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Sample records for metallic fuel lmfbr

  1. Benchmark physics experiment of metallic-fueled LMFBR at FCA. 2

    International Nuclear Information System (INIS)

    Iijima, Susumu; Oigawa, Hiroyuki; Ohno, Akio; Sakurai, Takeshi; Nemoto, Tatsuo; Osugi, Toshitaka; Satoh, Kunio; Hayasaka, Katsuhisa; Bando, Masaru.

    1993-10-01

    An availability of data and method for a design of metallic-fueled LMFBR is examined by using the experiment results of FCA assembly XVI-1. Experiment included criticality and reactivity coefficients such as Doppler, sodium void, fuel shifting and fuel expansion. Reaction rate ratios, sample worth and control rod worth were also measured. Analysis was made by using three-dimensional diffusion calculations and JENDL-2 cross sections. Predictions of assembly XVI-1 reactor physics parameters agree reasonably well with the measured values, but for some reactivity coefficients such as Doppler, large zone sodium void and fuel shifting further improvement of calculation method was need. (author)

  2. Vibrating fuel grapple. [LMFBR

    Science.gov (United States)

    Chertock, A.J.; Fox, J.N.; Weissinger, R.B.

    A reactor refueling method is described which utilizes a vibrating fuel grapple for removing spent fuel assemblies from a reactor core. It incorporates a pneumatic vibrator in the grapple head which allows additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

  3. Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

    International Nuclear Information System (INIS)

    Roake, W.E.

    1977-01-01

    Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals

  4. Strategies in development of advanced fuels for LMFBR

    International Nuclear Information System (INIS)

    Handa, Muneo

    1976-12-01

    Overseas strategies in development of advanced fuels for LMFBR are reviewed. Recent irradiation experiment and out-of-pile test data of the fuels are given in detail. The present status of development of oxide fueled LMFBR is also treated. (auth.)

  5. Fueling method in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Inoue, Kotaro.

    1985-01-01

    Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

  6. CAT reconstruction and potting comparison of a LMFBR fuel bundle

    International Nuclear Information System (INIS)

    Betten, P.R.; Tow, D.M.

    1984-04-01

    A standard Liquid Metal Fast Breeder Reactor (LMFBR) subassembly used in the Experimental Breeder Reactor II (EBR-II) was investigated, by remote techniques, for fuel bundle distortion by both nondestructive and destructive methods, and the results from both methods were compared. The non-destructive method employed neutron tomography to reconstruct the locations of fuel elements through the use of a maximum entropy reconstruction algorithm known as MENT. The destructive method consisted of ''potting'' (a technique that embeds and permanently fixes the fuel elements in a solid matrix) the subassembly, and then cutting and polishing the individual sections. The comparison indicated that the tomography reconstruction provided good results in describing the bundle geometry and spacer-wire locations, with the overall resolution being on the order of a spacer-wire diameter. A dimensional consistency check indicated that the element and spacer-wire dimensions were accurately reproduced in the reconstruction

  7. Overview of the fast reactors fuels program. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

  8. Biological behavior of mixed LMFBR-fuel-sodium aerosols

    International Nuclear Information System (INIS)

    Mahlum, D.D.; Hackett, P.L.; Hess, J.O.; Allen, M.D.

    1979-01-01

    Immediately after exposure of rats to mixed aerosols of sodium-LMFBR fuel, about 80 to 90% of the body burden of 239 Pu is in the gastrointestinal tract; 1.5 to 4% is in the lungs. With fuel-only aerosols, less of the body burden was in the GI tract and more in the lung and the head. Blood and urine values suggest an increased absorption of 239 Pu from sodium-fuel than from fuel-only aerosols

  9. Advanced methods for fabrication of PHWR and LMFBR fuels

    International Nuclear Information System (INIS)

    Ganguly, C.

    1988-01-01

    For self-reliance in nuclear power, the Department of Atomic Energy (DAE), India is pursuing two specific reactor systems, namely the pressurised heavy water reactors (PHWR) and the liquid metal cooled fast breeder reactors (LMFBR). The reference fuel for PHWR is zircaloy-4 clad high density (≤ 96 per cent T.D.) natural UO 2 pellet-pins. The advanced PHWR fuels are UO 2 -PuO 2 (≤ 2 per cent), ThO 2 -PuO 2 (≤ 4 per cent) and ThO 2 -U 233 O 2 (≤ 2 per cent). Similarly, low density (≤ 85 per cent T.D.) (UPu)O 2 pellets clad in SS 316 or D9 is the reference fuel for the first generation of prototype and commercial LMFBRs all over the world. However, (UPu)C and (UPu)N are considered as advanced fuels for LMFBRs mainly because of their shorter doubling time. The conventional method of fabrication of both high and low density oxide, carbide and nitride fuel pellets starting from UO 2 , PuO 2 and ThO 2 powders is 'powder metallurgy (P/M)'. The P/M route has, however, the disadvantage of generation and handling of fine powder particles of the fuel and the associated problem of 'radiotoxic dust hazard'. The present paper summarises the state-of-the-art of advanced methods of fabrication of oxide, carbide and nitride fuels and highlights the author's experience on sol-gel-microsphere-pelletisation (SGMP) route for preparation of these materials. The SGMP process uses sol gel derived, dust-free and free-flowing microspheres of oxides, carbide or nitride for direct pelletisation and sintering. Fuel pellets of both low and high density, excellent microhomogeneity and controlled 'open' or 'closed' porosity could be fabricated via the SGMP route. (author). 5 tables, 14 figs., 15 refs

  10. SIEX: a correlated code for the prediction of liquid metal fast breeder reactor (LMFBR) fuel thermal performance

    International Nuclear Information System (INIS)

    Dutt, D.S.; Baker, R.B.

    1975-06-01

    The SIEX computer program is a steady state heat transfer code developed to provide thermal performance calculations for a mixed-oxide fuel element in a fast neutron environment. Fuel restructuring, fuel-cladding heat conduction and fission gas release are modeled to provide assessment of the temperature. Modeling emphasis has been placed on correlations to measurable quantities from EBR-II irradiation tests and the inclusion of these correlations in a physically based computational scheme. SIEX is completely modular in construction allowing the user options for material properties and correlated models. Required code input is limited to geometric and environmental parameters, with a ''consistent'' set of material properties and correlated models provided by the code. 24 references. (U.S.)

  11. LMFBR fuel cycle studies progress report, August 1972, No. 42

    International Nuclear Information System (INIS)

    Unger, W.E.; Blanco, R.E.; Crouse, D.J.; Irvine, A.R.; Watson, C.D.

    1972-10-01

    This report continues a series outlining progress in the development of methods for reprocessing of LMFBR fuels. Development work is reported on problems of irradiated fuel transport to the processing facility, the dissolution of the fuel and the chemical recovery of PuO 2 --UO 2 values, the containment of volatile fission products, product purification, conversion of fuel processing plant product nitrate solutions to solids suitable for shipping and for subsequent fuel fabrication. Pertinent experimental results are presented for the information of those immediately concerned with the field. Detailed description of experimental work and data are included in the topical reports and in the Chemical Technology Division Annual Reports

  12. Blockages in LMFBR fuel assemblies: a review

    International Nuclear Information System (INIS)

    Han, J.T.; Fontana, M.H.

    1977-01-01

    Experimental and analytical investigations performed in the United States, Germany, Great Britain, and Japan on the effects of partial flow blockages in liquid-metal fast breeder reactor fuel assemblies are reviewed and the results presented. Generalized models are developed from experimental data obtained for blockages of various sizes, shapes, and porosity, with and without pins, utilizing water and sodium as the coolant. Generally, the recirculating flow in the wake behind a blockage is a relatively effective heat transfer mechanism. Experiments where sodium boiling was made to occur behind the blockages indicate that boiling is stable for the configurations tested; these results are predicted by analytical models. Blockages at the inlet of fuel assemblies tend to have insignificant effects in the fuel assembly unless flow is reduced grossly and therefore would be detectable. Blockages in the heat generating zone have to be quite large to cause sodium boiling under normal reactor operating conditions

  13. Blockages in LMFBR fuel assemblies: a review

    Energy Technology Data Exchange (ETDEWEB)

    Han, J T; Fontana, M H

    1977-01-01

    Experimental and analytical investigations performed in the United States, Germany, Great Britain, and Japan on the effects of partial flow blockages in liquid-metal fast breeder reactor fuel assemblies are reviewed and the results presented. Generalized models are developed from experimental data obtained for blockages of various sizes, shapes, and porosity, with and without pins, utilizing water and sodium as the coolant. Generally, the recirculating flow in the wake behind a blockage is a relatively effective heat transfer mechanism. Experiments where sodium boiling was made to occur behind the blockages indicate that boiling is stable for the configurations tested; these results are predicted by analytical models. Blockages at the inlet of fuel assemblies tend to have insignificant effects in the fuel assembly unless flow is reduced grossly and therefore would be detectable. Blockages in the heat generating zone have to be quite large to cause sodium boiling under normal reactor operating conditions.

  14. Fire protection device for LMFBR spent fuel shipping casks

    International Nuclear Information System (INIS)

    Moss, M.; Heckman, R.C.

    1977-08-01

    This report describes the analysis and experimental evaluation of a device to provide fire protection for a shipping cask used to transport liquid metal fast breeder reactor (LMFBR) fuel rods. Thermal analyses of various fire protection schemes were conducted by means of the finite difference code, CINDA. The choice of materials for the cask body was determined to be less important than the dimensions and number of cooling fins attached to the cask. Of several protection methods considered, radiation/convection shields between the cooling fins looked most attractive; these were tested on a quarter-scale cask model. The shields consisted of bimetallic bands which expanded when heated, blocking the fire, and retracted when cooled. This automatic reversibility would allow the fins to dissipate internally generated heat of a full-size cask once the fire was out. Tests showed that, even in an asymmetric fire exposure, the bands expanded to provide protection to the cask. The directly exposed surface of the cask model, protected this way, reached a temperature of only 180 0 C compared with 295 0 C in the unprotected state, when subjected to a butane flame for 0.5 h at a distance of 15 cm. Greater relative effectiveness could be expected in an engulfing fire in which the bands would expand more symmetrically

  15. Natural circulation in simulated LMFBR fuel assemblies

    International Nuclear Information System (INIS)

    Levin, A.E.; Carbajo, J.J.; Lloyd, D.B.; Montgomery, B.H.; Rose, S.D.; Wantland, J.L.

    1985-01-01

    Natural circulation experiments have been performed using simulated liquid metal fast breeder reactor fuel assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale sodium loop. Objective of these tests has been to provide experimental data under conditions that might be encountered during a partial or total loss of the shutdown heat removal system (SHRS) in a reactor. The experiments have included single- and two-phase tests under quasi-steady and transient conditions, at both nominal and non-nominal system conditions. Results from these test indicate that the potential for reactor damage during degraded SHRS operation is extremely slight, and that natural circulation can be a major contributor to safe operation of the system in both single- and two-phase flow during such operation

  16. LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

    The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

  17. Evaluation of LMFBR fuel-motion diagnostics instrumentation with PARKA

    International Nuclear Information System (INIS)

    Evans, A.E. Jr.; Orndoff, J.D.; Talbert, W.L. Jr.

    1978-01-01

    To aid in the design of LMFBR safety test experiments and safety test facilities (STF), a program of evaluation of concepts for fuel-motion diagnostics instrumentation has been undertaken. A part of this evaluation is beng done at PARKA, a Rover project critical assembly which has been modified to study the self nuclear image from driven FTR-type fuel assemblies. Feasibility of obtaining fast-neutron images of single-pin voids in assemblies of up to 127 fuel pins has been demonstrated, albeit marginally for the larger fuel bundles. The feasibility of using in-core detectors as fuel-motion monitors has also been studied. Use of PARKA in a pulsed mode to study STF transient phenomena is discussed

  18. LMFBR spent fuel and new fuel inspection and handling

    International Nuclear Information System (INIS)

    Youngquist, C.H.

    1975-01-01

    If the hot cells used for inspection and preparation of liquid metal fast breeder reactor fuel assemblies can be located so that the fuel assemblies can be moved directly to and from the sodium filled decay storage tank, the fuel handling problems can be simplified and a fuel handling machine may be eliminated. The hot cells can also serve as a transfer lock through the outer containment barrier, thereby permitting uninterrupted reactor operation during fuel transfer to and from the decay storage tank. 5 references

  19. LMFBR fuel assembly design for HCDA fuel dispersal

    Science.gov (United States)

    Lacko, Robert E.; Tilbrook, Roger W.

    1984-01-01

    A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

  20. Dissolution of LMFBR fuel-sodium aerosols

    International Nuclear Information System (INIS)

    Allen, M.D.; Moss, O.R.

    1979-01-01

    Plutonium dioxide, normally insoluble in biological fluids, becomes much more soluble when mixed with sodium as the aerosol is formed. Sodium-fuel aerosols are approximately 20 times less soluble in simulated lung fluid than in distilled water. Solubility of sodium-fuel aerosols increases when Na 2 CO 3 are added to the distilled-water dissolution fluid. Mixed-oxide fuel aerosols without sodium present are relatively insoluble in distilled water, simulated lung fluid, and distilled water with Na 2 CO 3 and NaHCO 3 added

  1. Instrumentation for core and coolant monitoring in liquid-metal fast breeder reactors (LMFBR)

    International Nuclear Information System (INIS)

    Hess, B.; Ruppert, E.

    1975-01-01

    The review on core and coolant instrumentation for liquid metal fast breeders aims to give a short survey of measurement methods and the variety of appropriate instrumentation developed and tested for reactor application throughout the world. The introductory part gives a general outline of instrumentation development, partly as the refinement of well-known thermal reactor instrumentation and partly as the special instrumentation demanded for LMFBR safety requirements, some aspects of which are also discussed briefly. The in-core LMFBR instrumentation is surveyed, classifying the measurement or monitoring of coolant properties such as temperature, pressure, flow and acoustic emission and the measurement of core-kinetic quantities such as neutron flux and reactivity. Without considering the fundamentals of the measurements, the state of instrument development is reviewed and, where known, future aspects are indicated. An additional review on fuel failure detection methods and the related instrumentation distinguishes between global or whole-core detection methods and those used for localization of failures. Special attention is paid to the aspect of reactor safety and its reliability as one of the major objectives of these detection methods. A summary of the protective systems and instrumentation already used or foreseen for LMFBR plants forms a transition to a very brief discussion of handling and interpretation of the multitude of data derived from the rather comprehensive LMFBR instrumentation. This state of the art review claims neither to be complete at the time published nor to be a detailed guide to special problems of instrumentation development, the solutions to which are normally part of industrial knowhow. (author)

  2. Compatibility of niobium, titanium, and vanadium metals with LMFBR cladding

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1975-10-01

    A series of laboratory capsule annealing experiments were conducted to assess the compatibility of niobium, vanadium, and titanium with 316 stainless steel cladding in the temperature range of 700 to 800 0 C. Niobium, vanadium, and titanium are cantidate oxygen absorber materials for control of oxygen chemistry in LMFBR fuel pins. Capsule examination indicated good compatibility between niobium and 316 stainless steel at 800 0 C. Potential compatibility problems between cladding and vanadium or titanium were indicated at 800 0 C under reducing conditions. In the presence of Pu/sub 0.25/U/sub 0.75/O/sub 1.98/ fuel (Δanti G 02 congruent to -160 kcal/mole) no reaction was observed between vanadium or titanium and cladding at 800 0 C

  3. Compatibility of niobium, titanium, and vanadium metals with LMFBR cladding

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, C.N.

    1975-10-01

    A series of laboratory capsule annealing experiments were conducted to assess the compatibility of niobium, vanadium, and titanium with 316 stainless steel cladding in the temperature range of 700 to 800/sup 0/C. Niobium, vanadium, and titanium are cantidate oxygen absorber materials for control of oxygen chemistry in LMFBR fuel pins. Capsule examination indicated good compatibility between niobium and 316 stainless steel at 800/sup 0/C. Potential compatibility problems between cladding and vanadium or titanium were indicated at 800/sup 0/C under reducing conditions. In the presence of Pu/sub 0.25/U/sub 0.75/O/sub 1.98/ fuel (..delta..anti G/sub 02/ congruent to -160 kcal/mole) no reaction was observed between vanadium or titanium and cladding at 800/sup 0/C.

  4. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. First, a thermal creep damage index is set up through a sufficiently sophisticated clad physical analysis including arbitrary time dependence of power and neutron flux as well as effects of sodium temperature, burnup and steel mechanical behavior. Although this strain limit approach implies a more general but time consuming model., on the counterpart the net output is improved and e.g. clad temperature, stress and strain maxima may be easily assessed. A full spectrum of variables are statistically treated to account for their probability distributions. Creep damage probability may be obtained and can contribute to a quantitative fuel probability estimation

  5. Upon local blockage formations in LMFBR fuel rod bundles with wire-wrapped spacers

    International Nuclear Information System (INIS)

    Minden, C. v.; Schultheiss, G.F.

    1982-01-01

    A theoretical and experimental study, to improve understanding of local particle depositions in a wire-wrapped LMFBR fuel bundle, has been performed. Theoretical considerations show, that a preferentially axial process of particle depositions occurs. The experiments confirm this and clarify that the blockages arise near the particle source and settle at the spatially arranged minimum gaps in the bundle. The results suggest that, considering flow reduction, cooling and DND-detection, such fuel particle blockages are less dangerous. With reference to these safety-relevant factors, wire-wrapped LMFBR fuel bundles seem to gain advantages compared to the grid design. (orig.) [de

  6. A LMFBR for thorium utilization and for the U233/Th fuel rods specification

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Dias, A.F.

    1982-01-01

    The use of U 233 /Th as fuel in the middle part of LMFBR core and the Pu/U in the external part of the core, are proposed. The basic neutronic and safety characteristics and the specifications of fuel rods to be used in the internal core, are presented. (E.G.) [pt

  7. The state of art of the methods for thermohydraulics design of LMFBR fuel elements

    International Nuclear Information System (INIS)

    Fernandez y Fernandez, E.; Carajilescov, P.

    1981-09-01

    The present (experimental and analytical) state of art of the methods for thermohydraulics design of LMFBR fuel elements is analyzed. A development program is suggested, in order to obtain a computer code for modelling the distribution of coolant enthalpy in reactor core. This computer code is in development. (Author) [pt

  8. Thermohydraulic and thermal stress aspects of a porous blockage in an LMFBR fuel assembly

    International Nuclear Information System (INIS)

    Kuzay, T.M.; Marr, W.W.; Helenberg, H.W.; Ariman, T.; Wilson, R.E.; Pedersen, D.R.

    1979-01-01

    The current safety scenarios of Liquid Metal Fast Breeder Reactors (LMFBR) under local fault propagation include the study of a hypothetical accident initiated by the formation of an external debris porous blockage in a fuel subassembly. In this preliminary experimental and analytical investigation, a non-heat-generating porous blockage was postulated to cover 18 flow channels of a 37 pin Fast Test Reactor (FTR) type fuel subassembly. The axial extent of the blockage is 50 mm. The blockage material is stainless steel (SS 316) with 30 percent average porosity (percent void volume). The blockage and the pins were modeled with a finite element technique and the thermal field in the blockage was predicted. This thermal field was utilized to do a planar thermal stress analysis of the postulated blockage. To verify the analytical model and also to better understand the thermal-hydraulics of such a porous blockage out-of-pile tests were conducted in a sodium loop. Data from the out-of-pile tests was utilized to calibrate and improve the analytical model

  9. Refractory metal carbide coatings for LMFBR applications: a systems approach

    International Nuclear Information System (INIS)

    Gotschall, H.L.; Ople, F.S.; Riccardella, P.C.

    1975-01-01

    The selection, testing and improvement of high density, tightly bonded plasma and detonation gun coatings designed to meet LMFBR core component criteria are described. The process descriptions include a review of the important developments in substrate surface preparation which were required to ensure strong bonding and to minimize interface contamination. Coating finishing techniques which were developed to optimize friction behavior are also described

  10. Contribution of Clinch River Breeder Reactor plant design and development to the LMFBR fuel cycle

    International Nuclear Information System (INIS)

    Riley, D.R.; Dickson, P.W.

    1981-01-01

    This paper describes how the CRBRP development and CRBRP focus of the LMFBR base technology program have led to advances in the state of the art in physics, thermal-hydraulics, structural analysis, core restraint, seismic analysis, and analysis of hypothetical core-disruptive accident energetics, all of which have been incorporated through disciplined engineering into the final CRBRP design. The total development in the US of fuels and materials, the analytical advances made on CRBRP design, and the incorporation of the latest experimental results into that design have put the US technology in general and the CRBRP design in particular at the forefront of technology. This has placed the US in a position to develop the most favorable LMFBR fuel cycle

  11. LMFBR source term experiments in the Fuel Aerosol Simulant Test (FAST) facility

    Energy Technology Data Exchange (ETDEWEB)

    Petrykowski, J.C.; Longest, A.W.

    1985-01-01

    The transport of uranium dioxide (UO/sub 2/) aerosol through liquid sodium was studied in a series of ten experiments in the Fuel Aerosol Simulant Test (FAST) facility at Oak Ridge National Laboratory (ORNL). The experiments were designed to provide a mechanistic basis for evaluating the radiological source term associated with a postulated, energetic core disruptive accident (CDA) in a liquid metal fast breeder reactor (LMFBR). Aerosol was generated by capacitor discharge vaporization of UO/sub 2/ pellets which were submerged in a sodium pool under an argon cover gas. Measurements of the pool and cover gas pressures were used to study the transport of aerosol contained by vapor bubbles within the pool. Samples of cover gas were filtered to determine the quantity of aerosol released from the pool. The depth at which the aerosol was generated was found to be the most critical parameter affecting release. The largest release was observed in the baseline experiment where the sample was vaporized above the sodium pool. In the nine ''undersodium'' experiments aerosol was generated beneath the surface of the pool at depths varying from 30 to 1060 mm. The mass of aerosol released from the pool was found to be a very small fraction of the original specimen. It appears that the bulk of aerosol was contained by bubbles which collapsed within the pool. 18 refs., 11 figs., 4 tabs.

  12. A Swiss contribution to a secure LMFBR fuel cycle

    International Nuclear Information System (INIS)

    Nicolet, M.; Bischoff, K.; Hausmann, W.; Stofer, B.

    1978-12-01

    Since 1967, EIR has been using the sphere-pac fuel concept, which takes advantage of the wet route fabrication of (U,Pu) carbide-microspheres using an internal gelation method, followed by carbothermic reduction of the precipitated metal-oxides. Some of the promises of the wet process are a shorter fabrication route than for pellet manufacture, no dust problems, reduced fire hazard for carbides, and last but not least the improvement of Pu safeguards. The method is particularly suitable for direct coupling to a reprocessing plant, where coprocessing of both U and Pu and spiked solutions will be possible. (Auth.)

  13. Safety research needs for carbide and nitride fueled LMFBR's. Final report

    International Nuclear Information System (INIS)

    Kastenberg, W.E.

    1975-01-01

    The results of a study initiated at UCLA during the academic year 1974--1975 to evaluate and review the potential safety related research needs for carbide and nitride fueled LMFBR's are presented. The tasks included the following: (1) Review Core and primary system designs for any significant differences from oxide fueled reactors, (2) Review carbide (and nitride) fuel element irradiation behavior, (3) Review reactor behavior in postulated accidents, (4) Examine analytical methods of accident analysis to identify major gaps in models and data, and (5) Examine post accident heat removal. (TSS)

  14. Performance of LMFBR fuel pins with (Pu,Th)O/sub 2-x/ and UO2

    International Nuclear Information System (INIS)

    Lawrence, L.A.

    1983-09-01

    The irradiation performance of (Pu,Th)O/sub 2-x/ and UO 2 fueled pins for breeder reactor application were compared to the extensive performance data base for the (U,Pu)O/sub 2-x/ fuel system. Th-Pu and 238 U- 233 U based fuel systems were candidate fuel fertile/fissile isotopic combinations for development of alternatives to the current LMFBR fuel cycle. Initial screening tests were conducted in the EBR-II to obtain comparative performance data because of the limited experience with these fuel systems. In some cases, 235 U was used as a substitute for 233 U because of the difficulties in fabrication of available 233 U due to its high gamma ray emission rate

  15. The safety of storage facilities for LMFBR spent fuels

    International Nuclear Information System (INIS)

    Lefort, G.; Puit, J.C.

