry. A fission track method was developed to determine 239 Pu in the presence of 238 Pu and 240 Pu, using Lexan TM polycarbonate detectors. Combining the two techniques allowed the determination of the 240 Pu/ 239 Pu activity and atom ratios. Data from selected USTUR cases are presented. (author)

  15. Apparatus for surface treatment of U-Pu carbide fuel samples

    International Nuclear Information System (INIS)

    Fukushima, Susumu; Arai, Yasuo; Handa, Muneo; Ohmichi, Toshihiko; Shiozawa, Ken-ichi.

    1979-05-01

    Apparatus has been constructed for treating the surface of U-Pu carbide fuel samples for EPMA. The treatment is to clean off oxide layer on the surface, then coat with an electric-conductive material. The apparatus, safe in handling plutonium, operates as follows. (1) To avoid oxidation of the analyzing surface by oxygen and water in the air, series of cleaning and coating, i.e. ion-etching and ion-coating or ion-etching and vacuum-evaporation is done at the same time in an inert gas atmosphere. (2) Ion-etching is possible on samples embedded in non-electric-conductive and low heat-conductive resin. (3) Since the temperature rise in (2) is negligible, there is no deterioration of the samples. (author)

  16. Deformation-strengthening during rolling Cu60Zr20Ti20 bulk metallic glass

    DEFF Research Database (Denmark)

    Cao, Q.P.; Li, J.F.; Hu, Yuyan

    2007-01-01

    Mechanical strength evolutions during rolling the Cu60Zr20Ti20 bulk metallic glass (BMG) at room temperature (RT) and cryogenic temperature (CT) have been investigated by measuring the microhardness. The hardness slightly increases during the initial rolling stage as a result of the gradually...

  17. Measurements Conducted on an Unknown Object Labeled Pu-239

    International Nuclear Information System (INIS)

    Hoteling, Nathan

    2013-01-01

    Measurements were carried out on 12 November 2013 to determine whether Pu-239 was present on an object discovered in a plastic bag with label ''Pu-239 6 uCi''. Following initial survey measurements to verify that the object was not leaking or contaminated, spectra were collected with a High Purity Germanium (HPGe) detector with object positioned in two different configurations. Analysis of the spectra did not yield any direct evidence of Pu-239. From the measured spectra, minimum detectable activity (MDA) was determined to be approximately 2 uCi for the gamma ray measurements. Although there was no direct evidence of Pu-239, a peak at 60 keV characteristic of Am-241 decay was observed. Since it is very likely that Am-241 would be present in aged plutonium samples, this was interpreted as indirect evidence for the presence of plutonium on the object. Analysis of this peak led to an estimated Pu-239 activity of 0.02-0.04 uCi, or <1x10 -6 grams.

  18. Study of uranium - 20 Wt per cent plutonium-niobium alloys (1963); Etude d'alliages U-Pu-Nb a 20 pour cent en poids de plutonium (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Abgrall, J; Barthelemy, P; Boucher, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1963-07-01

    U-Pu-Nb alloys containing 20 wt per cent Pu and 10 - 20 - 30 - 40 - 50 or 60 wt per cent Nb have been studied principally to determine the feasibility of their use as fuel element. The fabrication, casting and homogenisation presented certain difficulties due specially to niobium. The transformation temperatures, thermal expansion coefficients and nature of phases have been determined by thermal analysis, dilatometry, micrography and X Rays diffraction. For similar compositions, U-Pu-Mo and U-Pu-Nb alloys have many common points concerning the presence of zeta phase (up to 40 wt per cent Nb), the coefficients of expansion, the good behaviour during thermal cycling and the good resistance to air oxidation in spite of zeta phase. In consequence, irradiation tests in EL{sub 3} reactor (Saclay) will be carried out in the near future. (authors) [French] Les alliages a 20 pour cent de plutonium, 10 - 20 - 30 - 40 - 50 - 60 pour cent de niobium et le complement en uranium ont ete etudies du point de vue de leur possibilite d'emploi comme combustible. Les problemes d'elaboration, de mise en forme et d'homogeneisation sont presentes. Ils sont relativement delicats. On a determine par analyse thermique, dilatometrie, micrographie et diffraction des rayons X les temperatures de transformation a l'etat solide, les coefficients de dilatation et la nature des phases. Pour des teneurs analogues, on retrouve de nombreux points communs avec les alliages U-Pu-Mo: presence de la phase zeta des U-Pu a temperature moyenne, coefficients de dilatation analogues, bonne tenue en cyclage thermique et bonne resistance a l'oxydation dans l'air malgre la presence de la phase zeta. Des essais d'irradiation dans EL{sub 3} vont etre entrepris. (auteurs)

  19. Study of the Effect of (U0.8Pu0.2O2 Uranium–Plutonium Mixed Fuel Fission Products on a Living Organism

    Directory of Open Access Journals (Sweden)

    Ayagoz Baimukhanova

    2016-08-01

    Full Text Available The article describes the results of experiments conducted on pigs to determine the effect of plutonium, which is the most radiotoxic and highly active element in the range of mixed fuel (U0.8Pu0.2O2 fission products, on living organisms. The results will allow empirical prediction of the emergency plutonium radiation dose for various organs and tissues of humans in case of an accident in a reactor running on mixed fuel (U0.8Pu0.2O2.

  20. Elastic-constant systematics in f.c.c. metals, including lanthanides-actinides

    Energy Technology Data Exchange (ETDEWEB)

    Ledbetter, Hassel [Mechanical Engineering Department, University of Colorado, Boulder, Colorado 80309 (United States); Migliori, Albert [Los Alamos National Laboratory (E536), Los Alamos, New Mexico 87545 (United States)

    2008-01-15

    For f.c.c. metals, using Blackman's diagram of dimensionless elastic-constant ratios, we consider the systematics of physical properties and interatomic bonding. We focus especially on the lanthanides-actinides La, Ce, Yb, Th, U, Pu, those for which we know some monocrystal elastic constants. Their behavior differs from the other f.c.c. metals, and all except La show a negative Cauchy pressure, contrary to most f.c.c. metals, which show a positive Cauchy pressure. Among the lanthanides-actinides, {delta}-Pu stands apart, consistent with its many odd physical properties. Based on elastic-constant correlations, we suggest that {delta}-Pu possesses a strong s-electron interatomic-bonding component together with a covalent component. Elastically, {delta}-Pu shows properties similar to Yb. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)