    1982-04-01

    Storage conditions for the cooling of fast neutron spent fuel assembly, before reprocessing, will be adapted specifically to this fuel cycle, but in a preliminary testing period it is better to take advantage of the industrial storage conditions already used for spent fuel of light water reactors, even if no extrapolation is possible in the future. In this aim fuel assemblies are dismantled, as soon as thermal conditions allow it, and pins are gathered in leak proof containers in a gas atmosphere and put in a cooling pool. This solution gives good results but studies and experiments are resumed, for future processing, for safe, easy and economical transport and storage conditions [fr

  16. On LMFBR corrosion. Part II: Consideration of the in-reactor fuel-cladding system

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Walker, C.T.; Whitlow, W.H.

    1976-05-01

    The scientific and technological aspects of LMFBR cladding corrosion are discussed in detail. Emphasis is placed on the influence of the irradiation environment and the effect of fuel and filler-gas impurities on the corrosion process. These studies are complemented by a concise review of out-of-pile simulation experiments that endeavour to clarify the role of the aggressive fission products cesium, tellurium and iodine. The principal models for cladding corrosion are presented and critically assessed. Areas of uncertainty are exposed and some pertinent experiments are suggested. Consideration is also given to some new observations regarding the role of stress in fuel-cladding reactions and the formation of ferrite in the corrosion zone of the cladding during irradiation. Finally, two technological solutions to the problem of cladding corrosion are proposed. These are based on the use of an oxygen buffer in the fuel and the application of a protective coating to the inner surface of the cladding

  17. Materials properties utilization in a cumulative mechanical damage function for LMFBR fuel pin failure analysis

    International Nuclear Information System (INIS)

    Jacobs, D.C.

    1977-01-01

    An overview is presented of one of the fuel-pin analysis techniques used in the CRBRP program, the cumulative mechanical damage function. This technique, as applied to LMFBR's, was developed along with the majority of models used to describe the mechanical properties and environmental behavior of the cladding (i.e., 20 percent cold-worked, 316 stainless steel). As it relates to fuel-pin analyses the Cumulative Mechanical Damage Function (CDF) continually monitors cladding integrity through steady state and transient operation; it is a time dependent function of temperature and stress which reflects the effects of both the prior mechanical history and the variations in mechanical properties caused by exposure to the reactor environment

  18. Nonlinear transient deformation of LMFBR fuel elements under impulsive loading

    International Nuclear Information System (INIS)

    Liebe, R.

    1975-01-01

    Subject of this paper is the elastoplastic transient behavior of one subassembly under given space- and time-dependent pressure loading. The interaction of several colliding fuel elements including coolant dynamics is briefly discussed. Theoretical models: description is given of physical mechanisms (bending of the slender subassembly duct, flattening of the thin-walled hexagonal cross section, pin-bundle and coolant action). Discrete structural models are compared and applied to the coupled fluid-structure problem. The use of a special beam element with lumped masses and discrete elastoplastic hinges with concentrated moments allows the effective simulation of the structure. An incompressible Eulerian nonstationary flow model including friction is used with a 4th order variable step Runge-Kutta method. Static and impact experiments: description is given of a single subassembly experimental program on a special droptower facility which is used for static and impact loading of 1:1 SNR-300 type fuel element models. Also described is a load and impact controlling device utilizing honeycomb crushing material which allows to produce desired pressure pulses (rise time, peak load, impulse, pulse shape). Forces, accelerations and strains are measured at the subassembly; its deformation is filmed with a high-speed camera and recorded by specially designed photodiode-arrays. Results and conclusions: good agreement is observed between experimental and theoretical response data. Parameter studies confirmed the significant role of duct flexure, local cross section flattening, material ductility and strain-hardening. The stiffening effects of the combined deformation of pin-bundle and spacer-grids is very strong, and most of the energy is dissipated by localized yielding of the hexcan

  19. Post-accident fuel relocation and heat removal in the LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Kazimi, M S; Tsai, S S; Gasser, R D

    1976-08-01

    Assessment of the dynamics of post-accident fuel relocation and heat removal is an important aspect of the evaluation of the consequences of a hypothetical accident in an LMFBR. Such an assessment is of particular importance in the evaluation of the post-accident radiological doses around the reactor site. In the present evaluation particular attention is given to the design features of the Clinch River Breeder Reactor Plant (CRBR). Fuel relocation and heat removal, assuming certain conditions have resulted in core disruption, are discussed. The discussion of events and phenomena involved in the relocation processes is centered around the resulting patterns of heat source distribution. The factors influencing fuel relocation and distribution in the inlet and outlet plena of the reactor vessel are discussed. The current technology of in-vessel heat removal is applied to the design of the CRBR reactor. Both fuel debris cooling limits and overall coolant flow in the reactor under natural convection conditions are explored. Some of the uncertainties in ex-vessel fuel behavior are addressed. In particular, the effect of melting the cavity bed on the rate of growth of a molten fuel pool is investigated.

  20. Synthesis Report on the understanding of failed LMFBR fuel element performance

    International Nuclear Information System (INIS)

    Plitz, H.; Bagley, K.; Harbourne, B.

    1990-07-01

    In the coarse of LMFBR operation fuel element failures cannot entirely be avoided as experienced during the operation of PFR, PHENIX and KNK II, where 44 failed fuel elements have been registered between 1978 and 1989. In earlier irradiations, post irradiation examinations showed mixed oxide pin diameter increases up to pin pitch distance, urging to stress reactor safety questions on the potential of fuel pin failure propagation within pin bundles. The chemical interaction of sodium with mixed oxide fuel is regarded to be the key for the understanding of failed fuel behavior. Valuable results on the failed fuel pin behavior during operation were obtained from the SILOE sodium loop test. Based on the bulk of experience with the detection of fuel pin failures, with the continued operation and with the handling of failed pins respectively elements, one can state: 1. All fuel pin failures have been detected securely in time and have been located. 2. Small defects are developing slowly. 3. Even large defects at end-of-life pins resulted in limited fuel loss. 4. Clad failures behave benign in main aspects. 5. The chemical interaction of sodium with mixed oxide is an important factor in the behavior of failed fuel pins, especially at high burnup. 6. Despite different pin designs and different operation conditions, on the basis of 44 failed elements in PFR, PHENIX and KNK II no pin-to-pin propagation was observed and fuel release was rather low, often not detectable. 7. In no case hazard conditions affecting reactor safety have been experienced

  1. Specialists' meeting on theoretical modelling of LMFBR fuel pin behaviour. Summary report

    International Nuclear Information System (INIS)

    1979-12-01

    The purpose of the meeting was to provide an opportunity for exchanging views of theoretical modelling of LMFBR fuel pin behaviour and to summarise the IWGFR member countries' knowledge in this field. The special emphasis was placed on normal operating conditions. The technical part of the meeting was divided into six sessions, as follows: An overview of fuel modelling studies; Key factors and basic phenomena relevant to fuel pin behaviour modelling; Application to steady state operation and normal transients; Experimental validation through pins in service and specific irradiation experiments; Advanced fuels; and Brief review of existing codes. During the meeting, papers were presented by the delegates on behalf of their countries or organization. The papers, which are included in this report, were either in the form of a general survey of the subject, or on specific technical subjects. In each subject area presentations appropriate to the subject were made from the submitted papers. The presentations were followed by discussions of the questions raised and summary is made

  2. Specialists' meeting on LMFBR fuel rod behaviour under operational transients, Kalpakkam, India, 3-6 December 1985

    International Nuclear Information System (INIS)

    1985-12-01

    IWGFR Specialists' Meeting on ''LMFBR Fuel Rod Behaviour Under Operational Transients'' was held in Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 3-6 December 1985. The meeting was attended by the representatives of Belgium, France, FRG, India, Italy, US, USSR, IAEA and observers from India. The purpose of the meeting was to provide a forum for the exchange of information on the subject of LMFBR Fuel Rod Behaviour Under Operational Transients. The meeting presentations were divided into sessions devoted to the following topics: Overview of National Programmes (3 papers); In-pile and out-pile experimental facilities and results (5 papers); Modelling and Code work (5 papers). A separate abstract was prepared for each of these papers

  3. Detailed design consideration on wire-spaced LMFBR fuel subassemblies under the effects of uncertainties and non-nominal geometries

    International Nuclear Information System (INIS)

    Hishida, H.

    1979-01-01

    This paper explains some analytical methods for evaluating the effects of deviation in subchannel coolant flow rate from the nominal value due to fuel pin bundle deflection and manufacturing tolerances and of inter-sub-channel coolant mixing and local temperature rise due to a wire-spacer on the hot spot temperature. Numerical results are given in each chapter with respect to a prototype LMFBR core. (author)

  4. Noise and DC balanced outlet temperature signals for monitoring coolant flow in LMFBR fuel elements

    International Nuclear Information System (INIS)

    Edelmann, M.

    1977-01-01

    Local cooling disturbances in LMFBR fuel elements may have serious safety implications for the whole reactor core. They have to be detected reliably in an early stage of their formation therefore. This can be accomplished in principle by individual monitoring of the coolant flow rate or the coolant outlet temperature of the sub-assemblies with high precision. In this paper a method is proposed to increase the sensitivity of outlet temperature signals to cooling disturbances. Using balanced temperature signals provides a means for eliminating the normal variations from the original signals which limit the sensitivity and speed of response to cooling disturbances. It is shown that a balanced signal can be derived easily from the original temperature signal by subtracting an inlet temperature and a neutron detector signal with appropriate time shift. The method was tested with tape-recorded noise signals of the KNK I reactor at Karlsruhe. The experimental results confirm the theoretical predictions. A significant reduction of the uncertainty of measured outlet temperatures was achieved. This enables very sensitive and fast response monitoring of coolant flow. Furthermore, it was found that minimizing the variance of the balanced signal offers the possibility for a rough determination of the heat transfer coefficient of the fuel rods during normal reactor operation at power. (author)

  5. FMEF profilometry and visual examination feasibility and conceptual design. [Fuels and Materials Examination Facility; LMFBR and GCFR

    Energy Technology Data Exchange (ETDEWEB)

    Philipp, L.D.; Dilbeck, R.A.; Hartman, J.S.; Hildebrand, B.P.; Reich, F.R.; Swinth, K.L.

    1976-07-01

    The Fuels and Materials Examination Facility (FMEF) is being scoped to provide postirradiation examination capabilities for FFTF, LMFBR and GCFR fuels and materials. The Hanford Engineering Development Laboratory has requested that the Battelle Pacific Northwest Laboratory (PNL) complete a feasibility study for the development of equipment to meet the FMEF Measurement Requirements for irradiated fuel pin and absorber rod bow, length, profile and visual examination stations. The purpose of the report is to provide a conceptual design for development of the examination equipment. The design analysis assumes that fuel pins and absorber rods to be examined are in the main cell. The cell's environment will be argon or nitrogen gas at a pressure between --1 and --4 in. of water and at a temperature between 70 and 100/sup 0/F. Oxygen content of the cell gas will normally be controlled between 25 and 50 ppM. Water content will be controlled within the same limits.

  6. PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kirk, W.L. (comp.)

    1976-08-01

    The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters.

  7. CRT code and results of some studies for power, reactivity and temperatures for LMFBR fuel rods under operational transients

    International Nuclear Information System (INIS)

    Om Pal Singh; Ponpondi, S.; Parikh, M.V.

    1985-01-01

    The paper describes the details of the computer code, CRT, that has been developed to study the reactor power, reactivity and temperature transients in LMFBR fuel rods under operational transients. The code is based upon suitable modelling of reactor neutron kinetics, heat transfer phenomena and reactivity feedback effects coming from: axial/radial expansion of the fuel/clad/coolant; core boundary movement to axial and radial blankets; sodium entry/expulsion during the radial expansion of the core and apparent insertion/removal of control rods during core/reactor vessel expansion. Further, the results of some studies like, transient behaviour of reactor power and temperature distribution in fuel rods for reactor at low and high initial powers and for fresh and irradiated fuels; comparison of lumped and exact heat transfer models and evaluation of limiting reactivity addition rates for the lumped model to be adequate; influence of the gap conductance on the temperature distributions inside the fuel pellet and the clad and break up of reactivity coming from different feedback mechanisms and their dependence upon the heat transfer parameters are also presented. (author)

  8. Metal fuel safety performance

    International Nuclear Information System (INIS)

    Miles, K.J. Jr.; Tentner, A.M.

    1988-01-01

    The current development of breeder reactor systems has lead to the renewed interest in metal fuels as the driver material. Modeling efforts were begun to provide a mechanistic description of the metal fuel during anticipated and hypothetical transients within the context of the SAS4A accident analysis code system. Through validation exercises using experimental results of metal fuel TREAT tests, confidence is being developed on the nature and accuracy of the modeling and implementation. Prefailure characterization, transient pin response, margins to failure, axial in-pin fuel relocation prior to cladding breach, and molten fuel relocation after cladding breach are considered. Transient time scales ranging from milliseconds to many hours can be studied with all the reactivity feedbacks evaluated

  9. Influence of strain rate and temperature on the yield and fracture toughness behavior of selected steels for an LMFBR spent fuel shipping cask, a literature assessment

    International Nuclear Information System (INIS)

    Rack, H.J.

    1976-02-01

    The literature has been reviewed to determine the possible influences of strain rate and temperature on the yield and fracture toughness behavior of selected steels suggested for use in an LMFBR Spent Fuel Shipping Cask. Based on this information, recommendations have been made for further work which is intended to alleviate potential problems prior to their having a major impact on the shipping cask program

  10. Commercial liquid-metal MHD conversion systems coupled to LMFBR and coal-fired fluidized bed combustors

    International Nuclear Information System (INIS)

    Amend, W.E.; Brunsvold, A.; Pierson, E.S.

    1975-01-01

    The constraints imposed on two-phase liquid-metal MHD (LMMHD) when employed in commercial power plants with practical heat sources have not previously been studied. The coupling of a LMMHD power system with an LMFBR and a coal-fired fluidized bed combustor are considered. Two MHD systems are considered. The first is a dual cycle where heat is added to both the liquid metal and the gas, and the gas may expand through a gas turbine after the MHD generator. The second system, a binary cycle, differs in that a significant portion of the sensible heat in the gas entering the compression loop is converted to useful power in a steam bottoming cycle. The effect of liquid-metal vapor carry-over into the gas loop is included. The couplings of the LMMHD system with the heat sources and with the steam plants were studied in depth. The results of the study of each interface are presented parametrically for each heat source and energy conversion system. Operating points have been selected and the complete schematic of each system considered is presented along with all thermodynamic state points and fluid flow rates. All system parameters and component efficiencies were selected to be consistent with near term technology and good engineering design principles. These criteria yielded a system performance of 37 percent for an LMFBR operating with a maximum reactor coolant temperature of 1200 0 F when the pure LMMHD energy converter was used. A LMMHD/steam binary cycle is shown to be capable of achieving a thermal efficiency of 44.8 percent when used with the same heat source. Results with the coal fluidized bed combustor as a heat source show even higher performance levels (about 50 percent efficiency) since the maximum cycle temperature is increased

  11. Accident considerations in LMFBR design

    International Nuclear Information System (INIS)

    Simpson, D.E.; Alter, H.; Fauske, H.K.; Hikido, K.; Keaten, R.W.; Stevenson, M.G.; Strawbridge, L.

    1975-12-01

    LMFBR safety design criteria are discussed from the standpoints of accident severity classification and damage criteria, and the following design events are considered: fuel failure propagation, reactivity addition faults, heat transport system events, steam generator faults, sodium spills, fuel handling and storage faults, and external events

  12. On the hazard accumulation of actinide waste in a Pu-fueled LMFBR power economy with and without by-product actinide recycling

    International Nuclear Information System (INIS)

    Anselmi, L.; Caruso, K.; Hage, W.; Schmidt, E.

    1979-01-01

    The actinide waste arisings in terms of hazard potential for ingestion and inhalation are given for a Pu-fueled LMFBR Power Economy as function of decay time. The data were assessed for two simplified fuel cycles, one considering the recycling of by-product actinides and the other their complete discharge to the high-level waste. Two durations of nuclear power and several loss fractions of actinides to the waste were considered. The major contributors in form of chemical elements or isotopes to the actinide waste hazard built up during the nuclear power duration were identified for various decay intervals

  13. Effects of duct configuration on flow and temperature structure in sodium-cooled 19-rod simulated LMFBR fuel bundles with helical wire-wrap spacers

    International Nuclear Information System (INIS)

    Wantland, J.L.; Fontana, M.H.; Gnadt, P.A.; Hanus, N.; MacPherson, R.E.; Smith, C.M.

    1976-01-01

    Thermal-hydrodynamic testing of sodium-cooled 19-rod simulated LMFBR fuel bundles is being conducted at the O ak Ridge National Laboratory in the Fuel Failure Mockup (FFM), an engineering-scale high-temperature sodium facility which provides prototypic flows, temperatures and power densities. Electrically heated bundles have been tested with two scalloped and two hexagonal duct configurations. Peripheral helical flows, attributed to the spacers, have been observed with strengths dependent upon the evenness and relative sizes of the peripheral flow areas. Diametral sodium temperature profiles are more uniform with smaller peripheral flow areas

  14. FFTF metal fuel pin fabrication

    International Nuclear Information System (INIS)

    Dittmer, J.O.; Benecke, M.W.; Feigenbutz, L.V.

    1989-01-01

    A major new initiative to develop, irradiate, and qualify a binary uranium/zirconium metal-fuel system in the Fast Flux Test Facility (FFTF) has been implemented by the Westinghouse Hanford Company for the US Department of Energy. Metal-fuel test assemblies have been designed and fabricated, and are now being irradiated in FFTF to provide the data needed to support the potential use of binary metal fuels in FFTF and other liquid-metal reactors. These development efforts support licensing activities for metal-fuel use in near-term advanced liquid-metal reactors

  15. Metallic fuel design development

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

    1999-04-01

    This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual

  16. Metallic fuel design development

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

    1999-04-01

    This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual design technology on metallic fuel

  17. Fuel-disruption experiments under high-ramp-rate heating conditions. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Wright, S.A.; Worledge, D.H.; Cano, G.L.; Mast, P.K.; Briscoe, F.

    1983-10-01

    This topical report presents the preliminary results and analysis of the High Ramp Rate fuel-disruption experiment series. These experiments were performed in the Annular Core Research Reactor at Sandia National Laboratories to investigate the timing and mode of fuel disruption during the prompt-burst phase of a loss-of-flow accident. High-speed cinematography was used to observe the timing and mode of the fuel disruption in a stack of five fuel pellets. Of the four experiments discussed, one used fresh mixed-oxide fuel, and three used irradiated mixed-oxide fuel. Analysis of the experiments indicates that in all cases, the observed disruption occurred well before fuel-vapor pressure was high enough to cause the disruption. The disruption appeared as a rapid spray-like expansion and occurred near the onset of fuel melting in the irradiated-fuel experiments and near the time of complete fuel melting in the fresh-fuel experiment. This early occurrence of fuel disruption is significant because it can potentially lower the work-energy release resulting from a prompt-burst disassembly accident.

  18. Research on nuclear energy in the fields of fuel cycle, PWR reactors and LMFBR reactors

    International Nuclear Information System (INIS)

    Barre, B.; Camarcat, N.

    1995-01-01

    In this article we present the CEA research programs to improve the safety of the next generation of reactors, to manage the Plutonium and the wastes of the fuel cycle end and to ameliorate the competitiveness. 6 refs

  19. Development of the delyed-neutron triangulation technique for locating failed fuel in LMFBR

    International Nuclear Information System (INIS)

    Kryter, R.C.

    1975-01-01

    Two major accomplishments of the ORNL delayed neutron triangulation program are (1) an analysis of anticipated detector counting rates and sensitivities to unclad fuel and erosion types of pin failure, and (2) an experimental assessment of the accuracy with which the position of failed fuel can be determined in the FFTF (this was performed in a quarter-scale water mockup of realistic outlet plenum geometry using electrolyte injections and conductivity cells to simulate delayed-neutron precursor releases and detections, respectively). The major results and conclusions from these studies are presented, along with plans for further DNT development work at ORNL for the FFTF and CRBR. (author)

  20. 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.

  1. Prediction of thermal capability for a proposed LMFBR spent fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Reed, A.W.

    1979-02-01

    Prediction of the thermal capability of a fast breeder spent fuel shipping cask requires consideration of a large number of variables, such as geometry, canister packing fluid, gaseous coolant pressure, surface emissivities, and fin densities. In order to evaluate the individual influences of the large number of controlling parameters, a parametric analysis of a conceptual cask has been performed. Results of the analysis are applied to four options, each having a different configuration. Comparison of the options indicates that substantial increases in cask thermal capability may be obtained by the use of liquid sodium as a primary coolant and aluminum as a basket material.

  2. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, September 1, 1976--November 30, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1976-01-01

    Information is presented concerning bundle geometry with wrapped and bare rods, subchannel geometry with bare rods, LMFBR outlet plenum flow mixing, and theoretical determination of local temperature fields in LMFBR fuel rod bundles.

  3. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, March 1, 1977--May 31, 1977

    International Nuclear Information System (INIS)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1977-01-01

    Progress is summarized in the following tasks: (1) bundle flow studies (wrapped and bare rods); (2) subchannel flow studies (bare rods); (3) LMFBR outlet plenum flow mixing; and (4) theoretical determination of local temperature fields in LMFBR fuel rod bundles

  4. Fuel Coolant Thermal Interaction Project. Quarterly progress report No. 2, October 1, 1975--December 31, 1975. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.

    1976-03-01

    The objective of the work reported is to experimentally and analytically study the dominant mechanisms in fuel coolant thermal interactions which could lead to vapor explosions. The exploration of mechanisms is focused in two areas: (a) mechanisms responsible for fragmentation in molten metal droplet experiments (including assessment of the validity of the proposed spontaneous nucleation mechanism); and (b) thermal stress initiated fracture as a fragmentation mechanism. Work being performed in these areas is briefly described.

  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. LMFBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1988-01-01

    Purpose: To flatten the power distribution while maintaining the flattening in the axial power distribution in LMFBR type reactors. Constitution: Main system control rods are divided into control rods used for the operation and starting rods used for the starting of the reactor, and the starting rods are disposed in the radial periphery of the reactor core, while the control rods are disposed to the inside of the starting rods. With such a constitution, adjusting rods can be disposed in the region where the radial power peaking is generated to facilitate the flattening of the power distribution even in such a design that the ratio of the number of control rods to that of fuel assemblies is relatively large. That is, in this reactor, the radial power peaking is reduced by about 10% as compared with the conventional reactor core. As a result, the maximum linear power density during operation is reduced by about 10% to increase the thermal margin of the reactor core. If the maximum linear power density is set identical, the number of the fuel assemblies can be decreased by about 10%, to thereby reduce the fuel production cost. (K.M.)

  7. Issues in the selection of the LMFBR steam cycle

    International Nuclear Information System (INIS)

    Buschman, H.W.; McConnell, R.J.

    1983-01-01

    Unlike the light-water reactor, the liquid-metal fast breeder reactor (LMFBR) allows the designer considerable latitude in the selection of the steam cycle. This latitude in selection has been exercised by both foreign and domestic designers, and thus, despite the fact that over 25 LMFBR's have been built or are under construction, a consensus steam cycle has not yet evolved. This paper discusses the LMFBR steam cycles of interest to the LMFBR designer, reviews which of these cycles have been employed to date, discusses steam-cycle selection factors, discusses why a consensus has not evolved, and finally, concludes that the LMFBR steam-cycle selection is primarily one of technical philosophy with several options available

  8. Metallic Reactor Fuel Fabrication for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of (1) fuel slug casting, (2) loading and fabrication of the fuel rods, and (3) fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.

  9. Utility industry evaluation of the metal fuel facility and metal fuel performance for liquid metal reactors

    International Nuclear Information System (INIS)

    Burstein, S.; Gibbons, J.P.; High, M.D.; O'Boyle, D.R.; Pickens, T.A.; Pilmer, D.F.; Tomonto, J.R.; Weinberg, C.J.

    1990-02-01

    A team of utility industry representatives evaluated the liquid metal reactor metal fuel process and facility conceptual design being developed by Argonne National Laboratory (ANL) under Department of Energy sponsorship. The utility team concluded that a highly competent ANL team was making impressive progress in developing high performance advanced metal fuel and an economic processing and fabrication technology. The utility team concluded that the potential benefits of advanced metal fuel justified the development program, but that, at this early stage, there are considerable uncertainties in predicting the net overall economic benefit of metal fuel. Specific comments and recommendations are provided as a contribution towards enhancing the development program. 6 refs

  10. Performance of metallic fuels in liquid-metal fast reactors

    International Nuclear Information System (INIS)

    Seidel, B.R.; Walters, L.C.; Kittel, J.H.

    1984-01-01

    Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

  11. CEC activities in the field of LMFBR safety

    International Nuclear Information System (INIS)

    Balz, W.; Finzi, S.; Klersy, R.

    1976-01-01

    The aim of the ECC is to reach a common LMFBR Safety strategy in Europe. To this end the Commission promotes collaboration between the different fast reactor projects in the Community through working groups and collaborative arrangements and contributes with a research activity executed in its Joint Research Centre Ispra. A short description is given of the activity in the working groups and of the Ispra programme on LMFBR Safety. This programme covers: LMFBR thermohydraulics, fuel coolant interactions, dynamic structure loading and response, safety related material properties and whole core accident code development

  12. Pyroprocessing of IFR Metal Fuel

    International Nuclear Information System (INIS)

    Laidler, J.J.

    1993-01-01

    The Integral Fast Reactor (IFR) fuel cycle features the use of an innovative reprocessing method, known as open-quotes pyroprocessingclose quotes featuring fused-salt electrofining of the spent fuel. Electrofining of IFR spent fuel involves uranium recovery by electro-transport to a solid steel cathode. The thermodynamics of the system preclude plutonium recovery in the same way, so a liquid cadmium cathode located in the electrolyte salt phase is utilized. The deposition of Pu, Am, Np, and Cm takes place at the liquid cadmium cathode in the form of cadmium intermetallic compounds (e.g, PuCd 6 ), and uranium deposits as the pure metal when cadmium saturation is reached. A small amount of rare earth fission products deposit together with the heavy metals at both the solid and liquid cadmium cathodes, providing a significant degree of self-protection. A full scope demonstration of the IFR fuel cycle will begin in 1993, using fuel irradiated in EBR-II

  13. Irradiation behavior of metallic fast reactor fuels

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985

  14. In-pile observations of fuel and clad relocation during LMFBR initiation phase accident experiments - the STAR experiments

    International Nuclear Information System (INIS)

    Wright, S.A.; Schumacher, G.; Henkel, P.R.; Royl, P.

    1987-01-01

    A series of seven in-pile experiments (the STAR experiments) were performed in which clad motion and fuel dispersal were observed in small pin bundles with high-speed cinematography. The experimental heating conditions reproduced a range of Loss of Flow (LOF) accident scenarios for the lead subassemblies in LMFBRs. The experiments show strong tendencies for limited clad motion in multiple pin bundles, early fuel disruption and dispersal (prior to fuel melting) in moderate power transients having simultaneous clad melting and fuel disruption. The more recent experiments indicate a possibility of steel vapor driven fuel dispersal after fuel breakup and intimate fuel/steel mixing. (author)

  15. Improvements in fabrication of metallic fuels

    International Nuclear Information System (INIS)

    Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.

    1989-12-01

    Argonne National Laboratory is currently developing a new liquid- metal cooled breeder reactor known as the Integral Fast Reactor (IFR). IFR fuels represent the state-of-the-art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, to be discussed below, will support the fully remote fuel cycle facility that as an integral part of the IFR concept will be demonstrated at the EBR-II site. 3 refs

  16. A decade of advances in metallic fuel

    International Nuclear Information System (INIS)

    Seidel, B.R.; Batte, G.L.; Dodds, N.E.; Hofman, G.L.; Lahm, C.E.; Pahl, R.G.; Porter, D.L.; Tsai, H.; Walters, L.C.

    1990-01-01

    Significant advances in the understanding of behavior and performance of metallic fuels to high burnup have been achieved over the past four decades. Metallic fuels were the first fuels for liquid-metal-cooled fast reactors (LMR) but in the late 1960s worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved. Now metallic fuels are recognized as a preferred viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last decade and highlights the behavior and performance features which have demonstrated a much greater potential than previously expected. 28 refs., 2 figs., 1 tab

  17. Improvements in the fabrication of metallic fuels

    International Nuclear Information System (INIS)

    Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.

    1989-01-01

    Argonne National Laboratory (ANL) is currently developing a new liquid-metal-cooled breeder reactor known as the Integral Fast Reactor (IFR). The IFR represents the state of the art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, discussed in this paper, will support ANL-West's (ANL-W) fully remote fuel cycle facility, which is an integral part of the IFR concept

  18. Low Loss Advanced Metallic Fuel Casting Evaluation

    International Nuclear Information System (INIS)

    Kim, Kihwan; Ko, Youngmo; Kim, Jonghwan; Song, Hoon; Lee Chanbock

    2014-01-01

    The fabrication process for SFR fuel is composed of fuel slug casting, loading and fabrication of the fuel rods, and the fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycles streams in the fabrication process. Recycle streams include fuel slug reworks, returned scraps, and fuel casting heels, which are a special concern in the counter gravity injection casting process because of the large masses involved. Large recycle and waste streams result in lowering the productivity and the economic efficiency of fuel production. To increase efficiency the fuel losses in the furnace chamber, crucible, and the mold, after casting a considerable amount of fuel alloy in the casting furnace, will be quantitatively evaluated. After evaluation the losses will be identified and minimized. It is expected that this study will contribute to the minimization of fuel losses and the wastes streams in the fabrication process of the fuel slugs. Also through this study the technical readiness level of the metallic fuel fabrication process will be further enhanced. In this study, U-Zr alloy system fuel slugs were fabricated by a gravity casting method. Metallic fuel slugs were successfully fabricated with 19 slugs/batch with diameter of 5mm and length of 300mm. Fuel losses was quantitatively evaluated in casting process for the fuel slugs. Fuel losses of the fuel slugs were so low, 0.1∼1.0%. Injection casting experiments have been performed to reduce the fuel loss and improve the casting method. U-Zr fuel slug having φ5.4-L250mm was soundly fabricated with 0.1% in fuel loss. The fuel losses could be minimized to 0.1%, which showed that casting technology of fuel slugs can be a feasible approach to reach the goal of the fuel losses of 0.1% or less in commercial scale

  19. Equation of state for L.M.F.B.R. fuel (measurement of fission gas release during transients)

    International Nuclear Information System (INIS)

    Combette, P.; Barthelemy, P.

    1979-01-01

    A sample of fuel (UO 2 or UPuO 2 ) can be heated by fission in a heating transient up to energy deposition 4000 j/g, in the Silene reactor. The Kistler type capsule, the calorimeter device and the radiochemical analysis of fission products enable the pressure pulse and the fuel energy deposition to be measured. So, the relationship between the fuel vapour pressure and the fuel specific energy can be deduced. Peaks pressure (about 1 MPa) coming from fresh UO 2 vaporization, have been measured on a 7 milliseconds time scale. There is a good agreement with the E.O.S. for fresh UO 2 , which is well known for low pressure (1 MPa). Numerous tests have been done with 93% enriched UO 2 and a first test with highly active fuel containing plutonium (15 at %) has been performed. The capsule allows the released gas coming from the irradiated fuel to be retained for measurements and analysis. To investigate the mode of fuel disruption, in-pile fission-heated fuel pellets has been recorded by high speed cinematography

  20. Recent metal fuel safety tests in TREAT

    International Nuclear Information System (INIS)

    Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

    1986-01-01

    In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described

  1. Transactions of the second technical exchange meeting on fuel- and clad-motion diagnostics for LMFBR safety test facilities

    Energy Technology Data Exchange (ETDEWEB)

    DeVolpi, A. (comp.)

    1976-01-01

    Papers are presented which deal with diagnostic requirements and fuel motion monitoring capabilities of hodoscopes, coded aperture systems, x-ray radiography, and in-core detectors. Separate abstracts and indexing were prepared for each paper. (DG)

  2. Assessment of accident energetics in LMFBR core-disruptive accidents

    International Nuclear Information System (INIS)

    Fauske, H.K.

    1977-01-01

    An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggests that such events are highly unlikely following a postulated core meltdown event

  3. Reprocessing technology of liquid metal cooled fast breeder reactor fuel

    International Nuclear Information System (INIS)

    Baetsle, L.H.; Broothaerts, J.; Heylen, P.R.; Eschrich, H.; Geel, J. van

    1974-11-01

    All the important aspects of LMFBR fuel reprocessing are critically reviewed in this report. Storage and transportation techniques using sodium, inert gas, lead, molten salts and organic coolants are comparatively discussed in connection with cooling time and de-activation techniques. Decladding and fuel disaggregation of UO 2 -PuO 2 fuel are reviewed according to the present state of R and D in the main nuclear powers. Strong emphasis is put on on voloxidation, mechanical pulverization and molten salt disaggregation in connection with volatilization of gaseous fission products. Release of fission gases and the resulting off-gas treatment are discussed in connection with cooling time, burn up and dissagregation techniques. The review is limited to tritium, iodine xenon-krypton and radioactive airborne particulates. Dissolution, solvent extraction and plutonium purification problems specifically connected to LMFBR fuel are reviewed with emphasis on the differences between LWR and fast fuel reprocessing. Finally the categories of wastes produced by reprocessing are analysed according to their origin in the plant and their alpha emitters content. The suitable waste treatment techniques are discussed in connection with the nature of the wastes and the ultimate disposal technique. (author)

  4. Composition of the corrosion product oxide phases in the gaps of three LMFBR-type mixed oxide fuel pins

    International Nuclear Information System (INIS)

    Walker, C.T.

    1978-01-01

    Identification of the corrosion product oxide phases in the gaps of mixed oxide fuel pins clad with austenitic stainless steel is an important step in understanding the mechanisms of corrosion. Once the oxide is identified the reaction paths by which it is formed can be established. The corrosion product oxide will also be a reliable indication of the oxygen potential in the gap. A grey oxide phase which is recognised to contain the products of cladding corrosion is often observed in the gaps of mixed oxide fuel pins. Its main constituents are known to be chromium, caesium and oxygen, and recently the composition of the grey oxide phase in the gaps of three fuel pins was shown to be broadly similar, and it was proposed that the phase was essentially chromium oxide. (Auth.)

  5. LMFBR with booster pump in pumping loop

    International Nuclear Information System (INIS)

    Rubinstein, H.J.

    1975-01-01

    A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation

  6. LMFBR core materials

    International Nuclear Information System (INIS)

    Roberts, J.T.A

    1981-01-01

    This paper discusses initial fuel and core designs that are conservative, emphasizing high reliability at the expense of breeding ration and doubling time. The progression in design improvements to better doubling time, plutonium inventory requirements, and power density are shown. The recent progress in advanced oxide fuel, cladding, and duct development is reported. Breeding ratios and compound system doubling times are computed for both the conventional homogeneous and alternative heterogeneous core configurations. Results are summarized and it is shown that both carbide and metal cores can achieve breeding ratios in the range of 1.5 to 1.6 versus 1.45 for oxide; the carbide and metal cores can achieve 10 year doubling times, which translate into millions of dollars savings in fuel cycle costs. Carbide fuel development is being continued at a modest level to provide backup to the mixed oxide design

  7. Measurements of dynamic shape factors of LMFBR aggregate aerosols

    International Nuclear Information System (INIS)

    Allen, M.D.; Moss, O.R.; Briant, J.K.

    1980-01-01

    Dynamic shape factors for branched, chain-like aggregates of LMFBR mixed-oxide fuels have been measured with a LAPS spiral-duct centrifuge. The aerosol was generated by repeatedly pulsing a focused laser beam onto the surface of a typical LMFBR fuel pellet. The measured values of the dynamic shape factor, corrected for slip, vary between kappa = 3.60 at D/sub ae/ = 0.5 μm, and kappa = 2.23 at D/sub ae/ = 1.5 μm

  8. Welding development for LMFBR applications

    International Nuclear Information System (INIS)

    Slaughter, G.M.; Edmonds, D.P.; Goodwin, G.M.; King, J.F.; Moorhead, A.J.

    1976-01-01

    High-quality welds with suitable properties for long-time elevated-temperature nuclear service are among the most critical needs in today's welding technology. Safe, reliable, and economic generation of future power depends on welded construction in systems such as Liquid Metal Fast Breeder Reactors (LMFBRs). Rapid thermal transients in LMFBR systems at coolant temperatures around 590 to 650 0 C (1000 to 1200 0 F) could cause creep and creep-fatigue damage that is not encountered in lower temperature reactor systems. The undesirable consequences of interaction between the two working fluids - sodium and steam - in the steam generators are also of major concern. Thus sound welds that have excellent reliability over a 30-year service life are essential. Several programs are actively underway at ORNL to satisfy this critical need and selected portions of three of these programs are discussed briefly

  9. Basic analysis and a comparison of the characteristics GCFRs and the LMFBR with the thorium cycle in one-group diffusion theory

    International Nuclear Information System (INIS)

    Sabundjian, G.; Ishiguro, Y.

    1991-09-01

    A preliminary study of neutronics of thorium cycle fast breeder reactor has been done using simplified reactor models and analyses methods with the aim of finding a type of breeder reactor suitable for an efficient utilization of thorium that is abundant in Brazil. Basic methods of cross section processing and reactor calculation are studied and applied to analyse breeding characteristics of GCFRs and LMFBRs. The GCFR is fueled with oxide pins and cooled with helium. The LMFBR is fueled with thin metallic pins to achieve high power densities. Neutronics characteristics are determined as functions of the average power density and the fuel volume fraction. Results show that a high power density and a high fuel volume fraction are desirable to achieve short doubling times, that the GCFR is inferior to the LMFBR in regard to the doubling time and that the LMFBR can achieve reactor doubling times ten years with an average power density of ∼ 600MW/m 3 and fuel volume fraction of 40%. (author)

  10. User's guide to EPIC, a computer program to calculate the motion of fuel and coolant subsequent to pin failure in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Pizzica, P.A.; Garner, P.L.; Abramson, P.B.

    1979-10-01

    The computer code EPIC models fuel and coolant motion which results from internal fuel pin pressure (from fission gas or fuel vapor) and possibly from the generation of sodium vapor pressure in the coolant channel subsequent to pin failure in a liquid-metal fast breeder reactor. The EPIC model is restricted to conditions where fuel pin geometry is generally preserved and is not intended to treat the total disruption of the pin structure. The modeling includes the ejection of molten fuel from the pin into a coolant channel with any amount of voiding through a clad breach which may be of any length or which may extend with time. One-dimensional Eulerian hydrodynamics is used to treat the motion of fuel and fission gas inside a molten fuel cavity in the fuel pin as well as the mixture of two-phase sodium and fission gas in the coolant channel. Motion of fuel in the coolant channel is tracked with a type of particle-in-cell technique. EPIC is a Fortran-IV program requiring 400K bytes of storage on the IBM 370/195 computer. 21 refs., 2 figs.

  11. Hot clamp design for LMFBR piping systems

    International Nuclear Information System (INIS)

    Kobayashi, T.; Tateishi, M.

    1993-01-01

    Thin-wall, large-diameter piping for liquid metal fast breeder reactor (LMFBR) plants can be subjected to significant thermal transients during reactor scrams. To reduce local thermal stresses, an insulated cold clamp was designed for the fast flux test facility and was also applied to some prototype reactors thereafter. However, the cost minimization of LMFBR requires much simpler designs. This paper presents a hot clamp design concept, which uses standard clamp halves directly attached to the pipe surface leaving an initial gap. Combinations of flexible pipe and rigid clamp achieved a self-control effect on clamp-induced pipe stresses due to the initial gap. A 3-D contact and inelastic history analysis were performed to verify the hot clamp concept. Considerations to reduce the initial stress at installation, to mitigate the clamp restraint on the pipe expansion during thermal shocks, and to maintain the pipe-clamp stiffness desired during a seismic event were discussed

  12. Development of metallic fuels for fast reactors

    International Nuclear Information System (INIS)

    Singh, R.P.

    2011-01-01

    A very rapid growth rate of 100-fold increase in the next 50 years has been targeted for nuclear energy in India for which metallic fuel will be introduced in FBRs after 2020. Many innovative and challenging fuel design concepts for metal fuels are under consideration, which require extensive research and development work to generate database for the designers. The designs under active considerations are one in which the fuel will either be sodium bonded ternary U-15Pu-6Zr alloy or mechanically/sodium bonded binary U-15Pu alloy with a Zr liner between the fuel and the clad. The decision on the choice of the fuel will be based on test fuel irradiations in FBTR, subsequent PIE results, modeling studies and closing the fuel cycle through pyrometallurgical route. The development of metallic fuel is being pursued jointly by BARC and IGCAR. While the mechanical bonding concept with Pu-U alloy as fuel is being pursued at BARC, sodium bonding concept is being developed at IGCAR. BARC has already established the fabrication route for the mechanical bonding of U-Zr alloy, through co-swaging route. The properties of the fuels are evaluated. The fabrication of U-Pu alloy fuel pins through this route is now under development. It is now proposed to carry out the irradiation of test fuel pins, fabricated through both the routes, in FBTR in order to select the route for further development. This will be followed by subassembly level irradiation in FBTR to obtain experience on large scale fabrication of metallic fuel and to establish its performance

  13. LMFBR operational safety

    International Nuclear Information System (INIS)

    Sackett, J.I.; Allen, N.L.; Dean, E.M.; Fryer, R.M.; Larson, H.A.; Lehto, W.K.

    1978-01-01

    The mission of the Experimental Breeder Reactor II (EBR-II) has envolved from that of a small LMFBR demonstration plant to a major irradiation-test facility. Because of that evolution, many operational-safety issues have been encountered. This paper describes the EBR-II operational-safety experience in four areas: protection-system design, safety-document preparation, tests of off-normal reactor conditions, and tests of elements with breached cladding. (author)

  14. LMFBR: safety aspects

    International Nuclear Information System (INIS)

    Natta, M.

    1990-01-01

    This presentation of LMFBR safety is limited at Super Phenix reactor. After a brief description of the reactor, some details on safety systems, in normal or accidental conditions, are given. The main functions studied are: chain reaction trip, residual power evacuation, reactor containment. In heavy accident the behaviour of Super Phenix is studied which its particular characteristics and the possibilities of operators reactions. The probability of appearance and the maximum consequences of heavy accidents are given [fr

  15. Metallic uranium as fuel for fast reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de

    1988-01-01

    This paper presents a first overview of the use of metallic uranium and its alloys as an option for fuel for rapid reactors. Aspects are discussed concerning uranium alloys which present high solubility in the gamma phase. (author)

  16. Metal fuel manufacturing and irradiation performance

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Walters, L.C.

    1992-01-01

    The advances in metal fuel by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, and improved passive safety. The goals and the safety philosophy of the Integral Fast Reactor Program are stressed

  17. 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.

  18. Shielding plug for LMFBR type reactors

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1979-01-01

    Purpose: To enable effective removal of liquid metals deposited, if any, in the gaps between a rotary plug and a fixed plug in LMFBR type reactors. Constitution: A plate incorporated with a heater and capable of projecting in a gap between a rotary plug and a fixed plug, and a scraper connected in perpendicular to it are provided to the rotary plug. Solidified liquid metals such as sodium deposited in the gap are effectively removed by the heating with the heater and the scraping action due to the rotation. (Horiuchi, T.)

  19. Fission gas retention in irradiated metallic fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Gruber, E.E.; Kramer, J.M.

    1987-01-01

    Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5Fs) are presented. The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations to the measurements shows quantitative agreement with both the magnitude and the axial variation of the retained gas content

  20. Status of gamma-ray heating characterization in LMFBR

    International Nuclear Information System (INIS)

    Gold, R.

    1975-11-01

    Efforts to define gamma-ray heating in Liquid Metal Fast Breeder Reactor (LMFBR) environments have been surveyed. Emphasis is placed on both current practice for the Experimental Breeder Reactor-II (EBR-II) and future needs of the Fast Flux Test Facility (FFTF). Experimental and theoretical work are included in this preliminary survey for both high and low power environments. Current ''state-of-the-art'' accuracies and limitations are assessed. On this basis, it is concluded that a broad and sustained effort be initiated to meet requested FFTF goal accuracies. To this end, recommendations are advanced for improving the current status of gamma heating characterization and temperature measurements in LMFBR

  1. 33 CFR 183.562 - Metallic fuel lines.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Metallic fuel lines. 183.562...) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Fuel Systems Manufacturer Requirements § 183.562 Metallic fuel lines. (a) Each metallic fuel line that is mounted to the boat structure must be connected to the...

  2. Analytical throughput-estimating methods for the Hot Fuel Examination Facility

    International Nuclear Information System (INIS)

    Keyes, R.W.; Phipps, R.D.

    1983-01-01

    The Hot Fuel Examination Facility (HFEF) supports the operation and experimental programs of the major Liquid Metal Fast Breeder Reactor (LMFBR) test facilities; specifically, the Fast Flux Test Facility (FFTF), the Experimental Breeder Reactor II (EBR-II), and the Transient Reactor Test (TREAT) Facility. Successful management of HFEF and of LMFBR safety and fuels and materials programs, therefore, requires reliable information regarding HFEF's capability to handle expected or proposed program work loads. This paper describes the 10-step method that has been developed to consider all variables which significantly affect the HFEF examination throughput and quickly provide the necessary planning information

  3. Casting Development of Metallic Fuel for SFR

    International Nuclear Information System (INIS)

    Song, H.; Kim, J.H.; Ko, Y.M.; Woo, Y.M.; Kim, K.H.; Lee, C.B.

    2015-01-01

    U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D programme of Korea. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, the generation of long-lived radioactive wastes and a loss of volatile species should be minimised during the recycled fuel fabrication step. In this study, fuel slug fabrication method has been introduced to develop an innovative fabrication process of metal fuel of SFR for preventing the evaporation of volatile elements such as Am. Metal fuel slugs were fabricated with an improved injection casting method in KAERI. Volatile species can be retained through the use of a cover gas with over pressure and covered crucibles. Experimental results show that the Mn was not volatilized and conserved in inert gas conditions compared to the vacuum condition. The volatility of Mn can be controlled by changing the casting process, and minimal Mn (and Am) loss is possible. An improved casting method under an inert atmosphere is more effective in the prevention of vaporisation than casting under a vacuum and reduced atmosphere. In addition, improved casting under a reduced atmosphere shows a considerable effect in the prevention of vaporisation. (authors)

  4. Advanced breeder cycle uses metallic fuel

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1991-01-01

    Scientists from Argonne National Laboratory have been developing a concept called the Integral fast Reactor (IFR). This fast breeder reactor could effectively increase Uranium resources a hundred fold making nuclear power essentially an inexhaustible energy source. The IFR is outlined. In the IFR, the inherent properties of liquid metal cooling are combined with a new metallic fuel which is allowed to swell and gives an improved burnup level and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics and waste management. (author)

  5. Non-noble metal fuel cell catalysts

    CERN Document Server

    Chen, Zhongwei; Zhang, Jiujun

    2014-01-01

    Written and edited by a group of top scientists and engineers in the field of fuel cell catalysts from both industry and academia, this book provides a complete overview of this hot topic. It covers the synthesis, characterization, activity validation and modeling of different non-noble metal and metalfree electrocatalysts for the reduction of oxygen, as well as their integration into acid or alkaline polymer exchange membrane (PEM) fuel cells and their performance validation, while also discussing those factors that will drive fuel cell commercialization. With its well-structured app

  6. Analysis of a postulated accident scenario involving loss of forced flow in a LMFBR

    International Nuclear Information System (INIS)

    Moreira, M.L.

    1985-01-01

    A model to analyse a postulated accident scenario involving loss of forced flow in the reactor vessel of a LMFBR is used. Five phases of the accident are analysed: Natural Circulation, Subcooled Boiling, Nucleate Boiling, Core Dryout and Cladding melt. The heat conduction in the fuel cladding, coolant and lower and upper plenum are calculated by a lump-parameter model. Physical data of a prototype LMFBR reactor were used for the calculation. (author)

  7. LMFBR design and its evolution. (2) Core design of LMFBR

    International Nuclear Information System (INIS)

    Uto, Nariaki; Mizuno, Tomoyasu

    2003-01-01

    Sodium-cooled core design studies are performed. MOX fuel core with axial blanket partial elimination subassembly due to safety consideration is studied. This type of core with high internal conversion ratio possesses capability of achieving 26 months of operation cycle length and 100 GWd/t of burnup averaged over core and blanket, which are superior characteristics in view of reducing cost of power generation. Metal fuel core is also studied, and its higher breeding capability reveals a potential of better core performance such as longer operation cycle length for the same level of electricity generation, though core outlet temperature is limited to lower level due to steel cladding-metal fuel compatibility concerns. Another metal fuel core concept using single Pu enrichment and two radial regions with individual fuel pin diameters achieves 550degC of core outlet temperature identical to that of MOX fuel core, keeping operation cycle length comparable with that of MOX fuel core. This series of study results show that sodium-cooled MOX and metal fuel cores have a high flexibility in satisfying various needs including fuel cycle cost and breeding capability, depending on the stage of introducing commercialized fast reactor cycle system. (author)

  8. Studies of Lanthanide Transport in Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Jinsuo; Taylor, Christopher

    2018-04-02

    Metallic nuclear fuels were tested in fast reactor programs and performed well. However, metallic fuels have shown the phenomenon of FCCI that are due to deleterious reactions between lanthanide fission products and cladding material. As the burnup is increased, lanthanide fission products that contact with the cladding could react with cladding constituents such as iron and chrome. These reactions produce higher-melting intermetallic compounds and low-melting alloys, and weaken the mechanical integrity. The lanthanide interaction with clad in metallic fuels is recognized as a long-term, high-burnup cause of the clad failures. Therefore, one of the key concerns of using metallic fuels is the redistribution of lanthanide fission products and migration to the fuel surface. It is believed that lanthanide migration is in part due to the thermal gradient between the center and the fuel-cladding interface, but also largely in part due to the low solubility of lanthanides within the uranium-based metal fuel. PIE of EBR-II fuels shows that lanthanides precipitate directly and do not dissolve to an appreciable extent in the fuel matrix. Based on the PIE data from EBR-II, a recent study recommended a so-called “liquid-like” transport mechanism for lanthanides and certain other species. The liquid-like transport model readily accounts for redistribution of Ln, noble metal fission products, and cladding components in the fuel matrix. According to the novel mechanism, fission products can transport as solutes in liquid metals, such as liquid cesium or liquid cesium–sodium, and on pore surfaces and fracture surfaces for metals near their melting temperatures. Transport in such solutions is expected to be much more rapid than solid-state diffusion. The mechanism could explain the Ln migration to the fuel slug peripheral surface and their deposition with a sludge-like form. Lanthanides have high solubility in liquid cesium but have low solubility in liquid sodium. As a

  9. Design study on metal fuel FBR cores

    International Nuclear Information System (INIS)

    Yokoo, T.; Tanaka, Y.; Ogata, T.

    1991-01-01

    A design approach for metal fuel FBR core to maintain fuel integrity during transient events by limiting eutectic/liquid phase formation is proposed based on the current status of metallic fuel development. Its impact as the limitation on the core outlet temperature is assessed through its application to two of CRIEPI's core concepts, high linear power 1000 MWe homogeneous design and medium linear power 300 MWe radially heterogeneous design. SESAME/SALT code is used in this study to analyze steady state and transient fuel behavior. SE2-FA code is developed based on SUPERENERGY-2 and used to analyze core thermal-hydraulics with uncertainties. As the result, the core outlet temperatures of both designs are found to be limited to ≤500degC if it is required to prevent eutectic/liquid phase formation during operational transients in order to guarantee the fuel integrity. Additional assessment is made assuming an advanced limiting condition that allows small liquid phase formation based on the liquid phase penetration rate derived from existing experimental results. The result indicates possibility of raising core outlet temperature to ∼ 530degC. Also, it is found that core design technology improvements such as hot spot factors reduction can contribute to the core outlet temperature extension by 10 ∼ 20degC. (author)

  10. Development of metallic fuel materials

    International Nuclear Information System (INIS)

    Kang, Young Ho; Lee, Chong Tak; Yang, Yeoung Seok; Kim, Ki Hwan; Hwang, Sung Chan; Joo, Keun Sik; Ann, Hyun Suk; Chang, Sae Jung.

    1997-09-01

    Through the control of melting and casting parameters, the sound and homogenous U-10wt.%Zr alloy could be fabricated. The yield and segregation of Zr elements were 85% and ±0.1wt.%, and the density of the alloy was about 16.6 g/cm 3 . The major phase were α-U and δ-UZr 2 . The microstructure showed the laminar structure with fiber morphology which was arranged alternatively with uranium and Zr-rich phase. This alloy will be used for KALIMER fuel material through developing the fabrication technology and the characteristics analysis. And electrorefining study was performed to separate uranium from uranium-neodymium and uranium-zirconium alloy by their different free energy for chloride formation. The liquid cadmium phase becomes the anode of the electrorefining cell. Uranium is electrolytically transported through a molten salt electrolyte to a low carbon steel cathode. The electrolyte is composed of KCl-LiCl eutectic and some UCl 3 , which are installed in the salt to facilitate the electrotransport of uranium. In pyrochemical process the reaction condition of chlorination and the maintenance its purity in preparing UCl 4 by chlorination of UO 2 is strongly dependent on the reaction temperature and time. (author).52 refs., 40 tabs., 129 figs

  11. Status of U.S. LMFBR programme

    International Nuclear Information System (INIS)

    Yevich, J.

    1978-01-01

    The determents of the decision for deterrence of commercial reprocessing and further demonstration of the plutonium breeder were based on two premises: time is needed to establish the programme for non-proliferating fuel cycle and there is a lessened sense of urgency for the USA to establish a commercial breeder in the near future. A strong, well funded base technology effort remains and will continue until institutional and technical solutions can be found to minimize or eliminate the proliferation risk. An LMFBR option will be maintained. The FFTF will be coming on line providing a powerful tool in breeder fuel and materials development and a baseline from which to scale up heat transfer systems and components. Sodium system hardware development and testing will continue to have high priority

  12. Treatment of Corroded Metallic Uranium Fuel

    International Nuclear Information System (INIS)

    Stridsman, H.; Ekeroth, E.; Hallberg, B.; Hellsten, E.; Lindberg, M.; Nordlinder, S.

    2009-01-01

    This paper describes the extensive planning and treatment of corroded metallic uranium fuel performed in Studsvik in 2007. This included conversion of possible pyrophoric uranium hydride to uranium oxide and separation of intact parts of fuel rods from corrosion products. The first nuclear reactor in Sweden was the R1 reactor in Stockholm, with fuel of natural metallic uranium. After shut-down of the reactor, part of the fuel had been placed in waterproof canisters in an interim storage pool at Studsvik. In 1988, corrosion of the fuel had been discovered in one of the canisters due to water leakage. Exposure of metallic uranium to water causes pyrophoric uranium hydride to be formed, which poses a severe risk of fire or explosion if subjected to oxygen in air. The canister with corroded fuel had then been placed in an outer container to which inert gas with low oxygen concentration was supplied, in order to slowly oxidize hydride. A project to take care of the corroded fuel was set up in 2006, comprising several preparatory steps: research, equipment design, pre-treatment and safety-case demonstrations. Treatment comprised two stages, pre-treatment at the storage location, and final treatment in a hot-cell. The pre-treatment stage was performed in order to study the reaction inside the container with a controlled oxygen supply and regularly control of atmosphere inside the container. In addition, equipment was developed for controlled movement of the container. The results from the pre-treatment stage were used for the planning and acceptance of the transport to and handling in the hot-cell facility in Studsvik. The work performed in the hot-cell consisted of opening the container, sequentially cutting the container and canister and separation of intact parts of fuel rods from the corrosion products and cladding, top and bottoms of aluminium. No evidence of uranium hydride was found throughout the operation. The corrosion products was expected to be in form of

  13. Breeding nuclear fuels with accelerators: replacement for breeder reactors

    International Nuclear Information System (INIS)

    Grand, P.; Takahashi, H.

    1984-01-01

    One application of high energy particle accelerators has been, and still is, the production of nuclear fuel for the nuclear energy industry; tantalizing because it would create a whole new industry. This approach to producing fissile from fertile material was first considered in the early 1950's in the context of the nuclear weapons program. A considerable development effort was expended before discovery of uranium ore in New Mexico put an end to the project. Later, US commitment to the Liquid Metal Fast Breeder Reactors (LMFBR) killed any further interest in pursuing accelerator breeder technology. Interest in the application of accelerators to breed nuclear fuels, and possibly burn nuclear wastes, revived in the late 1970's, when the LMFBR came under attack during the Carter administration. This period gave the opportunity to revisit the concept in view of the present state of the technology. This evaluation and the extensive calculational modeling of target designs that have been carried out are promising. In fact, a nuclear fuel cycle of Light Water Reactors and Accelerator Breeders is competitive to that of the LMFBR. At this time, however, the relative abundance of uranium reserves vs electricity demand and projected growth rate render this study purely academic. It will be for the next generation of accelerator builders to demonstate the competitiveness of this technology versus that of other nuclear fuel cycles, such as LMFBR's or Fusion Hybrid systems. 22 references, 1 figure, 5 tables

  14. 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

  15. Methods used and kind of results obtained in SCARABEE facility about fuel and clad motion diagnostics. Previsional development for SCARABEE N and CABRI facilities

    International Nuclear Information System (INIS)

    Bardy, J.; Manent, G.; Roche, L.; Tattegrain, A.

    1975-01-01

    Techniques are described for planned LMFBR loss of flow simulation studies using irradiated fuel (SCARABEE N program) and for LMFBR transient overpower studies using both fresh and irradiated fuel (CABRI program)

  16. LMFBR type reactor

    International Nuclear Information System (INIS)

    Kumaoka, Yoshio; Kawamura, Yutaka.

    1990-01-01

    A main vessel support skirt is supported by a base mat of reactor buildings and the base mat is supported by means of an earthquake-proof device on concretes of a lower raft disposed on ground rocks. The earthquake-proof device is constituted by alternately stacking, laminating and press-bonding thin steel plates together with thin rubber layers between an upper flange and a lower raft. Thus, for the horizontal seismic vibrations, the period of the swinging in the horizontal direction of the buildings is made greater than the swinging period of earthquakes by the earthquake-proof device to reduce the impact shocks of earthquakes. Further, for the vertical seismic vibrations, the input seismic movements are not amplified during transmission from the base mat of the buildings to the reactor structure by way of the support skirt of the main vessel, due to the shortened load transmission path and the seismic power design to the reactor structure can be moderated sufficiently. A safety LMFBR type reactor with reduced construction cost and improved reliability can be attained. (N.H.)

  17. Nuclear welding, application for an LMFBR

    International Nuclear Information System (INIS)

    Patriarca, P.; Goodwin, G.M.

    1975-01-01

    Fabrication of an LMFBR system is discussed, with emphasis on areas where joint welding innovations have been introduced. Each major component of the system, including reactor vessel, intermediate heat exchanger, steam generator, and sodium-containment piping, is treated separately. Developmet of special filler metals to avoid the low elevated-temperature creep ductility obtained with conventional austenitic stainless steel weldments is reported. Bore-side welding of steam generator tube-to-tubesheet joints with and without filler metal is desirable to improve inspectability and eliminate the crevice inherent with face-side weld design, thus minimizing corrosion problems. Automated welding methods for sodium-containment piping are summarized which iminimize and control distortion and ensure welds of high integrity. Selection of materials for the various components is discussed for plants presently under construction, and materials predictions are made for future concepts. (U.S.)

  18. Fast Flux Test Facility metal fuel pin fabrication

    International Nuclear Information System (INIS)

    Benecke, M.W.; Dittmer, J.O.; Feigenbutz, L.V.

    1989-05-01

    A new initiative to develop, irradiate, and qualify a binary uranium/zirconium metal fuel system in the Fast Flux Test Facility (FFTF) has been implemented by the Westinghouse Hanford Company for the US Department of Energy. Metal fuel test assemblies have been designed and fabricated and are now being irradiated in FFTF to provide the data needed to support the potential use of binary fuels in FFTF and other liquid metal reactors. These development efforts support licensing activities for metal fuel use in near-term advanced liquid metal reactors. New metal fuel pin design features, fabrication development, and fabrication processes for three metal fuel tests will be described and their irradiation status reported in this paper. 11 figs

  19. Quality assurance of metallic fuel pins using Eddy current technique

    International Nuclear Information System (INIS)

    Sasi, B.; Thirunavukkarasu, S.; Rao, B.P.C.; Jayakumar, T.; Baldev Raj

    2009-01-01

    Metallic fuels (U-Pu-Zr) is promising for future fast reactors in view of their higher breeding ratio and burn-up as compared to oxide, carbide and other ceramic fuels. The metallic fuel pins are made of T91 tubes filled with fuel slugs with sodium as bonding material. Quality assurance of metallic fuel pins using non destructive evaluation (NDE) techniques is important. For this, eddy current (EC) technique is a natural choice in view of its sensitivity, speed, versatility and ease of use. NDE of metallic fuel pins comprises of detection of defects in T91 tubes, defects in metallic fuel slugs (2.84 mm diameter) and voids in sodium. For these inspection requirements, separate EC techniques would be employed and in this direction, finite element (FE) modelling has been performed and experimental simulation studies have been carried out. This paper discusses the results of these studies

  20. Miscibility and oxidation rate of the simulated metallic spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    You, K. S.; Joo, J. S.; Shin, Y. J.; Oh, S. C. [KAERI, Taejon (Korea, Republic of)

    1999-10-01

    The simulated metallic spent fuel was fabricated by using Uranium, Neodymium and Palladium in order to study the miscibility of Neodymium and Palladium with Uranium. For analysis of long-term safty on the metallized spent fuel, the simulated metallic spent fuel was oxidized under pure oxygen environment at 183{approx}250 deg C. From the results, the oxidation rate correlation and activation energy were obtained.

  1. Hot clamp design for LMFBR piping systems

    International Nuclear Information System (INIS)

    Kobayashi, T.; Tateishi, M.

    1990-01-01

    Thin wall large diameter piping for liquid metal fast breeder reactor plants can be subjected to significant thermal transients during reactor scrams. To reduce local thermal stresses, an insulted cold clamp was designed for the Fast Flux Test Facility and was also applied to some following prototype reactors. However, the cost minimization of LMFBR requires much simpler designs. This paper presents a hot clamp design concept, which uses standard clamp halves directly attached to the pipe surface with an initial gap. Combinations of flexible pipe and rigid clamp achieved a self-control effect on clamp induced pipe stresses due to the initial gap. A 3-D contact and inelastic history analysis were performed to verify the hot clamp concept. Considerations to reduce the initial stress at installation, to mitigate the clamp restraint on the pipe expansion during thermal shocks, and to maintain the pipe-clamp stiffness desired during a seismic event are discussed

  2. Recent progress in the development of metallic fuel

    International Nuclear Information System (INIS)

    Seidel, B.R.; Batte, G.L.; Dodds, N.E.; Lahm, C.E.; Pahl, R.G.; Tsai, H.C.

    1990-01-01

    Tests to date demonstrate that metallic fuel for advanced liquid metal reactors performs well, is easily reprocessed and refabricated and provides inherent reactor safety within an economic design. The behavior and performance of metallic fuel is key to the demonstration of the Integral Fast Reactor (IFR) concept at Argonne National Laboratory. Since 1985, more than 40 assemblies of experimental fuel in addition to the standard metallic driver fuel for Experimental Breeder Reactor 2 (EBR-2)have been irradiated; several more continue to be designed and fabricated. Results have characterized the influence of a wide range of fabrication, design and material variables upon irradiation behavior throughout the fuel lifetime under normal and upset conditions including operation with breached cladding. Results of test, both in- and out-of-reactor, indicate that metallic fuel is readily and economically fabricated, capable of achieving high exposure and long reactor residence times, and possesses unique and promising safety features. 9 refs., 6 figs

  3. Advances in Metallic Fuels for High Burnup and Actinide Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.

    2016-10-01

    Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.

  4. 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

  5. Influence of fission product transport on delayed neutron precursors and decay heat sources in LMFBR accidents

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.

    1981-01-01

    A method is presented for studying the influence of fission product transpot on delayed neutron precursors and decay heat sources during Liquid Metal Fast Breeder Reactor (LMFBR) unprotected accidents. The model represents the LMFBR core as a closed homogeneous cell. Thermodynamic phase equilibrium theory is used to predict fission product mobility. Reactor kinetics behavior is analyzed by an extension of point kinetics theory. Group dependent delayed neutron precursor and decay heat source retention factors, which represent the fraction of each group retained in the fuel, are developed to link the kinetics and thermodynamics analysis. Application of the method to a highly simplified model of an unprotected loss-of-flow accident shows a time delay on the order of 10 ms is introduced in the predisassembly power history if fission product motion is considered when compared to the traditional transient solution. The post-transient influence of fission product transport calculated by the present model is a 24 percent reduction in the decay heat level in the fuel material which is similar to traditional approximations. Isotopes of the noble gases, Kr and Xe, and the elements I and Br are shown to be very mobile and are responsible for a major part of the observed effects. Isotopes of the elements Cs, Se, Rb, and Te were found to be moderately mobile and contribute to a lesser extent to the observed phenomena. These results obtained from the application of the described model confirm the initial hypothesis that sufficient fission product transport can occur to influence a transient. For these reasons, it is concluded that extension of this model into a multi-cell transient analysis code is warranted

  6. Fabrication of metallic channel-containing UO2 fuels

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Song, Kun Woo; Kim, Keon Sik; Jung, Youn Ho

    2004-01-01

    The uranium dioxide is widely used as a fuel material in the nuclear industry, owing to many advantages. But it has a disadvantage of having the lowest thermal conductivity of all kinds of nuclear fuels; metal, carbide, nitride. It is well known that the thermal conductivity of UO 2 fuel is enhanced by making, so called, the CERMET (ceramic-metal) composite which consists of both continuous body of highly thermal-conducting metal and UO 2 islands. The CERMET fuel fabrication technique needs metal phase of at least 30%, mostly more than 50%, of the volume of the pellet in order to keep the metal phase interconnected. This high volume fraction of metal requires such a high enrichment of U that the parasitic effect of metal should be compensated. Therefore, it is attractive to develop an innovative composite fuel that can form continuous metal phase with a small amount of metal. In this investigation, a feasibility study was made on how to make such an innovative fuel. Candidate metals (W, Mo, Cr) were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO 2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet

  7. A state of the art on metallic fuel technology development

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Kang, Hee Young; Nam, Cheol; Kim, Jong Oh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Since worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved in the late 1960`s, the development of metallic fuels continued throughout the 1970`s at ANL`s experimental breeder reactor II (EBR-II) because EBR-II continued to be fueled with the metallic uranium-fissium alloy, U-5Fs. During this decade the performance limitations of metallic fuel were satisfactorily resolved resolved at EBR-II. The concept of the IFR developed at ANL since 1984. The technical feasibility had been demonstrated and the technology database had been established to support its practicality. One key features of the IFR is that the fuel is metallic, which brings pronounced benefits over oxide in improved inherent safety and lower processing costs. At the outset of the 1980`s, it appeared that metallic fuels are recognized as a professed viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last score and summarizes the state-of the art on metallic fuel technology development. (author). 29 refs., 1 tab.

  8. A state of the art on metallic fuel technology development

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang, Hee Young; Nam, Cheol; Kim, Jong Oh

    1997-01-01

    Since worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved in the late 1960's, the development of metallic fuels continued throughout the 1970's at ANL's experimental breeder reactor II (EBR-II) because EBR-II continued to be fueled with the metallic uranium-fissium alloy, U-5Fs. During this decade the performance limitations of metallic fuel were satisfactorily resolved resolved at EBR-II. The concept of the IFR developed at ANL since 1984. The technical feasibility had been demonstrated and the technology database had been established to support its practicality. One key features of the IFR is that the fuel is metallic, which brings pronounced benefits over oxide in improved inherent safety and lower processing costs. At the outset of the 1980's, it appeared that metallic fuels are recognized as a professed viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last score and summarizes the state-of the art on metallic fuel technology development. (author). 29 refs., 1 tab

  9. Viscosity Meaurement Technique for Metal Fuels

    International Nuclear Information System (INIS)

    Ban, Heng

    2015-01-01

    Metallic fuels have exceptional transient behavior, excellent thermal conductivity, and a more straightforward reprocessing path, which does not separate out pure plutonium from the process stream. Fabrication of fuel containing minor actinides and rare earth (RE) elements for irradiation tests, for instance, U-20Pu-3Am-2Np-1.0RE-15Zr samples at the Idaho National Laboratory, is generally done by melt casting in an inert atmosphere. For the design of a casting system and further scale up development, computational modeling of the casting process is needed to provide information on melt flow and solidification for process optimization. Therefore, there is a need for melt viscosity data, the most important melt property that controls the melt flow. The goal of the project was to develop a measurement technique that uses fully sealed melt sample with no Americium vapor loss to determine the viscosity of metallic melts and at temperatures relevant to the casting process. The specific objectives of the project were to: develop mathematical models to establish the principle of the measurement method, design and build a viscosity measurement prototype system based on the established principle, and calibrate the system and quantify the uncertainty range. The result of the project indicates that the oscillation cup technique is applicable for melt viscosity measurement. Detailed mathematical models of innovative sample ampoule designs were developed to not only determine melt viscosity, but also melt density under certain designs. Measurement uncertainties were analyzed and quantified. The result of this project can be used as the initial step toward the eventual goal of establishing a viscosity measurement system for radioactive melts.

  10. Viscosity Meaurement Technique for Metal Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Ban, Heng [Utah State Univ., Logan, UT (United States). Mechanical and Aerospace Engineering; Kennedy, Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-09

    Metallic fuels have exceptional transient behavior, excellent thermal conductivity, and a more straightforward reprocessing path, which does not separate out pure plutonium from the process stream. Fabrication of fuel containing minor actinides and rare earth (RE) elements for irradiation tests, for instance, U-20Pu-3Am-2Np-1.0RE-15Zr samples at the Idaho National Laboratory, is generally done by melt casting in an inert atmosphere. For the design of a casting system and further scale up development, computational modeling of the casting process is needed to provide information on melt flow and solidification for process optimization. Therefore, there is a need for melt viscosity data, the most important melt property that controls the melt flow. The goal of the project was to develop a measurement technique that uses fully sealed melt sample with no Americium vapor loss to determine the viscosity of metallic melts and at temperatures relevant to the casting process. The specific objectives of the project were to: develop mathematical models to establish the principle of the measurement method, design and build a viscosity measurement prototype system based on the established principle, and calibrate the system and quantify the uncertainty range. The result of the project indicates that the oscillation cup technique is applicable for melt viscosity measurement. Detailed mathematical models of innovative sample ampoule designs were developed to not only determine melt viscosity, but also melt density under certain designs. Measurement uncertainties were analyzed and quantified. The result of this project can be used as the initial step toward the eventual goal of establishing a viscosity measurement system for radioactive melts.

  11. A review of liquid metal anode solid oxide fuel cells

    Directory of Open Access Journals (Sweden)

    ALIYA TOLEUOVA

    2013-06-01

    Full Text Available This review discusses recent advances in a solid oxide fuel cell (SOFC variant that uses liquid metal electrodes (anodes with the advantage of greater fuel tolerance and the ability to operate on solid fuel. Key features of the approach are discussed along with the technological and research challenges that need to be overcome for scale-up and commercialisation.

  12. Fuel exchanger in FBR type reactor

    International Nuclear Information System (INIS)

    Shinden, Kazuhiko; Tanaka, Osamu.

    1990-01-01

    The present invention concerns a fuel exchanger for exchanging fuels in an LMFBR type reactor using liquid metals as coolants. An outer gripper cylinder rotating device for rotating an outer gripper cylinder that holds a gripper is driven, to lower the gripper driving portion and the outer gripper cylinder, fuels are caught by the finger at the top end of the outer gripper cylinder and elevated to extract the fuels from the reactor core. Then, the gripper driving portion casing and the outer gripper cylinder are rotated to rotate the fuels caught by the gripper. Subsequently, the gripper driving portion and the outer gripper cylinder are lowered to charge the fuels in the reactor core. This can directly shuffle the fuels in the reactor core without once transferring the fuels into a reactor storing pot and replacing with other fuels, thereby shortening the shuffling time. (I.N.)

  13. LARA: Expert system for acoustic localization of robot in a LMFBR

    International Nuclear Information System (INIS)

    Lhuillier, C.; Malvache, P.

    1986-12-01

    The expert system LARA (Acoustic Localization of Autonomic Robot) has been developed to show the interest of introducing artificial intelligency for fine automatic positioning of refuelling machine in a LMFBR reactor. LARA which is equipped with an acoustic detector gives rapidly a good positioning on the fuel [fr

  14. Design of metallic bipolar plates for PEM fuel cells.

    Science.gov (United States)

    2012-01-01

    This project focused on the design and production of metallic bipolar plates for use in PEM fuel cells. Different metals were explored : and stainless steel was found out to be best suited to our purpose. Following the selection of metal, it was calc...

  15. 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.

  16. Fission and corrosion products behavior in primary circuits of LMFBR's

    International Nuclear Information System (INIS)

    Feuerstein, H.; Thorley, A.W.

    1987-08-01

    Most of the 20 presented papers report items belonging to more than one session. The equipment results of primary circuits of LMFBR's relative to corrosion and fission products, release and chemistry of fuel, measurement techniques and analytical procedures of sodium sampling, difficulties with radionuclides and particles, reactor experiences with EBR-II, FFTF, BR10, BOR60, BN350, BN600, JOYO, and KNK-II, DFR, PFR, RAPSODIE, PHENIX, and SUPERPHENIX, and at least the verification of codes for calculation models of radioactive products accumulation and distribution are described. All 20 papers presented at the meeting are separately indexed in the database. (DG)

  17. Performance of HT9 clad metallic fuel at high temperature

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Hayes, S.L.

    1992-01-01

    Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area

  18. Dimensional, microstructural and compositional stability of metal fuels

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, A.A.; Dayananda, M.A.

    1993-03-15

    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.

  19. 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

  20. Advancing liquid metal reactor technology with nitride fuels

    International Nuclear Information System (INIS)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.; Matthews, R.B.

    1991-08-01

    A review of the use of nitride fuels in liquid metal fast reactors is presented. Past studies indicate that both uranium nitride and uranium/plutonium nitride possess characteristics that may offer enhanced performance, particularly in the area of passive safety. To further quantify these effects, the analysis of a mixed-nitride fuel system utilizing the geometry and power level of the US Advanced Liquid Metal Reactor as a reference is described. 18 refs., 2 figs., 2 tabs

  1. Behavior of metallic fuel in treat transient overpower tests

    International Nuclear Information System (INIS)

    Bauer, T.H.; Wright, A.E.; Robinson, W.R.; Klickman, A.E.

    1988-01-01

    Results and analyses are reported for TREAT in-pile transient overpower tests of margin to cladding failure and pre-failure axial expansion of metallic fuel. In all cases the power rise was exponential on an 8 s period until either incipient or actual cladding failure was achieved. Test fuel included EBR-II driver fuel and ternary alloy, the reference fuel of the Intergral Fast Reactor concept. Test pin burnup spanned the widest range available. The nature of the observed cladding failure and resultant fuel dispersals is described. Simple models are presented which describe observed cladding failures and pre-failure axial expansions yet are general enough to apply to all metal fuel types

  2. Fuel transfer manipulator for liquid metal nuclear reactors

    International Nuclear Information System (INIS)

    Sturges, R.H.

    1983-01-01

    A manipulator for transferring fuel assemblies between inclined fuel chutes of a liquid metal nuclear reactor installation. Hoisting means are mounted on a mount supported by beams pivotably attached by pins to the mount and to the floor in such a manner that pivoting of the beams causes movement and tilting of a hoist tube between positions of alignment with the inclined chutes. (author)

  3. 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

  4. LMFBR steam generator leak detection development in the United States

    International Nuclear Information System (INIS)

    Magee, P.M.; Gerrels, E.E.; Greene, D.A.; McKee, J.

    1978-01-01

    Leak detection for Liquid Metal Fast Breeder Reactor steam generators is an important economic factor in the shutdown, repair and restart of a plant. Development of leak detection systems in the U.S. has concentrated on four areas: (1) chemical (H 2 and O 2 ) leak detection meters; (2) acoustic leak detection/location techniques; (3) investigation of leak behavior (enlargement, damage effects, plugging and unplugging); and (4) data management for plant operations. This paper discusses the status, design aspects, and applications of leak detection technology for LMFBR plants. (author)

  5. Experiences in the fabrication of aluminium clad metallic uranium fuel

    International Nuclear Information System (INIS)

    Vijayaraghavan, R.

    1989-01-01

    With a view to achieve self-sufficiency and self-reliance in the fabrication of metallic natural uranium fuel, a full fledged fuel fabrication facility was set up in 1958. Based on the then technical information available and the development work carried out, a flow-sheet for the fabrication of metallic uranium fuel, starting from uranium ingots, was worked out and the first fuel element was successfully fabricated in June 1959. More than half the first charge for the initial criticality of CIRUS, a 40 MWt research reactor at Trombay, was fabricated and supplied. Since then, this facility has been regularly catering to the replacement fuel and component requirements for CIRUS. The fuel for Dhruva, a 100 MWt research reactor at Trombay, is in the form of a cluster consisting of 7 fuel pins as compared to the rigid single fuel element for CIRUS. The fabrication process chosen for making this fuel is more or less on the same lines as that has been followed for CIRUS fuel. However, because of the smaller diameter of uranium metal rod, higher length to diameter ratio, configurations of fins on aluminium sheaths, cluster assembly etc., extensive development work was required to be undertaken for optimising various production parameters. Several prototype fuel clusters of different designs were made and subjected to rigorous out-of-pile and in-pile testing. Based on the reliable satisfactory performance, type II-B SPT cluster design was finally frozen for production. Stringent quality control is of prime importance in ensuring good performance of the fuel in the reactor. Hence, appropriate quality control measures have been adopted at various stages of fuel fabrication to ensure conformance with the specifications. (author) 9 refs., 11 figs., 1 tab

  6. Conceptual design of KALIMER uranium metallic fueled core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young In; Kim, Sang Ji; Kim, Young Gyun; Kim, Young Jin

    1999-03-15

    As a part of the core design development of KALIMER(150 MWe), the KALIMER core design which uses U-Zr binary fuel not in excess of 20% enrichment was performed. Starting from the former uranium metallic fueled core design, a more economic and safer equilibrium core design was first established based on extensive researches for the possible enrichment gains over various design options and in-core fuel management strategies. Further optimization to extend fuel discharge burnup has been achieved by employing strategic loading schemes for initial and transition cycles to reach the equilibrium cycle early. The core performance analysis based on a once-through equilibrium fuel cycle scenario shows that the core has an average breeding ratio of 0.67 and core average discharge burnup of 61.6 MWD/kg. The negative sodium void reactivity over the core shows a beneficial potential to assure inherent safety characteristics. When comparing with conventional plutonium metallic fueled cores of the same power level, the present KALIMER uranium metallic fueled core has an increased physical core size to meet the enrichment restriction, and, as a result, a lower power density to realize the minimum one-year cycle operation. The KALIMER uranium metallic fueled core characterized by its negative sodium void reactivity and low power density can be operated with maximizing its core safety characteristics as a first generation LMR. The present uranium metallic fueled core allows an easy replacement with different fuel compositions by its demands, with the accumulation of operation experience and design data verification. (author). 34 refs., 34 tabs., 12 figs.

  7. Use of reliability in the LMFBR industry

    International Nuclear Information System (INIS)

    Penland, J.R.; Smith, A.M.; Goeser, D.K.

    1977-01-01

    This mission of a Reliability Program for an LMFBR should be to enhance the design and operational characteristics relative to safety and to plant availability. Successful accomplishment of this mission requires proper integration of several reliability engineering tasks--analysis, testing, parts controls and program controls. Such integration requires, in turn, that the program be structured, planned and managed. This paper describes the technical integration necessary and the management activities required to achieve mission success for LMFBR's

  8. Thermal analysis of metallic fuel for future FBRs

    International Nuclear Information System (INIS)

    Verma, Vishnu; Ghosh, A.K.

    2009-08-01

    A conceptual design of metallic fuel for future Fast Breeder Reactors (FBRs) has been studied. Main objective of metallic fuel is to increase breeding ratio besides power production. The fuel consists of U-Pu binary alloy cladded in T91 alloy with Zr layer between the fuel slug and cladding material and helium gas in semicircular groove region. Temperature distribution of this fuel has been evaluated for different Linear Heat Rate (LHR) for two types of geometry having two and four longitudinal grooves. Analysis has been done in two steps. In first step location of the maximum centerline and clad temperature with corresponding LHR and coolant temperature have been evaluated based on analytical method for circular cylindrical fuel without groove. In second step detailed temperature distribution obtained with the help of Finite Element Method by knowing LHR and coolant temperature at those locations. First step is necessary for 2-D FEM analysis. Contact element has been considered between fuel and clad. Contact pressure due to differential thermal expansion between fuel and clad has been evaluated by finite element method. Input data required for gap conductance was obtained from various models available for gaseous and solid to solid conductance. Effect of contact pressure and thermal conductivity of gas on gap conduction has been studied. Based on detailed temperature distribution it was found that Limiting value of LHR is first reached by clad eutectic temperature between T91 and U-Pu than fuel centerline temperature based on solidus temperature of fuel. (author)

  9. Evaluation of Metal-Fueled Surface Reactor Concepts

    International Nuclear Information System (INIS)

    Poston, David I.; Marcille, Thomas F.; Kapernick, Richard J.; Hiatt, Matthew T.; Amiri, Benjamin W.

    2007-01-01

    Surface fission power systems for use on the Moon and Mars may provide the first use of near-term reactor technology in space. Most near-term surface reactor concepts specify reactor temperatures <1000 K to allow the use of established material and power conversion technology and minimize the impact of the in-situ environment. Metal alloy fuels (e.g. U-10Zr and U-10Mo) have not traditionally been considered for space reactors because of high-temperature requirements, but they might be an attractive option for these lower temperature surface power missions. In addition to temperature limitations, metal fuels are also known to swell significantly at rather low fuel burnups (∼1 a/o), but near-term surface missions can mitigate this concern as well, because power and lifetime requirements generally keep fuel burnups <1 a/o. If temperature and swelling issues are not a concern, then a surface reactor concept may be able to benefit from the high uranium density and relative ease of manufacture of metal fuels. This paper investigates two reactor concepts that utilize metal fuels. It is found that these concepts compare very well to concepts that utilize other fuels (UN, UO2, UZrH) on a mass basis, while also providing the potential to simplify material safeguards issues

  10. The LMFBR safety case

    International Nuclear Information System (INIS)

    Smith, D.

    1990-01-01

    The main objective of this report is to review the present status of the safety case for the liquid metal cooled fast reactor (FBR). A particular emphasis is placed on activities in Europe where the FBR has been progressively developed for many years during which time systems have passed from small experimental plants to the 1200 MWe SPX-1. The FBR has been found to be an easily controlled plant with low impact on the environment and low dose rates to operational personnel. Aspects of reactor design and associated R and D that are required for FBRs to be licensed and the progress made to meet these requirements are described. Fault conditions in the credible range can be dealt with safely, the FBR having several advantageous characteristics which assist safety. Also measures are foreseen to mitigate potential consequences of more severe but improbable accidents. This study sponsored by the Commission of the European Communities was carried out by Colenco Ltd in vlose collaboration with the Safety Working Group (SWG) which is a subgroup of the CEC Fast Reactor Coordinating Committee (FRCC)

  11. The use of metal hydrides in fuel cell applications

    Directory of Open Access Journals (Sweden)

    Mykhaylo V. Lototskyy

    2017-02-01

    Full Text Available This paper reviews state-of-the-art developments in hydrogen energy systems which integrate fuel cells with metal hydride-based hydrogen storage. The 187 reference papers included in this review provide an overview of all major publications in the field, as well as recent work by several of the authors of the review. The review contains four parts. The first part gives an overview of the existing types of fuel cells and outlines the potential of using metal hydride stores as a source of hydrogen fuel. The second part of the review considers the suitability and optimisation of different metal hydrides based on their energy efficient thermal integration with fuel cells. The performances of metal hydrides are considered from the viewpoint of the reversible heat driven interaction of the metal hydrides with gaseous H2. Efficiencies of hydrogen and heat exchange in hydrogen stores to control H2 charge/discharge flow rates are the focus of the third section of the review and are considered together with metal hydride – fuel cell system integration issues and the corresponding engineering solutions. Finally, the last section of the review describes specific hydrogen-fuelled systems presented in the available reference data.

  12. Transfer hook for nuclear fuel assemblies and nuclear reactor having a such hook

    International Nuclear Information System (INIS)

    Thevenot, L.P.

    1990-01-01

    For removing irradiated nuclear fuel assemblies above the level of the liquid metal in the vessel without loss of cooling, the hook mechanism has a guide tube with two annular cavities and a pump to circulate the reactor cooling fluid which flows out by gravity. A such hook used in a LMFBR reduces the height of the reactor vessel and consequently the initial capital cost [fr

  13. Computer simulation of LMFBR piping systems

    International Nuclear Information System (INIS)

    A-Moneim, M.T.; Chang, Y.W.; Fistedis, S.H.

    1977-01-01

    Integrity of piping systems is one of the main concerns of the safety issues of Liquid Metal Fast Breeder Reactors (LMFBR). Hypothetical core disruptive accidents (HCDA) and water-sodium interaction are two examples of sources of high pressure pulses that endanger the integrity of the heat transport piping systems of LMFBRs. Although plastic wall deformation attenuates pressure peaks so that only pressures slightly higher than the pipe yield pressure propagate along the system, the interaction of these pulses with the different components of the system, such as elbows, valves, heat exchangers, etc.; and with one another produce a complex system of pressure pulses that cause more plastic deformation and perhaps damage to components. A generalized piping component and a tee branching model are described. An optional tube bundle and interior rigid wall simulation model makes such a generalized component model suited for modelling of valves, reducers, expansions, and heat exchangers. The generalized component and the tee branching junction models are combined with the pipe-elbow loop model so that a more general piping system can be analyzed both hydrodynamically and structurally under the effect of simultaneous pressure pulses

  14. Review of PRA methodology for LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Yang, J.E

    1999-02-01

    Probabilistic Risk Assessment (PRA) has been widely used as a tool to evaluate the safety of NPPs (Nuclear Power Plants), which are in the design stage as well as in operation. Recently, PRA becomes one of the licensing requirements for many existing and new NPPs. KALIMER is a Liquid Metal Fast Breeder Reactor (LMFBR) being developed by KAERI. Since the design concept of KALIMER is similar to that of the PRISM plant developed by GE, it would be appropriate to review the PRA methodology of PRISM as the first step of KALIMER PRA. Hence, in this report summarizes the PRA methodology of PRISM plant, and the required works for the PSA of KALIMER based on the reviewed results. The PRA technology of PRISM plant consists of following five major tasks: (1) development of initiating event list, (2) development of system event tree, (3) development of core response event tree, (4) development of containment response event tree, and (5) consequences and risk estimation. The estimated individual and societal risk measures show that the risk from a PRISM module is substantially less than the NRC goal. Each task is compared to the PRA methodology of Light Water Reactor (LWR)/Pressurized Heavy Water Reactor (PHWR). In the report, each task of PRISM PRA methodology is reviewed and compared to the corresponding part of LWR/PHWR PSA performed in Korea. The parts that are not modeled appropriately in PRISM PRA are identified, and the recommendations for KALIMER PRA are stated. (author). 14 refs., 9 tabs., 4 figs.

  15. Modeling of constituent redistribution in U Pu Zr metallic fuel

    Science.gov (United States)

    Kim, Yeon Soo; Hayes, S. L.; Hofman, G. L.; Yacout, A. M.

    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.

  16. Metal Matrix Microencapsulated Fuel Technology for LWR Applications

    International Nuclear Information System (INIS)

    Terrani, Kurt A.; Bell, Gary L.; Kiggans, Jim; Snead, Lance Lewis

    2012-01-01

    An overview of the metal matrix microencapsulated (M3) fuel concept for the specific LWR application has been provided. Basic fuel properties and characteristics that aim to improve operational reliability, enlarge performance envelope, and enhance safety margins under design-basis accident scenarios are summarized. Fabrication of M3 rodlets with various coated fuel particles over a temperature range of 800-1300 C is discussed. Results from preliminary irradiation testing of LWR M3 rodlets with surrogate coated fuel particles are also reported.

  17. 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

  18. Research on plant of metal fuel fabrication using casting process

    International Nuclear Information System (INIS)

    Senda, Yasuhide; Mori, Yukihide

    2003-12-01

    This document presents the plant concept of metal fuel fabrication system (38tHM/y) using casting process in electrolytic recycle, which based on recent studies of its equipment design and quality control system. And we estimate the cost of its construction and operation, including costs of maintenance, consumed hardware and management of waste. The content of this work is as follows. (1) Designing of fuel fabrication equipment: We make material flow diagrams of the fuel fabrication plant and rough designs of the injection casting furnace, demolder and inspection equipment. (2) Designing of resolution system of liquid waste, which comes from analytical process facility. Increased analytical items, we rearrange analytical process facility, estimate its chemicals and amount of waste. (3) Arrangement of equipments: We made a arrangement diagram of the metal fuel fabrication equipments in cells. (4) Estimation of cost data: We estimated cost to construct the facility and to operate it. (author)

  19. Apparatus for injection casting metallic nuclear energy fuel rods

    Science.gov (United States)

    Seidel, Bobby R.; Tracy, Donald B.; Griffiths, Vernon

    1991-01-01

    Molds for making metallic nuclear fuel rods are provided which present reduced risks to the environment by reducing radioactive waste. In one embodiment, the mold is consumable with the fuel rod, and in another embodiment, part of the mold can be re-used. Several molds can be arranged together in a cascaded manner, if desired, or several long cavities can be integrated in a monolithic multiple cavity re-usable mold.

  20. Irradiation experience with HT9-clad metallic fuel

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Tsai, H.; Billone, M.C.

    1991-01-01

    The safe and reliable performance of metallic fuel is currently under study and demonstration in the Integral Fast Reactor program. In-reactor tests of HT9-clad metallic fuel have now reached maturity and have all shown good performance characteristics to burnups exceeding 17.5 at. % in the lead assembly. Because this low-swelling tempered martensitic alloy is the cladding of choice for high fluence applications, the experimental observations and performance modelling efforts reported in this paper play an important role in demonstrating reliability

  1. Dynamic simulation of LMFBR systems

    International Nuclear Information System (INIS)

    Agrawal, A.K.; Khatib-Rahbar, M.

    1980-01-01

    This review article focuses on the dynamic analysis of liquid-metal-cooled fast breeder reactor systems in the context of protected transients. Following a brief discussion on various design and simulation approaches, a critical review of various models for in-reactor components, intermediate heat exchangers, heat transport systems and the steam generating system is presented. A brief discussion on choice of fuels as well as core and blanket system designs is also included. Numerical considerations for obtaining system-wide steady-state and transient solutions are discussed, and examples of various system transients are presented. Another area of major interest is verification of phenomenological models. Various steps involved in the code and model verification are briefly outlined. The review concludes by posing some further areas of interest in fast reactor dynamics and safety. (author)

  2. Generic considerations of LMFBR hypothetical accident energetics

    Energy Technology Data Exchange (ETDEWEB)

    Fauske, H.K.; Cho, D.H.; Epstein, M.; Grolmes, M.A.; Henry, R.E.

    1978-01-01

    The paper provides a preliminary assessment of generic accident energetics issues associated with alternatives relative to the reference (U,Pu) oxide fuel in liquid metal fast breeder reactors. The alternatives considered include thorium- and uranium-based oxide, carbide and metal fuel types. This assessment is made within the context of low probability, but potentially large consequence accidents, e.g., core-disruptive accidents.

  3. Closure plug for alkali-metal-bounded fuel rods

    International Nuclear Information System (INIS)

    Guettler, R.

    1974-01-01

    The fuel rod consists of a cladding tube containing an alkali metal which surrounds the fuel pellets. The alkali metal improves the heat transfer. The cladding tube is closed with an end cap at its front end which cap is welded to the cladding tube. Its outside diameter is smaller than the inside diameter of the cladding tube so that the gas can flow out over the alkali metal column during the filling process. The length of the cap is such that the alkali metal is not heated during the welding process. The weld proper is made on a welding collar following a forced fit of the end cap. The end cap may be hollow. (DG) [de

  4. Studies on PEM Fuel Cell Noble Metal Catalyst Dissolution

    DEFF Research Database (Denmark)

    Ma, Shuang; Skou, Eivind Morten

    Incredibly vast advance has been achieved in fuel cell technology regarding to catalyst efficiency, improvement of electrolyte conductivity and optimization of cell system. With breathtakingly accelerating progress, Proton Exchange Membrane Fuel Cells (PEMFC) is the most promising and most widely....... Membrane Electrode Assembly (MEA) is commonly considered as the heart of cell system [2]. Degradation of the noble metal catalysts in MEAs especially Three-Phase-Boundary (TPB) is a key factor directly influencing fuel cell durability. In this work, electrochemical degradation of Pt and Pt/Ru alloy were...

  5. Seismic behaviour of LMFBR reactor cores. The SYMPHONY program

    International Nuclear Information System (INIS)

    Broc, Daniel

    2001-01-01

    As part of a comprehensive program on the seismic behaviour of the LMFBR reactor cores, the SYMPHONY experimental program, performed at the CEA Saclay, is carried out from 1993 up to now. LMFBR reactor cores are composed of fuel assemblies and neutronic shields, immersed in sodium (the primary coolant) or water (for the experimental tests). The main objective of the seismic studies is to evaluate the assembly motions, with consequences on the reactivity and the control rod insertability, and to verify the structural integrity of the assemblies under the impact forces. The experimental program has reached its objectives. Tests have been performed in a satisfying way. Instrumentation allowed to collect displacements, accelerations, and shock forces. All the results constitute a comprehensive base of valuable and reliable data. The interpretation of the tests is based on beam models, taking into account the Fluid Structure Interaction, and the shocks between the assemblies. Theoretical results are in a quite good agreement with the experimental ones. The interpretation of the hexagonal tests in water pointed out very strong coupling between the assemblies and lead to the development of a specific Fluid Structure Interaction, taking into account not only inertial effects, but dissipative effects also. (author)

  6. Fuel upgrading and reforming with metal organic framework

    KAUST Repository

    Eddaoudi, Mohamed

    2016-03-31

    Systems and methods for separating hydrocarbons on an internal combustion powered vehicle via one or more metal organic frameworks are disclosed. Systems and methods can further include utilizing separated hydrocarbons and exhaust to generate hydrogen gas for use as fuel. In one aspect, a method for separating hydrocarbons can include contacting a first component containing a first metal organic framework with a flow of hydrocarbons and separating hydrocarbons by size. In certain embodiments, the hydrocarbons can include alkanes.

  7. Performance analysis of LMFBR control rods

    International Nuclear Information System (INIS)

    Pitner, A.L.; Birney, K.R.

    1975-01-01

    Control rods in the FFTF and LMFBR's will consist of pin bundles of stainless steel-clad boron carbide pellets. In the FFTF reference design, sixty-one pins of 0.474-inch diameter each containing a 36-inch stack of 0.362-inch diameter boron carbide pellets comprise a control rod. Reactivity control is provided by the 10 B (n,α) 7 Li reaction in the boron carbide. This reaction is accompanied by an energy release of 2.8 MeV, and heating from this reaction typically approaches 100 watts/cm 3 for natural boron carbide pellets in an LMFBR flux. Performance analysis of LMFBR control rods must include an assessment of the thermal performance of control pins. In addition, irradiation performance with regard to helium release, pellet swelling, and reactivity worth depletion as a function of service time must be evaluated

  8. Fission gas retention and axial expansion of irradiated metallic fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.; Johanson, E.W.

    1986-05-01

    Out-of-reactor experiments utilizing direct electrical heating and infrared heating techniques were performed on irradiated metallic fuel. The results indicate accelerated expansion can occur during thermal transients and that the accelerated expansion is driven by retained fission gases. The results also demonstrate gas retention and, hence, expansion behavior is a function of axial position within the pin

  9. Calculation of Doses Due to Accidentally Released Plutonium From An LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Fish, B.R.

    2001-08-07

    Experimental data and analytical models that should be considered in assessing the transport properties of plutonium aerosols following a hypothetical reactor accident have been examined. Behaviors of released airborne materials within the reactor containment systems, as well as in the atmosphere near the reactor site boundaries, have been semiquantitatively predicted from experimental data and analytical models. The fundamental chemistry of plutonium as it may be applied in biological systems has been used to prepare models related to the intake and metabolism of plutonium dioxide, the fuel material of interest. Attempts have been made to calculate the possible doses from plutonium aerosols for a typical analyzed release in order to evaluate the magnitude of the internal exposure hazards that might exist in the vicinity of the reactor after a hypothetical LMFBR (Liquid-Metal Fast Breeder Reactor) accident. Intake of plutonium (using data for {sup 239}Pu as an example) and its distribution in the body were treated parametrically without regard to the details of transport pathways in the environment. To the extent possible, dose-response data and models have been reviewed, and an assessment of their adequacy has been made so that recommended or preferred practices could be developed.

  10. LMFBR Emergency Deployment Assuming 45 year Time-Delay Excess CO{sub 2} Removal

    Energy Technology Data Exchange (ETDEWEB)

    Schenewerk, William Ernest [5060 San Rafael Avenue, Los Angeles, CA, 90042-3239 (United States)

    2008-07-01

    Atmospheric CO{sub 2} is presently increasing 2.25% per year in proportion to 2.25% per year exponential fossil fuel consumption increase. CO{sub 2} removal is modeled as being proportional to 45-year-earlier CO{sub 2} amount above 280 ppmV-C. This is: Exp (-0.0225/year * 45 years) = 0.36 fraction CO{sub 2} removed from anthropological emissions, apparently by seawater. LMFBRs use 15 year doubling time. Deploying 30000 GWe atomic power by year-2080 results in CO{sub 2} doubling year-2065 if World primary energy consumption continues increasing 2.25% per year. CO{sub 2} remains roughly twice pre-industrial until year-2100. Beginning year-2080, CO{sub 2} declines at 2.25% per year. CO{sub 2} will presumably decline back to roughly the year-2000 value by year-2200 if the 45-year-delay sink remains effective. LMFBR and GCFR fleet expands to 30000 GWe by 2080. 1000 GWe LWR fleet consumes 5 Mt HM (Heavy Metal). Breeder first cores require 1 Mt HM. (author)

  11. Metal-Free Motifs for Solar Fuel Applications

    Science.gov (United States)

    Ilic, Stefan; Zoric, Marija R.; Kadel, Usha Pandey; Huang, Yunjing; Glusac, Ksenija D.

    2017-05-01

    Metal-free motifs, such as graphitic carbon nitride, conjugated polymers, and doped nanostructures, are emerging as a new class of Earth-abundant materials for solar fuel devices. Although these metal-free structures show great potential, detailed mechanistic understanding of their performance remains limited. Here, we review important experimental and theoretical findings relevant to the role of metal-free motifs as either photoelectrodes or electrocatalysts. First, the light-harvesting characteristics of metal-free photoelectrodes (band energetics, exciton binding energies, charge carrier mobilities and lifetimes) are discussed and contrasted with those in traditional inorganic semiconductors (such as Si). Second, the mechanistic insights into the electrocatalytic oxygen reduction and evolution reactions, hydrogen evolution reaction, and carbon dioxide reduction reaction by metal-free motifs are summarized, including experimental surface-sensitive spectroscopy findings, studies on small molecular models, and computational modeling of these chemical transformations.

  12. Selection of engineering materials and fabrication of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Patriarca, P.

    1975-01-01

    Information is presented graphically and pictorially concerning the need for nuclear power; basic nuclear concepts including BWR, PWR, HTGR, and LMFBR; the fissioning process; nuclear reactor fuel; fabrication of reactor vessels for LMFBR's; fabrication of intermediate heat exchangers for LMFBR's; piping fabrication for LMFBR's; transition welds; steam generators for LMFBR demonstration plants worldwide; stress corrosion cracking of steam generator materials and weldments; post--test examination of the Alco/BLH sodium-heated steam generator; alternate steam generator designs; and alternate structural materials. (DCC)

  13. Fuel damage during off-normal transients in metal-fueled fast reactors

    International Nuclear Information System (INIS)

    Kramer, J.M.; Bauer, T.H.

    1990-01-01

    Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs

  14. Degradation of solid oxide fuel cell metallic interconnects in fuels containing sulfur

    Energy Technology Data Exchange (ETDEWEB)

    Ziomek-Moroz, M.; Hawk, Jeffrey A.

    2005-01-01

    Hydrogen is the main fuel for all types of fuel cells except direct methanol fuel cells. Hydrogen can be generated from all manner of fossil fuels, including coal, natural gas, diesel, gasoline, other hydrocarbons, and oxygenates (e.g., methanol, ethanol, butanol, etc.). Impurities in the fuel can cause significant performance problems and sulfur, in particular, can decrease the cell performance of fuel cells, including solid oxide fuel cells (SOFC). In the SOFC, the high (800-1000°C) operating temperature yields advantages (e.g., internal fuel reforming) and disadvantages (e.g., material selection and degradation problems). Significant progress in reducing the operating temperature of the SOFC from ~1000 ºC to ~750 ºC may allow less expensive metallic materials to be used for interconnects and as balance of plant (BOP) materials. This paper provides insight on the material performance of nickel, ferritic steels, and nickel-based alloys in fuels containing sulfur, primarily in the form of H2S, and seeks to quantify the extent of possible degradation due to sulfur in the gas stream.

  15. Perfomance analysis of boron carbide in LMFBR

    International Nuclear Information System (INIS)

    Pitner, A.L.; Birney, K.R.

    1975-01-01

    Reactivity control in the FFTF and LMFBR's will be maintained by control elements utilizing boron carbide pellets contained in stainless steel pins. Computer performance codes predict irradiation service conditions of absorber pellets and identify required experimental testing. Test results are incorporated in the codes to improve performance prediction capabilities

  16. Simulation of the injection casting of metallic fuels

    International Nuclear Information System (INIS)

    Nakagawa, Tomokazu; Ogata, Takanari; Tokiwai, Moriyasu.

    1989-01-01

    For the fabrication of metallic fuel pins, injection casting is a preferable process because the simplicity of the process is suitable for remote operation. In this process, the molten metal in the crucible is injected into evacuated molds (suspended above the crucible) by pressurizing the casting furnace. Argonne National Laboratory has already adopted this process in the Integral Fast Reactor program. To obtain fuel pins with good quality, the casting parameters, such as the molten metal temperature, the magnitude of the pressure applied, the pressurizing rate, the cooling time, etc., must be optimized. Otherwise, bad-quality castings (short castings, rough surfaces, shrinkage cavities, mold fracture) may result. Therefore, it is very important in designing the casting equipment and optimizing the operation conditions to be able to predict the fluid and thermal behavior of the castings. This paper describes methods to simulate the heat and mass transfer in the molds and molten metallic fuel during injection casting. The results obtained by simulation are compared with experimental ones. Also, appropriate casting conditions for the uranium-plutonium-zirconium alloy are discussed based on the simulated results

  17. LMFBR conceptual design study: an overview of environmental and safety concerns

    International Nuclear Information System (INIS)

    Brenchley, D.L.

    1981-06-01

    The US Department of Energy (DOE) initiated the Liquid Metal Fast Breeder (LMFBR) Conceptual Design Study (CDS) with the objective of maintaining a viable breeder option. The project is scheduled to be completed in FY-1981 but decisions regarding plant construction will be delayed until at least 1985. This report provides a review of the potential environmental and safety engineering concerns for the CDS and recommends specific action for the Environmental and Safety Engineering Division of DOE

  18. Development of an 85,000 gpm (19,303 m3/h) LMFBR primary pump

    International Nuclear Information System (INIS)

    Zerinvary, M.C.; Wagner, E.W.

    1984-01-01

    The development of an 85,000 gpm two-stage primary pump for liquid metal fast breeder reactor (LMFBR) applications is described. The design was supported by air and cavitation model testing of the hyraulics, and development and feature testing of the level control system and the adjustable frequency solid state power supply. Important fabrication and water test items are also discussed, along with some unique assembly tooling requirements

  19. LMFBR conceptual design study: an overview of environmental and safety concerns

    Energy Technology Data Exchange (ETDEWEB)

    Brenchley, D.L.

    1981-06-01

    The US Department of Energy (DOE) initiated the Liquid Metal Fast Breeder (LMFBR) Conceptual Design Study (CDS) with the objective of maintaining a viable breeder option. The project is scheduled to be completed in FY-1981 but decisions regarding plant construction will be delayed until at least 1985. This report provides a review of the potential environmental and safety engineering concerns for the CDS and recommends specific action for the Environmental and Safety Engineering Division of DOE.

  20. Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Cahalan, J.; Wigeland, R.; Friedel, G.; Kussmaul, G.; Royl, P.; Moreau, J.; Perks, M.

    1990-01-01

    In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs

  1. Development of metal cask for nuclear spent fuel

    International Nuclear Information System (INIS)

    Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S.

    2001-01-01

    It is one of the realistic solutions against increasing demand on interim storage of spent fuel assemblies arising from nuclear power plants in Japan to apply dual purpose (transport and storage) metal casks. Since 1980's Mitsubishi Heavy Industries, Ltd. (MHI) has been contributing to develop metal cask technologies for utilities, etc. in Japan, and have established transport and storage cask design ''MSF series'' which realizes higher payload and reliability for long term storage. MSF series transport and storage casks use various new design concepts and materials to improve thermal performance of the cask, structural integrity of the basket, durability of the neutron shielding material and so on. This paper summarizes an outline of the cask design that can accommodate BWR spent fuel assemblies as well as the new technologies applied to the design and fabrication. (author)

  2. Isotopic composition and radiological properties of uranium in selected fuel cycles

    International Nuclear Information System (INIS)

    Fleischman, R.M.; Liikala, R.C.

    1975-04-01

    Three major topic areas are discussed: First, the properties of the uranium isotopes are defined relative to their respective roles in the nuclear fuel cycle. Secondly, the most predominant fuel cycles expected in the U. S. are described. These are the Light Water Reactor (LWR), High Temperature Gas Cooled Reactor (HTGR), and Liquid Metal Fast Breeder Reactor (LMFBR) fuel cycles. The isotopic compositions of uranium and plutonium fuels expected for these fuel cycles are given in some detail. Finally the various waste streams from these fuel cycles are discussed in terms of their relative toxicity. Emphasis is given to the high level waste streams from reprocessing of spent fuel. Wastes from the various fuel cycles are compared based on projected growth patterns for nuclear power and its various components. (U.S.)

  3. Platinum redispersion on metal oxides in low temperature fuel cells

    DEFF Research Database (Denmark)

    Tripkovic, Vladimir; Cerri, Isotta; Nagami, Tetsuo

    2013-01-01

    We have analyzed the aptitude of several metal oxide supports (TiO2, SnO2, NbO2, ZrO2, SiO2, Ta2O5 and Nb2O5) to redisperse platinum under electrochemical conditions pertinent to the Proton Exchange Membrane Fuel Cell (PEMFC) cathode. The redispersion on oxide supports in air has been studied in ...

  4. Probabilistic assessment of critically flawed LMFBR PHTS piping elbows

    International Nuclear Information System (INIS)

    Balkey, K.R.; Wallace, I.T.; Vaurio, J.K.

    1982-01-01

    One of the important functions of the Primary Heat Transport System (PHTS) of a large Liquid Metal Fast Breeder Reactor (LMFBR) plant is to contain the circulating radioactive sodium in components and piping routed through inerted areas within the containment building. A significant possible failure mode of this vital system is the development of cracks in the piping components. This paper presents results from the probabilistic assessment of postulated flaws in the most-critical piping elbow of each piping leg. The criticality of calculated maximum sized flaws is assessed against an estimated material fracture toughness to determine safety factors and failure probability estimates using stress-strength interference theory. Subsequently, a different approach is also employed in which the randomness of the initial flaw size and loading are more-rigorously taken into account. This latter approach yields much smaller probability of failure values when compared to the stress-strength interference analysis results

  5. 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)

  6. Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

    OpenAIRE

    Chan Bock Lee; Jin Sik Cheon; Sung Ho Kim; Jeong-Yong Park; Hyung-Kook Joo

    2016-01-01

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU)–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochem...

  7. 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

  8. Technology of Fabrication for Sodium-cooled Fast Reactor Metallic Fuel

    International Nuclear Information System (INIS)

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

    2008-02-01

    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

  9. Method for producing hydrocarbon fuels and fuel gas from heavy polynuclear hydrocarbons by the use of molten metal halide catalysts

    Science.gov (United States)

    Gorin, Everett

    1979-01-01

    In a process for hydrocracking heavy polynuclear carbonaceous feedstocks to produce lighter hydrocarbon fuels by contacting the heavy feedstocks with hydrogen in the presence of a molten metal halide catalyst in a hydrocracking zone, thereafter separating at least a major portion of the lighter hydrocarbon fuels from the spent molten metal halide and thereafter regenerating the spent molten metal halide by incinerating the spent molten metal halide by combustion of carbon and sulfur compounds in the spent molten metal halide in an incineration zone, the improvement comprising: (a) contacting the heavy feedstocks and hydrogen in the presence of the molten metal halide in the hydrocracking zone at reaction conditions effective to convert from about 60 to about 90 weight percent of the feedstock to lighter hydrocarbon fuels; (b) separating at least a major portion of the lighter hydrocarbon fuels from the spent molten metal halide; (c) contacting the spent molten metal halide with oxygen in a liquid phase gasification zone at a temperature and pressure sufficient to vaporize from about 25 to about 75 weight percent of the spent metal halide, the oxygen being introduced in an amount sufficient to remove from about 60 to about 90 weight percent of the carbon contained in the spent molten metal halide to produce a fuel gas and regenerated metal halide; and (d) incinerating the spent molten metal halide by combusting carbon and sulfur compounds contained therein.

  10. Bubble behavior in LMFBR core disruptive accidents. Annual report, June 1, 1975--June 30, 1976

    International Nuclear Information System (INIS)

    Reynolds, A.B.; Erdman, C.A.; Garner, P.L.; Kennedy, M.F.; Rao, S.P.; Refling, J.G.

    1976-08-01

    The work reported here is part of the Aerosol Release and Transport program for LMFBR safety assessment for the Reactor Safety Research Division of the U.S. Nuclear Regulatory Commission. Six areas were at various stages of investigation during this reporting period. A study of nonequilibrium mass transfer during fuel expansion and a study of the dynamics of fuel expansion into the sodium pool were completed. Studies are underway on condensation on above-core structures and on generation of aerosols from condensation. Studies were initiated on small-particle generation from hydrodynamic fragmentation, on particle kinematics and on particle-surface interaction

  11. Fast reactors fuel Cycle: State in Europe

    International Nuclear Information System (INIS)

    1991-01-01

    In this SFEN day we treat all aspects (economics-reactor cores, reprocessing, experience return) of the LMFBR fuel cycle in Europe and we discuss about the development of this type of reactor (EFR project) [fr

  12. High performance, high durability non-precious metal fuel cell catalysts

    Science.gov (United States)

    Wood, Thomas E.; Atanasoski, Radoslav; Schmoeckel, Alison K.

    2016-03-15

    This invention relates to non-precious metal fuel cell cathode catalysts, fuel cells that contain these catalysts, and methods of making the same. The fuel cell cathode catalysts are highly nitrogenated carbon materials that can contain a transition metal. The highly nitrogenated carbon materials can be supported on a nanoparticle substrate.

  13. Chemical factors affecting fission product transport in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly

  14. Chemical factors affecting fission product transport in severe LMFBR accidents

    Energy Technology Data Exchange (ETDEWEB)

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly.

  15. Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Benoit, Timothy [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hlotke, John Daniel [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2017-07-05

    This document contains an evaluation of the applicability of the current Quality Assurance Standards from the American Society of Mechanical Engineers Standard NQA-1 (NQA-1) criteria and identifies and describes the quality assurance process(es) by which attributes of historical, analytical, and other data associated with sodium-cooled fast reactor [SFR] metallic fuel and/or related reactor fuel designs and constituency will be evaluated. This process is being instituted to facilitate validation of data to the extent that such data may be used to support future licensing efforts associated with advanced reactor designs. The initial data to be evaluated under this program were generated during the US Integral Fast Reactor program between 1984-1994, where the data includes, but is not limited to, research and development data and associated documents, test plans and associated protocols, operations and test data, technical reports, and information associated with past United States Nuclear Regulatory Commission reviews of SFR designs.

  16. Fuel pin transient behavior technology applied to safety analyses. Presentation to AEC Regulatory Staff 4th Regulatory Briefing on safety technology, Washington, D.C., November 19--20, 1974

    International Nuclear Information System (INIS)

    1974-11-01

    Information is presented concerning LMFBR fuel pin performance requirements and evaluation; fuels behavior codes with safety interfaces; performance evaluations; ex-reactor materials and simulation tests; models for fuel pin failure; and summary of continuing fuels technology tasks. (DCC)

  17. Fully-Coupled Metallic Fuel Performance Simulations using BISON

    Energy Technology Data Exchange (ETDEWEB)

    Galloway, Jack D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Unal, Cetin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-08-27

    This document is a set of slides intended to accompany a talk at a meeting. The first topic taken up is zirconium redistribution. The rod edge Zr increase is evidently due to the Soret term and temperature gradient. Then metallic fission gas release modeling is considered. Based on a GRSIS/FEAST model, the approach of generating fission gas in the fuel matrix is described. A sensitivity study on parameters is presented, including sodium bond & diffusion coefficient sensitivity along with dt sensitivity. Finally, results of some coupled simulations are shown, with ideas about future work.

  18. Thermal performance of fresh mixed-oxide fuel in a fast flux LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.

    1985-01-01

    A test was designed and irradiated to provide power-to-melt (heat generation rate necessary to initiate centerline fuel melting) data for fresh mixed-oxide UO 2 -PuO 2 fuel irradiated in a fast neutron flux under prototypic liquid metal reactor (LMR) conditions. The fuel pin parameters were selected to envelope allowable fabrication ranges and address mass production of LMR fuel using sintered-to-size techniques. The test included fuel pins with variations in fabrication technique, pellet density, fuel-to-cladding gap, Pu concentration, and fuel oxygen-to-metal ratios. The resulting data base has reestablished the expected power-to-melt in mixed-oxide fuels during initial reactor startup when the fuel temperatures are expected to be the highest. Calibration of heat transfer models of fuel pin performance codes with these data are providing more accurate capability for predicting steady-state thermal behavior of current and future mixed-oxide LMR fuels

  19. Epsilon metal waste form for immobilization of noble metals from used nuclear fuel

    Science.gov (United States)

    Crum, Jarrod V.; Strachan, Denis; Rohatgi, Aashish; Zumhoff, Mac

    2013-10-01

    Epsilon metal (ɛ-metal), an alloy of Mo, Pd, Rh, Ru, and Tc, is being developed as a waste form to treat and immobilize the undissolved solids and dissolved noble metals from aqueous reprocessing of commercial used nuclear fuel. Epsilon metal is an attractive waste form for several reasons: increased durability relative to borosilicate glass, it can be fabricated without additives (100% waste loading), and in addition it also benefits borosilicate glass waste loading by eliminating noble metals from the glass, thus the processing problems related to their insolubility in glass. This work focused on the processing aspects of the epsilon metal waste form development. Epsilon metal is comprised of refractory metals resulting in high alloying temperatures, expected to be 1500-2000 °C, making it a non-trivial phase to fabricate by traditional methods. Three commercially available advanced technologies were identified: spark-plasma sintering, microwave sintering, and hot isostatic pressing, and investigated as potential methods to fabricate this waste form. Results of these investigations are reported and compared in terms of bulk density, phase assemblage (X-ray diffraction and elemental analysis), and microstructure (scanning electron microscopy).

  20. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    International Nuclear Information System (INIS)

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong

    2012-01-01

    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced

  1. Analysis of metal fuel transient overpower experiments with the SAS4A accident analysis code

    International Nuclear Information System (INIS)

    Tentner, A.M.; Kalimullah; Miles, K.J.

    1990-01-01

    The results of the SAS4A analysis of the M7 TREAT Metal fuel experiment are presented. New models incorporated in the metal fuel version of SAS4A are described. The computational results are compared with the experimental observations and this comparison is used in the interpretation of physical phenomena. This analysis was performed using the integrated metal fuel SAS4A version and covers a wide range of events, providing an increased degree of confidence in the SAS4A metal fuel accident analysis capabilities

  2. Casting technology for manufacturing metal rods from simulated metallic spent fuels

    Science.gov (United States)

    Leeand, Y. S.; Lee, D. B.; Kim, C. K.; Shin, Y. J.; Lee, J. H.

    2000-09-01

    A uranium metal rod 13.5 mm in diameter and 1,150 mm long was produced from simulated metallic spent fuels with advanced casting equipment using the directional-solidification method. A vacuum casting furnace equipped with a four-zone heater to prevent surface oxidation and the formation of surface shrinkage holes was designed. By controlling the axial temperature gradient of the casting furnace, deformation by the surface shrinkage phenomena was diminished, and a sound rod was manufactured. The cooling behavior of the molten uranium was analyzed using the computer software package MAGMAsoft.

  3. Development of models for the sodium version of the two-phase three-dimensional thermal hydraulics code THERMIT. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.J.; Kazimi, M.S.

    1980-05-01

    Several different models and correlations were developed and incorporated in the sodium version of THERMIT, a thermal-hydraulics code written at MIT for the purpose of analyzing transients under LMFBR conditions. This includes: a mechanism for the inclusion of radial heat conduction in the sodium coolant as well as radial heat loss to the structure surrounding the test section. The fuel rod conduction scheme was modified to allow for more flexibility in modelling the gas plenum regions and fuel restructuring. The formulas for mass and momentum exchange between the liquid and vapor phases were improved. The single phase and two phase friction factors were replaced by correlations more appropriate to LMFBR assembly geometry.

  4. Maintenance and repair of LMFBR steam generators: specialists` meeting, O-Arai Engineering Center, Japan, 4-8 June 1984. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1984-07-01

    The Specialists` Meeting on "Maintenance and Repair of LMFBR Steam Generators" was held in Oarai, Japan, from 4-8 June 1984. The meeting was sponsored by the International Atomic Energy Agency on the recommendation of the IAEA International Working Group on Fast Reactors and was hosted by the Power Reactor and Nuclear Fuel Development Corporation of Japan. The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topical areas were discussed by participants: national review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; research and development work on maintenance and repair; and experience on steam generator maintenance and repair.

  5. Cesium chemistry in GCFR fuel pins

    International Nuclear Information System (INIS)

    Fee, D.C.; Johnson, C.E.

    1979-01-01

    The fuel rod design for the Gas Cooled Fast-Breeder Reactor (GCFR) is similar to that employed for the Liquid Metal Fast Breeder Reactor (LMFBR) with the exception of the unique features inherent to the use of helium as the coolant. These unique design features include the use of (1) vented and pressure-equalized fuel rods, and (2) ribbed cladding along 75% of the fuel section. The former design feature enables reduction in cladding thickness and prevention of possible creep collapse of the cladding due to the high coolant pressure (8.5 MPa). The latter design feature brings about improved heat transfer characteristics. Each GCFR fuel rod is vented to a manifold whereby gaseous fission products diffusing out of the fuel pin are retained on charcoal traps. As a result, the internal pressure of a GCFR fuel pin does not increase during irradiation. In addition, the venting system also maintains the pressure within the fuel pin slightly below (0.3 to 0.5 MPa) the coolant pressure outside the fuel pin. Consequently, should a breach occur in the cladding, helium flows into the breached fuel pin thereby minimizing fission product contamination of the coolant. These desirable aspects of a GCFR fuel pin can be maintained only as long as axial gas transport paths are available and operating within the fuel pin

  6. Pyrometallurgical processing of Integral Fast Reactor metal fuels

    International Nuclear Information System (INIS)

    Battles, J.E.; Miller, W.E.; Gay, E.C.

    1991-01-01

    The pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor is now in an advanced state of development. This process involves electrorefining spent fuel with a cadmium anode, solid and liquid cathodes, and a molten salt electrolyte (LiCl-KCl) at 500 degrees C. The initial process feasibility and flowsheet verification studies have been conducted in a laboratory-scale electrorefiner. Based on these studies, a dual cathode approach has been adopted, where uranium is recovered on a solid cathode mandrel and uranium-plutonium is recovered in a liquid cadmium cathode. Consolidation and purification (salt and cadmium removal) of uranium and uranium-plutonium products from the electrorefiner have been successful. The process is being developed with the aid of an engineering-scale electrorefiner, which has been successfully operated for more than three years. In this electrorefiner, uranium has been electrotransported from the cadmium anode to a solid cathode in 10 kg quantities. Also, anodic dissolution of 10 kg batches of chopped, simulated fuel (U--10% Zr) has been demonstrated. Development of the liquid cadmium cathode for recovering uranium-plutonium is under way

  7. The evaluation of the use of metal alloy fuels in pressurized water reactors

    International Nuclear Information System (INIS)

    Lancaster, D.

    1992-01-01

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ''advanced reactors,'' it became clear that reactor design optimization has been under emphasized. Current ''advanced reactors'' are severely constrained. The AP-600 required the use of a fuel design from the 1970's. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing

  8. The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lancaster, D.

    1992-10-26

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

  9. FFTF [fast flux tests facility] metal fuel pin sodium bond quality verification

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1989-01-01

    The Fast Flux Tests Facility (FFTF) Series III driver fuel design consists of U-10 Zr fuel slugs contained in a ferritic alloy cladding. A liquid-metal, sodium bond between the fuel and cladding is required to prevent unacceptable temperatures during operation. Excessive voiding or porosity in the sodium thermal bond could result in localized fuel melting during irradiation. It is therefore imperative that bond quality be verified during fabrication of these metal fuel pins prior to irradiation. Cladding stripping of two spare metal fuel pins was successfully performed to inspect the quality of the sodium thermal bond and to compare observed defects in the bond with eddy current measurements. Bond quality was found to be generally good, thereby confirming the fabrication process used to bond the fuel and cladding with liquid-metal sodium to be acceptable. Observed porosity distribution in the sodium bond correlated well with eddy current indications

  10. Metal lagging

    International Nuclear Information System (INIS)

    Lemercier, Guy.

    1974-01-01

    The metal lagging described is characterized by the fact that it is formed of closed sacks composed of an elastic metal mass, compressed in an outer envelope made of a fine mesh metal fabric. The metal mass is composed of stainless steel wool stuffed into the envelope. This lagging is particularly intended for the thermal protection of the end slab of LMFBR type reactors [fr

  11. LMFBR Ultra Long Life Cores

    International Nuclear Information System (INIS)

    Schmidt, J.E.; Doncals, R.A.; Porter, C.A.; Gundy, L.M.

    1986-01-01

    The Ultra Long Life Core is an attractive and innovative design approach with several extremely beneficial attributes. Long Life cores are applicable to the full range of LMR plant sizes resulting in lifetimes up to 30 years. Core life is somewhat limited for smaller plant sizes, however significant benefits of this approach still exist for all plant sizes. The union of long life cores and the complementary inherent safety technology offer a means of utilizing the well-proven oxide fuel in a system with unsurpassed safety capability. A further benefit is that the uranium fuel cycle can be used in long life cores, especially for initial LMR plant deployment, thereby eliminating the need for reprocessing prior to starting LMR plant construction in the U.S. Finally the long life core significantly reduces power costs. With inherent safety capability designed into an LMR and with the ULLC fuel cycle, power costs competitive with light water plants are achievable while offering improved operational flexibility derived through extending refueling intervals

  12. Platinum redispersion on metal oxides in low temperature fuel cells.

    Science.gov (United States)

    Tripković, Vladimir; Cerri, Isotta; Nagami, Tetsuo; Bligaard, Thomas; Rossmeisl, Jan

    2013-03-07

    We have analyzed the aptitude of several metal oxide supports (TiO(2), SnO(2), NbO(2), ZrO(2), SiO(2), Ta(2)O(5) and Nb(2)O(5)) to redisperse platinum under electrochemical conditions pertinent to the Proton Exchange Membrane Fuel Cell (PEMFC) cathode. The redispersion on oxide supports in air has been studied in detail; however, due to different operating conditions it is not straightforward to link the chemical and the electrochemical environment. The largest differences reflect in (1) the oxidation state of the surface (the oxygen species coverage), (2) temperature and (3) the possibility of platinum dissolution at high potentials and the interference of redispersion with normal working potential of the PEMFC cathode. We have calculated the PtO(x) (x = 0, 1, 2) adsorption energies on different metal oxides' surface terminations as well as inside the metal oxides' bulk, and we have concluded that NbO(2) might be a good support for platinum redispersion at PEMFC cathodes.

  13. 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

  14. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Appendixes D and E. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  15. Response of Salix alba L. to heavy metals and diesel fuel ...

    African Journals Online (AJOL)

    Response of Salix alba L. to heavy metals and diesel fuel contamination. ... Ni and Pb led to reduced accumulation of Pb. Presence of 5 g/kg of diesel fuel in soil significantly increased toxic influence of applied heavy metals by further reducing ... Key words: Cd, Ni, Pb, phytoremediation, phytoextraction, willow, Salix alba.

  16. Metallic Fuel Casting Development and Parameter Optimization Simulations

    International Nuclear Information System (INIS)

    Fielding, Randall S.; Kennedy, J.R.; Crapps, J.; Unal, C.

    2013-01-01

    Conclusions: • Gravity casting is a feasible process for casting of metallic fuels: – May not be as robust as CGIC, more parameter dependent to find right “sweet spot” for high quality castings; – Fluid flow is very important and is affected by mold design, vent size, super heat, etc.; – Pressure differential assist was found to be detrimental. • Simulation found that vent location was important to allow adequate filling of mold; • Surface tension plays an important role in determining casting quality; • Casting and simulations high light the need for better characterized fluid physical and thermal properties; • Results from simulations will be incorporated in GACS design such as vent location and physical property characterization

  17. Inherent reactor power controller for a metal-fueled ALMR

    International Nuclear Information System (INIS)

    Wood, R.T.; Wilson, T.L. Jr.

    1990-01-01

    Inherent power control for metal-fueled ALMR designs involves using reactivity thermal feedback effects to control reactor power. This paper describes how, using classical control design techniques, a control system for normal load following maneuvers was deigned for a pool-type ALMR. This design provides active control of power removal in the balance of plant, direct control of selected primary and intermediate loop temperatures, and passive control of reactor power. The inherent stability of the strong, fast reactivity feedback effects bring heat production in the core into balance with the heat removal system temperatures, which are controlled to meet power demand. A simulation of the control system successfully responded to a 10% step change in power demand by changing power at an acceptable rate without causing large temperature fluctuations or exceeding thermal limits

  18. Studies on PEM fuel cell noble metal catalyst dissolution

    DEFF Research Database (Denmark)

    Andersen, S. M.; Grahl-Madsen, L.; Skou, E. M.

    2011-01-01

    A combination of electrochemical, spectroscopic and gravimetric methods was carried out on Proton Exchange Membrane (PEM) fuel cell electrodes with the focus on platinum and ruthenium catalysts dissolution, and the membrane degradation. In cyclic voltammetry (CV) experiments, the noble metals were...... found to dissolve in 1 M sulfuric acid solution and the dissolution increased exponentially with the upper potential limit (UPL) between 0.6 and 1.6 vs. RHE. 2-20% of the Pt (depending on the catalyst type) was found to be dissolved during the experiments. Under the same conditions, 30-100% of the Ru...... (depending on the catalyst type) was found to be dissolved. The faster dissolution of ruthenium compared to platinum in the alloy type catalysts was also confirmed by X-ray diffraction measurements. The dissolution of the carbon supported catalyst was found one order of magnitude higher than the unsupported...

  19. Design characteristics of metallic fuel rod on its in-LMR performance

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang Hee Young; Nam, Cheol; Kim, Jong Oh

    1997-01-01

    Fuel design is a key feature to assure LMR safety goals. To date, a large effort had been devoted to develop metallic fuels at ANL's experimental breeder reactor (EBR-II). The major design and performance parameters investigated include; thermal conductivity and temperature profile; smear density; axial plenum; FCMI and cladding deformation including creep, and fission gas release. In order to evaluate the sensitivity of each parameter, in-LMR performances of metallic fuels are not only reviewed by the experiment results in literatures, but also key design characteristics according to the variation of metallic fuel rod design parameters are analyzed by using the MACSIS code which simulates in-reactor behaviors of metal fuel rod. In this study, key design characteristics and the criteria which must be considered to design fuel rod in LMR, are proposed and discussed. (author). 14 refs., 4 figs

  20. Nuclear and thermal-hydraulic characteristics for an LMR core fueled with 20% enriched uranium metallic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young-In; Kim, Young-Gyun; Kim, Sang-Ji; Kim, Young-Jin

    1999-05-01

    As a part of the core design development of KALIMER (150 MWe), the KALIMER core was initially designed with 20% enriched uranium metallic fuel. In this core design, the primary emphasis was given to realize the metallic fueled core design to meet the specific design requirements; 20% and below uranium enrichment and a minimum fuel cycle length of one year. The core was defined by a radially homogeneous core configuration incorporated with several passive design features to give inherent passive means of negative reactivity insertion. The core nuclear performance based on a once-through equilibrium fuel cycle scenario shows that the core has an average breeding ratio of 0.67 and maximum discharge burnup of 47.3 MWD/kg. When comparing with conventional plutonium metallic fueled cores of the same power level, the present uranium metallic fueled core has a lower power density due to its increased physical core size. The negative sodium void reactivity over the core shows a beneficial potential to assure inherent safety characteristics. The transition from the uranium startup to equilibrium cycle is feasible without any design change. Core nuclear performance characteristics in the present core design are attributed to the specific design requirements of enrichment restriction and fuel cycle length.

  1. Ferritic steels for French LMFBR steam generators

    International Nuclear Information System (INIS)

    Aubert, M.; Mathieu, B.; Petrequin, P.

    1983-06-01

    Austenitic stainless steels have been widely used in many components of the French LMFBR. Up to now, ferritic steels have not been considered for these components, mainly due to their relatively low creep properties. Some ferritic steels are usable when the maximum temperatures in service do not exceed about 530 0 C. It is the case of the steam generators of the Phenix plant, where the exchange tubes of the evaporator are made of 2,25% Cr-1% Mo steel, stabilized or not by addition of niobium. These ferritic alloys have worked successfully since the first steam production in October 1973. For the SuperPhenix power plant, an ''all austenitic stainless alloy'' apparatus has been chosen. However, for the future, ferritic alloys offer potential for use as alternative materials in the evaporators: low alloys steels type 2,25% Cr-1% Mo (exchange tubes, tube-sheets, shells), or at higher chromium content type 9% Cr-2% Mo NbV (exchange tubes) or 12M Cr-1% Mo-V (tube-sheets). Most of these steels have already an industrial background, and are widely used in similar applications. The various potential applications of these steels are reviewed with regards to the French LMFBR steam generators, indicating that some points need an effort of clarification, for instance the properties of the heterogeneous ferritic/austenitic weldments

  2. Development of 4S and related technologies (2). Long life metallic fuel

    International Nuclear Information System (INIS)

    Yacout, A.M.; Tsuboi, Y.; Ueda, N.

    2009-01-01

    This paper provides an overview of the long life metallic fuel to be used in the 4S reactor. The 4S fuel design is presented and implications of its characteristics on fuel performance are discussed. Main design characteristics include the long fuel life time of 30 years and the wider and longer fuel pins compared to EBR-II and FFTF fuel pins. The LIFE-METAL fuel performance code was used to evaluate the performance of the 4S fuel design. The code has been validated using post irradiation examination data of metallic fuel irradiated in EBR-II. The performance evaluation shows the benign nature of the design. The design enables the fuel to perform adequately during reactor operations without violating any of a conservative set of steady state design criteria. A survey evaluation of the fuel performance is also presented. This performance bounding evaluation took into account possible fuel swelling behavior and cladding temperature range that represents worst case scenarios. The evaluation showed that the fuel maintains its integrity even under those worst case conditions. (author)

  3. 85,000-GPM, single-stage, single-suction LMFBR intermediate centrifugal pump

    International Nuclear Information System (INIS)

    Fair, C.E.; Cook, M.E.; Huber, K.A.; Rohde, R.

    1983-01-01

    The mechanical and hydraulic design features of the 85,000-gpm, single-stage, single-suction pump test article, which is designed to circulate liquid-sodium coolant in the intermediate heat-transport system of a Large-Scale Liquid Metal Fast Breeder Reactor (LS-LMFBR), are described. The design and analytical considerations used to satisfy the pump performance and operability requirements are presented. The validation of pump hydraulic performance using a hydraulic scale-model pump is discussed, as is the featute test for the mechanical-shaft seal system

  4. LMFBR safety. 1. Review of current issues and bibliography of literature, 1960--1969

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-08-16

    This report discusses the current status of liquid-metal fast breeder (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1960 through 1969. The bibliography consists of 1560 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

  5. LMFBR safety. 4. Review of current issues and bibliography of literature (1974--1975)

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-03-21

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1974 through 1975. The bibliography consists of approximately 1554 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

  6. LMFBR safety. 2. Review of current issues and bibliography of literature, 1970--1972

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-11-22

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1970 through 1972. The bibliography consists of approximately 1620 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

  7. LMFBR safety. 3. Review of current issues and bibliography of literature (1972--1974)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-01-01

    The report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1972 through 1974. The bibliography consists of approximately 1380 abstracts covering research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included

  8. LMFBR safety. 2. Review of current issues and bibliography of literature, 1970--1972

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-01-01

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1970 through 1972. The bibliography consists of approximately 1620 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness

  9. LMFBR safety. 3. Review of current issues and bibliography of literature (1972--1974)

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-02-24

    The report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1972 through 1974. The bibliography consists of approximately 1380 abstracts covering research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included.

  10. Transference of advanced LMFBR control technology to the aerospace power system program

    International Nuclear Information System (INIS)

    Chisholm, G.H.

    1984-01-01

    Much recent R and D has been devoted to the safety of liquid metal fast breeder reactors (LMFBR's). Part of the resulting technology, especially advanced control systems, appears to be directly transferable to the space nuclear power program. Some of the ideas described herein have been already culminated in successful products that are available for application, e.g. analytical redundancy and fault-tolerant computers. Others, in various stages of R and D, are being developed as elements to support the design goals outlined in the following section, e.g. automated software verification, automated hardware verification, and system validation

  11. LMFBR safety. 4. Review of current issues and bibliography of literature (1974--1975)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-01-01

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1974 through 1975. The bibliography consists of approximately 1554 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness

  12. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

  13. LMFBR safety. 5. Review of current issues and bibliography of literature (1975--1976)

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-06-08

    The current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA), are discussed. Bibliographic information on worldwide LMFBRs relative to the development and safety of the breeder reactor is presented for the period 1975 through 1976. The bibliography consists of approximately 1618 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Keyword, author, and permuted-title indexes are included for completeness.

  14. LMFBR safety. 1. Review of current issues and bibliography of literature, 1960--1969

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-01-01

    This report discusses the current status of liquid-metal fast breeder (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1960 through 1969. The bibliography consists of 1560 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness

  15. LMFBR subassembly response to local pressure loadings: an experimental approach

    International Nuclear Information System (INIS)

    Marciniak, T.J.; Ash, J.E.; Marchertas, A.H.; Cagliostro, D.J.

    1975-01-01

    An experimental program to determine the response of LMFBR-type subassemblies to local subassembly accidents caused by pressure loadings is described. Some results are presented and compared with computer calculations

  16. Loss-of-flow transient characterization in carbide-fueled LMFBRs

    Science.gov (United States)

    Rothrock, R. B.; Morgan, M. M.; Baars, R. E.; Elson, J. S.; Wray, M. L.

    One of the benefits derived from the use of carbide fuel in advanced liquid metal fast breeder reactors (LMFBRs) is a decreased vulnerability to certaiin accidents. This can be achieved through the combination of advanced fuel performance with the enhanced reactivity feedback effects and passive shutdown cooling systems characteristic of the current inherently safe plant concepts. The calculated core response to an unprotected loss of flow (ULOF) accident has frequently been used as a benchmark test of these designs, and the advantages of a high-conductivity fuel in relation to this type of transient have been noted in previous analyses. To evaluate this benefit in carbide-fueled LMFBRs incorporating representative current plant design features, limited calculations have been made of a ULOF transient in a small (modular) carbide-fueled LMFBR.

  17. A basic research on the transient behavior for a metallic fuel FBR

    International Nuclear Information System (INIS)

    Baba, Mamoru; Hirano, Go; Kawada, Ken-ichi; Niwa, Hajime

    1999-03-01

    A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'A basic research on the transient behavior for a metallic fuel FBR'. The results are the following. (1) Target and Results of analysis: The accident initiator considered is a LOF accident without scram. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transition phase. It is shown that the moderator is quite elective to mitigate the accident at the initiation phase. However, it is necessary to analyze the transition phase to know if the re-criticality is totally avoided after the initiation phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large scale material movement in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics, and examined

  18. Report on FY16 Low-dose Metal Fuel Irradiation and PIE

    Energy Technology Data Exchange (ETDEWEB)

    Edmondson, Philip D.

    2016-09-01

    This report gives an overview of the efforts into the low-dose metal fuel irradiation and PIE as part of the Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC) milestone M3FT-16OR020303031. The current status of the FCT and FCRP irradiation campaigns are given including a description of the materials that have been irradiated, analysis of the passive temperature monitors, and the initial PIE efforts of the fuel samples.

  19. Development of metal uranium fuel and testing of construction materials (I-VI); Part I

    International Nuclear Information System (INIS)

    Mihajlovic, A.

    1965-11-01

    This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors

  20. Behavior of spent nuclear fuel and storage-system components in dry interim storage

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

    1982-08-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom and organic-cooled reactor (OCR) fuel in silos in Canada. Dry storage demonstrations are under way for Zircaloy-clad fuel from boiling water reactors BWR's, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions

  1. Behavior of spent nuclear fuel and storage system components in dry interim storage.

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

    1982-08-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom and organic-cooled reactor (OCR) fuel in silos in Canada. Dry storage demonstrations are under way for Zircaloy-clad fuel from boiling water reactors BWR's, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions.

  2. Influence of metallic based fuel additives on performance and exhaust emissions of diesel engine

    Energy Technology Data Exchange (ETDEWEB)

    Keskin, Ali [Tarsus Technical Education Faculty, Mersin University, 33500 Mersin (Turkey); Guerue, Metin, E-mail: mguru@gazi.edu.t [Engineering and Architectural Faculty, Gazi University, 06570 Maltepe, Ankara (Turkey); Altiparmak, Duran [Technical Education Faculty, Gazi University, 06500 Ankara (Turkey)

    2011-01-15

    In this experimental study, influence of the metallic-based additives on fuel consumption and exhaust emissions of diesel engine were investigated. The metallic-based additives were produced by synthesizing of resin acid (abietic acid) with MnO{sub 2} or MgO. These additives were doped into diesel fuel at the rate of 8 {mu}mol/l and 16 {mu}mol/l for preparing test fuels. Both additives improved the properties of diesel fuel such as viscosity, flash point, cloud point and pour point. The fuels with and without additives were tested in a direct injection diesel engine at full load condition. Maximum reduction of specific fuel consumption was recorded as 4.16%. CO emission and smoke opacity decreased by 16.35% and by 29.82%, respectively. NO{sub x} emission was measured higher and CO{sub 2} emission was not changed considerably with the metallic-based additives.

  3. Proceedings of the third specialist meeting on sodium/fuel interaction in fast reactors

    International Nuclear Information System (INIS)

    1976-01-01

    This specialist meeting, sponsored by the OECD-NEA and organized by the Power Reactor and Nuclear Fuel Development Corporation, was attended by 56 delegates from 6 countries and the CEC (Commission of the European Communities). The purpose of the meeting was to bring together and discuss in depth the Fuel-Sodium Interaction, a phenomenon of major importance in the assessment of the Hypothetical Core Disruptive Accident in the Liquid Metal Fast Breeder Reactor. The meeting was essentially a follow-up of an earlier meeting held at Ispra in December 1973. In all, 29 papers were presented, covering the following topics: 1. Current perspective on sodium-fuel interaction in LMFBR safety; 2. Basic experimental and theoretical studies including other materials; 3. In-pile and out-of-pile experimental studies on sodium-fuel interaction; 4. Theoretical models for the interpretation of experiments and for application to reactor situations. The meeting is considered useful in narrowing down the chain of events necessary to get energetic interaction, large work potential, but many points are being clarified on the gap between the basic vapor explosions and the real fuel sodium interactions in the HCDA scenario of LMFBR. Finally another meeting of the same nature as this one has been recommended

  4. Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-01

    The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

  5. The basic research on the CDA initiation phase for a metallic fuel FBR

    International Nuclear Information System (INIS)

    Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime

    1998-03-01

    A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku University and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled. (1) Target and Results of analysis: The accident initiator considered is a LOF accident with ATWS. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transient phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large material movement to in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics. The results of a sample case, that is a metallic fueled core at the beginning of cycle, show this improvement is appropriate. (3) Conclusion: The behavior at CDA of a metallic fueled core of a fast reactor was analyzed using the CDA initiation phase analysis code and the knowledge of the important characteristics at the CDA initiation phase was obtained

  6. Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.

    1990-11-01

    The conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design has been performed. As a first step, an intensive literature survey was completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins were designed and analyzed using the SIEX computer code. The analysis predicted that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors. 13 refs., 10 figs., 2 tabs.

  7. Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.

    1990-11-01

    The conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design has been performed. As a first step, an intensive literature survey was completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins were designed and analyzed using the SIEX computer code. The analysis predicted that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors. 13 refs., 10 figs., 2 tabs

  8. Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.

    1991-01-01

    In this paper, the conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design is performed. As a first step, an intensive literature survey is completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins is designed and analyzed using the SIEX computer code. The analysis predicts that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors

  9. Spent fuel metal storage cask performance testing and future spent fuel concrete module performance testing

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Creer, J.M.

    1988-10-01

    REA-2023 Gesellshaft fur Nuklear Service (GNS) CASTOR-V/21, Transnuclear TN-24P, and Westinghouse MC-10 metal storage casks, have been performance tested under the guidance of the Pacific Northwest Laboratory to determine their thermal and shielding performance. The REA-2023 cask was tested under Department of Energy (DOE) sponsorship at General Electric's facilities in Morris, Illinois, using BWR spent fuel from the Cooper Reactor. The other three casks were tested under a cooperative agreement between Virginia Power Company and DOE at the Idaho National Engineering Laboratory (INEL) by EGandG Idaho, Inc., using intact spent PWR fuel from the Surry reactors. The Electric Power Research Institute (EPRI) made contributions to both programs. A summary of the various cask designs and the results of the performance tests is presented. The cask designs include: solid and liquid neutron shields; lead, steel, and nodular cast iron gamma shields; stainless steel, aluminum, and copper baskets; and borated materials for criticality control. 4 refs., 8 figs., 6 tabs

  10. Investigation on fuel-cladding chemical interaction in metal fuel for FBR

    International Nuclear Information System (INIS)

    Inagaki, Kenta; Nakamura, Kinya; Ogata, Takanari; Uwaba, Tomoyuki

    2013-01-01

    During steady-state irradiation of metallic fuel in fast reactors, rare-earth fission products can react with stainless steel cladding at the fuel-cladding interface. The authors conducted isothermal annealing tests with some diffusion couples to investigate the structure of the wastage layer formed at the interface. Candidate cladding alloys, ferritic-martensitic steel (PNC-FMS) and oxide-dispersion-strengthened (ODS) steel were assembled with rare-earth alloys, RE5 : La-Ce-Pr-Nd-Sm, which simulate the fission yield of rare-earth fission products. The diffusion couples were isothermally annealed in the temperature range of 500-650°C for up to 170 h. In both RE5/ODS-steel and RE5/PNC-FMS couples, the wastage layer of the two-phase region of the (Fe, Cr) 17 RE 2 matrix phase with the precipitation of the (Fe, RE, Cr) phase was formed. The structure was similar to that formed in RE5/Fe-12Cr and RE5/HT9 couples, which implies that the reaction between REs and steel is not significantly influenced by the minor alloying elements within the candidate cladding materials. It was also clarified that the increase in the wastage layer thickness was diffusion-controlled. The temperature dependence of the reaction rate constants were formulated, which can be the basis for the quantification of the wastage layer growth. (author)

  11. Effects of accelerated degradation on metal supported thin film-based solid oxide fuel cell

    DEFF Research Database (Denmark)

    Reolon, R. P.; Sanna, S.; Xu, Yu

    2018-01-01

    A thin film-based solid oxide fuel cell is deposited on a Ni-based metal porous support by pulsed laser deposition with a multi-scale-graded microstructure design. The fuel cell, around 1 μm in thickness, is composed of a stabilized-zirconia/doped-ceria bi-layered dense electrolyte and nanostruct......A thin film-based solid oxide fuel cell is deposited on a Ni-based metal porous support by pulsed laser deposition with a multi-scale-graded microstructure design. The fuel cell, around 1 μm in thickness, is composed of a stabilized-zirconia/doped-ceria bi-layered dense electrolyte...

  12. Acoustic leak detection of LMFBR steam generator

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Yoshida, Kazuo

    1993-01-01

    The development of a water leak detector with short response time for LMFBR steam generators is required to prevent the failure propagation caused by the sodium-water reaction and to maintain structural safety in steam generators. The development of an acoustic leak detector assuring short response time has attracted. The purpose of this paper is to confirm the basic detection feasibility of the active acoustic leak detector, and to investigate the leak detection method by erasing the background noise by spectrum analysis of the passive acoustic leak detector. From a comparison of the leak detection sensitivity of the active and the passive method, the active method is not influenced remarkably by the background noise, and it has possibility to detect microleakage with short response time. We anticipate a practical application of the active method in the future. (author)

  13. Hockey-stick steam generator for LMFBR

    International Nuclear Information System (INIS)

    Hallinan, G.J.; Svedlund, P.E.

    1981-01-01

    This paper presents the criteria and evaluation leading to the selection of the Hockey Stick Steam Generator Concept and subsequent development of that concept for LMFBR application. The selection process and development of the Modular Steam Generator (MSG) is discussed, including the extensive test programs that culminated in the manufacture and test of a 35 MW(t) Steam Generator. The design of the CRBRP Steam Generator is described, emphasizing the current status and a review of the critical structural areas. CRBRP steam generator development tests are evaluated, with a discussion of test objectives and rating of the usefulness of test results to the CRBRP prototype design. Manufacturing experience and status of the CRBRP prototype and plant units is covered. The scaleup of the Hockey Stick concept to large commercial plant application is presented, with an evaluation of scaleup limitations, transient effects, and system design implications

  14. Low cycle fatigue of irradiated LMFBR materials

    International Nuclear Information System (INIS)

    Blackburn, L.D.

    1976-01-01

    A review of low cycle fatigue data on irradiated LMFBR materials was conducted and extensive graphical representations of available data are presented. Representative postirradiation tensile properties of annealed 304 and 316 SS are selected and employed in several predictive methods to estimate irradiated material fatigue curves. Experimental fatigue data confirm the use of predictive methods for establishing conservative design curves over the range of service conditions relevant to such CRBRP components as core former, fixed radial shielding, core barrel, lower inlet module and upper internals structures. New experimental data on fatigue curves and creep-fatigue interaction in irradiated 20 percent cold worked (CW) 316 SS and Alloy 718 would support the design of removable radial shielding and upper internals in CRBRP. New experimental information on notched fatigue behavior and cyclic stress-strain curves of all these materials in the irradiated condition could provide significant design data

  15. Microprocessor-based integrated LMFBR core surveillance

    International Nuclear Information System (INIS)

    Gmeiner, L.

    1984-06-01

    This report results from a joint study of KfK and INTERATOM. The aim of this study is to explore the advantages of microprocessors and microelectronics for a more sophisticated core surveillance, which is based on the integration of separate surveillance techniques. Due to new developments in microelectronics and related software an approach to LMFBR core surveillance can be conceived that combines a number of measurements into a more intelligent decision-making data processing system. The following techniques are considered to contribute essentially to an integrated core surveillance system: - subassembly state and thermal hydraulics performance monitoring, - temperature noise analysis, - acoustic core surveillance, - failure characterization and failure prediction based on DND- and cover gas signals, and - flux tilting techniques. Starting from a description of these techniques it is shown that by combination and correlation of these individual techniques a higher degree of cost-effectiveness, reliability and accuracy can be achieved. (orig./GL) [de

  16. Computational analysis of coolant mixing in subassembly and hot pool of an LMFBR

    International Nuclear Information System (INIS)

    Natesan, K.; Velusamy, K.; Kasinathan, N.; Clement Ravichandar, S.; Selvaraj, P.; Ghosh, D.; Chellapandi, P.; Chetal, S.C.

    2005-01-01

    Full text of publication follows: The reactor core of a typical Liquid Metal cooled Fast Breeder Reactor (LMFBR) comprises of several fuel, blanket, storage, reflector and shielding subassemblies (SA) arranged in a hexagonal geometry. Each fuel subassembly (FSA) comprises of 217 fuel pins packed with pellets of fissile and fertile materials. The wire wrapped pins are arranged in triangular pitch within a hexagonal wrapper. Due to the non uniform flow resistance across the cross section, sodium velocity is non uniform within SA. This non uniform velocity distribution coupled with radial power profile causes the sodium coolant to come out of the pin bundles of SA with a non uniform temperature profile. From the fuel pin bundle top, sodium flows axially through various regions comprising of a sodium plenum, shielding bundle region and SA head and adapter assembly. These regions contribute in the mixing of sodium in SA. Due to the non uniform temperature distribution at FSA top, the temperature profile at the possible locations of core monitoring thermocouple (TC) would also be non uniform. During a power campaign of core, FSA top drift outwards with respect to TC location due to irradiation bowing. This causes the temperature recorded by the TC to change in accordance with the temperature of sodium stream washing it from time to time during the power campaign. Apart from this, the non uniform temperature at the FSA exits can also lead to fluctuation in the temperature of sodium at TC locations. Knowledge of these phenomena in hot pool is essential in fixing the SCRAM threshold on the parameters that are derived from the measurements made by these TC. Analyses to study the above phenomena have been carried out in three stages. In the first stage, velocity and temperature distributions in the wire wrap bundle region of FSA have been obtained through subchannel analysis using the SUPER ENERGY code. In the second stage, 3D mixing of sodium in FSA in the region above the

  17. LMFBR source term experiments with rupture disk discharge under sodium

    International Nuclear Information System (INIS)

    Minges, J.; Schuetz, W.

    1993-05-01

    In the frame of the KfK research program FAUST, contributions are given to the assessment of the instantaneous source term in case of an LMFBR loss-of-flow accident with expanding fuel or sodium vapour. The main goal of the program is to achieve information, mainly by experiments, on the retention capability of the primary sodium pool for fuel and fission products. For that purpose, it is necessary to investigate the interaction of bubble and aerosol behaviour after a pressure discharge, and the subsequent aerosol transport. After a series of water tests (FAUST-1), rupture disk discharge tests under 500 C sodium up to 3.81 MPa were performed during the phase FAUST-2 with the two test facilities 2A (about 2 liters of sodium) and 2B (about 200 liters of sodium). The discharge tests were performed with pressurized argon gas and admixtures of the simulation materials Cs, Csl, Nal, l 2 , SrO, and UO 2 . Cs was a liquid, l 2 vapour, and all other substances solid particles. Besides UO 2 , non-radioactive material was used (natural isotopes). The retention capability of liquid sodium is expressed by retention factors RF. In general, RF is defined as the mass ratio of discharged amount, and the amount which is detected in the cover gas for the relevant species. From sampling immediately after the discharge, the instantaneous retention factors are deduced. From retarded sampling, the 'delayed factors' follow. High pressure discharge creates two important removal mechanisms, namely 'impaction by inertia relative to the bubble oscillations' and 'wash-out by sedimentation of entrained sodium droplets'. On the other hand, the retention of particles enclosed in a buoyantly rising bubble is significantly smaller. (orig.) [de

  18. Development of eutectic free cladding materials for metallic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tokiwai, Moriyasu; Yuda, Ryoichi [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan); Ohuchi, Atsushi [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan); Amaya, Masaki [Global Nuclear Fuel-Japan Co., Ltd, Oarai, Ibaraki (Japan)

    2002-11-01

    Historically, it is well known that U base metallic fuel has a lower eutectic temperature with stainless steel cladding. In the phase diagram for the U-Fe binary system, the eutectic temperature is 998K. The eutectic reaction is a limiting factor for raising reactor operation temperature. For the purpose of development of eutectic-free cladding materials, three kinds of diffusion-couple tests with 10 mass%Zr alloy were conducted at a temperature of 1027K for 2250 hrs. We selected the following materials: (a) nitrogen charged zirconium foils, (b) vanadium foils of commercial grade, and (c) nitrogen charged ferritic stainless steel (HT-9). The results showed that typical Zr with layer was observed in all of these materials. Zr with layer appeared to act as a barrier against inter-diffusion of U, Fe. The barrier provided immunity to the eutectic reaction. Discussion was made on C-14 problems in relation to another desirable thermodynamic characteristics of Zr such as carbon-14 immobilization. EPMA analysis indicated relatively high nitrogen concentration at the barrier. The barrier is probably composed of ZrN. (author)

  19. Doped Graphene as Non-Metallic Catalyst for Fuel Cells

    Directory of Open Access Journals (Sweden)

    Adriana MARINOIU

    2017-08-01

    Full Text Available Aiming a commercial development of proton exchange membrane fuel cells (PEMFC, a low cost, sustainable and high performance electrocatalyst for oxygen reduction reaction (ORR with capability to replace/reduce rare metals, are high desirable. In this paper, we present a class of doped graphene, namely iodinated graphene with highly ORR electrochemical performances, synthesized by using the electrophilic substitution method. The prepared samples were characterized by different techniques, including Scanning Electron Microscopy SEM, X-ray photoelectron spectroscopy XPS, Raman spectroscopy, surface area measurement by BET method, that revealed the structure and morphology. The most highly iodinated graphene was tested in a single cell by measuring the cyclic voltammetry. The electrochemical performances were evaluated and compared with a typical PEMFC configuration, when a single cathodic peak at 0.2 V with a current density of – 3.67 mA cm-2 for the Pt/C electrode was obtained. The best electrochemical performances in terms of electrochemical active area, was obtained for a new concept of cathode composed from Pt/C – iodine doped graphene, when a well-defined peak centred at 0.23 V with a current density of approx. – 9.1 mA cm-2 was obtained, indicating a high catalytic activity for ORR.DOI: http://dx.doi.org/10.5755/j01.ms.23.2.16216

  20. Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    William J. Carmack

    2012-05-01

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

  1. In-pile measurement of the thermal conductivity of irradiated metallic fuel

    International Nuclear Information System (INIS)

    Bauer, T.H.; Holland, J.W.

    1995-01-01

    Transient test data and posttest measurements from recent in-pile overpower transient experiments are used for an in situ determination of metallic fuel thermal conductivity. For test pins that undergo melting but remain intact, a technique is described that relates fuel thermal conductivity to peak pin power during the transient and a posttest measured melt radius. Conductivity estimates and their uncertainty are made for a database of four irradiated Integral Fast Reactor-type metal fuel pins of relatively low burnup (<3 at.%). In the assessment of results, averages and trends of measured fuel thermal conductivity are correlated to local burnup. Emphasis is placed on the changes of conductivity that take place with burnup-induced swelling and sodium logging. Measurements are used to validate simple empirically based analytical models that describe thermal conductivity of porous media and that are recommended for general thermal analyses of irradiated metallic fuel

  2. A direct borohydride fuel cell with a polymer fiber membrane and non-noble metal catalysts

    OpenAIRE

    Yang, Xiaodong; Liu, Yongning; Li, Sai; Wei, Xiaozhu; Wang, Li; Chen, Yuanzhen

    2012-01-01

    Polymer electrolyte membranes (PEM) and Pt-based catalysts are two crucial components which determine the properties and price of fuel cells. Even though, PEM faces problem of fuel crossover in liquid fuel cells such as direct methanol fuel cell (DMFC) and direct borohydride fuel cell (DBFC), which lowers power output greatly. Here, we report a DBFC in which a polymer fiber membrane (PFM) was used, and metal oxides, such as LaNiO3 and MnO2, were used as cathode catalysts, meanwhile CoO was us...

  3. Status of liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    1985-01-01

    This document represents a compilation of the information on the status of fast breeder reactor development. It is intended to provide complete and authoritative information for academic, energy, industrial and planning organizations in the IAEA Member States. The Report also provides extended reference and bibliography lists. A summarized overview of the national programmes of LMFBR development is given in Chapter II. Chapter III on LMFBR experience provides a brief description and purpose of all fast reactors - experimental, demonstration and commercial size - that have been or are planned for construction and operation. Fast reactor physics is dealt with in Chapter IV. Besides the basic facts and definitions of neutronics and the compilation and measurement of nuclear data, a broad range of the calculation methods, codes, and the state of the art is described. In Chapter V, fuels and materials are described. The emphasis is on the design and development experience gained with mixed oxide fuel pins and subassemblies. Structural materials, blanket elements and absorber materials are also discussed. Chaper VI presents a broad overview of the technical and engineering aspects of LMFBR power plants. LMFBR core design is described in detail, followed by the components of the main heat transport system, the refuelling equipment, and auxiliary systems. Chapter VII on safety is a compilation of the current safety design concepts of LMFBRs and new trends in safety criteria and safety goals. The chapter concludes with risk analyses of LMFBR technology. In Chapter VIII, the systems approach has been emphasized in the consideration of the whole LMFBR fuel cycle. Special emphasis is placed on safeguards aspects and the environmental impact of the LMFBR fuel cycle. Chapter IX describes deployment considerations of LMFBRs. Special emphasis is placed on economic aspects of the LMFBR power plant and its related fuel cycle. Finally, Chapter X provides an overall summary and a

  4. Corrosion resistance of metallic materials for use in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Pinard Legry, G.; Pelras, M.; Turluer, G.

    1988-01-01

    The main of this review is to reassess the corrosion resistance properties required from metallic materials to be used in the various developments of the Purex process applied for nuclear fuel reprocessing

  5. Pumped lithium loop test to evaluate advanced refractory metal alloys and simulated nuclear fuel elements

    Science.gov (United States)

    Brandenburf, G. P.; Hoffman, E. E.; Smith, J. P.

    1974-01-01

    The performance was determined of refractory metal alloys and uranium nitride fuel element specimens in flowing 1900F (1083C) lithium. The results demonstrate the suitability of the selected materials to perform satisfactorily from a chemical compatibility standpoint.

  6. Dimensional, microstructural and compositional stability of metal fuels. Final performance report

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, A.A.; Dayananda, M.A.

    1993-03-15

    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.

  7. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    Energy Technology Data Exchange (ETDEWEB)

    SCHWINKENDORF, K.N.

    2006-05-12

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can

  8. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE and AFTER IRRADIATION

    International Nuclear Information System (INIS)

    SCHWINKENDORF, K.N.

    2006-01-01

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k eff = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can supply useful

  9. 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

  10. Wire-wrapped rod-bundle heat-transfer analysis for LMFBR

    International Nuclear Information System (INIS)

    Wong, C.N.C.; Todreas, N.E.

    1982-07-01

    Helical wire wraps are widely used in the LMFBR fuel and blanket assemblies to provide coolant mixing and maintain proper spacing between fuel pins. The presence of the helical wire, however, may possibly induce heat transfer problems, such as the uncertainty of the maximum clad temperature as a result of the contact between the wires and the pins. In this study, the detailed transient three dimensional velocity and temperature distributions for the coolant around the pin will be determined by solving the governing momentum and energy equation numerically. A computer code HEATRAN has been developed to perform this calculation. Before the computer code HEATRAN is applied to the wire wrapped rod bundle problem, it is used to analyze a wide range of fluid and heat transfer problem to verify its capabilities

  11. Status of the LMFBR thermo- and fluid-dynamic activities at KFK

    International Nuclear Information System (INIS)

    Hoffmann, H.; Hofmann, F.; Rehme, K.

    1979-01-01

    The aim of the thermo- and fluiddynamic analysis is to determine the spatial velocity and temperature distributions in LMFBR-core elements with high accuracy. Knowledge of these data is a necessary prerequisite for determining the mechanical behavior of fuel rods and of structural material. Three cases are distinguished: Nominal geometry and steady state conditions; non-nominal geometry and quasi-steady state conditions; nominal geometry and non-steady state conditions. The present situation for the design calculations of fuel elements is based mainly on undisturbed normal operation. Most of the thermo- and fluiddynamic activities performed under the Fast Breeder Programme at KFK are related to this case. The present status of theoretical and experimental research work briefly presented in this paper, can be subdivided into the following main topics: 1. Physical and mathematical modelling of single phase rod bundle thermo- and fluiddynamics, 2. Experimental investigations on heat transfer and fluid flow in rod bundles

  12. TRANSENERGY S: computer codes for coolant temperature prediction in LMFBR cores during transient events

    International Nuclear Information System (INIS)

    Glazer, S.; Todreas, N.; Rohsenow, W.; Sonin, A.

    1981-02-01

    This document is intended as a user/programmer manual for the TRANSENERGY-S computer code. The code represents an extension of the steady state ENERGY model, originally developed by E. Khan, to predict coolant and fuel pin temperatures in a single LMFBR core assembly during transient events. Effects which may be modelled in the analysis include temporal variation in gamma heating in the coolant and duct wall, rod power production, coolant inlet temperature, coolant flow rate, and thermal boundary conditions around the single assembly. Numerical formulations of energy equations in the fuel and coolant are presented, and the solution schemes and stability criteria are discussed. A detailed description of the input deck preparation is presented, as well as code logic flowcharts, and a complete program listing. TRANSENERGY-S code predictions are compared with those of two different versions of COBRA, and partial results of a 61 pin bundle test case are presented

  13. Thermochemical treatment of radioactive waste by using powder metal fuels

    International Nuclear Information System (INIS)

    Dmitriev, S.A.; Ojovan, M.I.; Karlina, O.K.

    2001-01-01

    Full text: A thermochemical approach was suggested for treating and conditioning specific streams of radioactive wastes for example spent ion exchange resins, mixed, organic or chlorine-containing radioactive waste as well as in order to decontaminate heavily contaminated surfaces. Conventional treatment methods of such waste encounters serious problems concerning complete destruction of organic molecules and possible emissions of radionuclides, heavy metals and chemically hazardous species or in case of contaminated materials - complete removal of contamination from surface. The thermochemical treatment of radioactive waste uses powdered metal fuels (PMF) that are specifically formulated for the waste composition and react chemically with the waste components. Thermochemical treatment technologies use the energy of chemical reactions in the mixture of waste with PMF to sustain both decomposition and synthesis processes as well as processes of isomorphic substitutions of hazardous elements into stable mineral forms. The composition of the PMF is designed in such a way as to minimise the release of hazardous components and radionuclides in the off gas and to confine the contaminants in the mineral or glass like final products. The thermochemical procedures allow decomposition of organic matter and capturing hazardous radionuclides and chemical species simultaneously. Thermochemical treatment technologies are very efficient, easy to apply, they have low capital investment and can be used both at large and small facilities. An advantage of thermochemical technologies is their autonomy. Thus these technologies can be successfully applied in order to treat small amount of waste without usage of complex and expensive equipment. They can be used also in emergency situations. Currently the thermochemical treatment technologies were developed and demonstrated to be feasible as follows: 1. Decontamination of surfaces; 2. Processing of organic waste; 3. Vitrification of dusty

  14. Development of advanced spent fuel management process / criticality safety analysis for integrated mockup and metallized spent fuel storage

    International Nuclear Information System (INIS)

    Ro, Seong Gy; Shin, Hee Sung; Shin, Young Joon; Bae, Kang Mok

    1999-02-01

    Benchmark calculation for SCALE4.3 CSAS6 module and burnup credit criticality analysis performed by CSAS6 module are described in this report. Calculation biases by the SCALE4.3 CSAS6 module for PWR spent fuel, metallized spent fuel and aqueous nuclear materials have been determined on the basis of the benchmark to be 0.011, 0.023 and 0.010, respectively. The maximum allowable multiplication factor for an integrated mockup and metallized spent fuel storage is conservatively determined to be 0.927. With the aid of this code system, K eff values as a function of metallization ratio for the integrated mockup have been calculated. The maximum values of K eff for normal and hypothetical accident conditions are 0.346 and 0.598, respectively, much less than the maximum allowable multiplication factor of 0.927. Besides, burnup credit criticality analysis has been performed for infinite arrays of square and hexagonal canisters containing metallized spent fuel rods with different canister wall thickness, canister surface-to-surface distance and water content. It is revealed that the effective multiplication factor for canister arrays as mentioned above is well below the subcritical limit regardless of external conditions when its wall thickness is over 9 mm. (Author). 37 refs., 27 tabs., 64 figs

  15. Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Chan Bock Lee

    2016-10-01

    Full Text Available Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  16. Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  17. Model-based temperature noise monitoring methods for LMFBR core anomaly detection

    International Nuclear Information System (INIS)

    Tamaoki, Tetsuo; Sonoda, Yukio; Sato, Masuo; Takahashi, Ryoichi.

    1994-01-01

    Temperature noise, measured by thermocouples mounted at each core fuel subassembly, is considered to be the most useful signal for detecting and locating local cooling anomalies in an LMFBR core. However, the core outlet temperature noise contains background noise due to fluctuations in the operating parameters including reactor power. It is therefore necessary to reduce this background noise for highly sensitive anomaly detection by subtracting predictable components from the measured signal. In the present study, both a physical model and an autoregressive model were applied to noise data measured in the experimental fast reactor JOYO. The results indicate that the autoregressive model has a higher precision than the physical model in background noise prediction. Based on these results, an 'autoregressive model modification method' is proposed, in which a temporary autoregressive model is generated by interpolation or extrapolation of reference models identified under a small number of different operating conditions. The generated autoregressive model has shown sufficient precision over a wide range of reactor power in applications to artificial noise data produced by an LMFBR noise simulator even when the coolant flow rate was changed to keep a constant power-to-flow ratio. (author)

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

    Science.gov (United States)

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

    2012-06-01

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

  19. Simplified process for leaching precious metals from fuel cell membrane electrode assemblies

    Science.gov (United States)

    Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ

    2009-12-22

    The membrane electrode assemblies of fuel cells are recycled to recover the catalyst precious metals from the assemblies. The assemblies are cryogenically embrittled and pulverized to form a powder. The pulverized assemblies are then mixed with a surfactant to form a paste which is contacted with an acid solution to leach precious metals from the pulverized membranes.

  20. Magnetron sputtered gadolinia-doped ceria diffusion barriers for metal-supported solid oxide fuel cells

    DEFF Research Database (Denmark)

    Sønderby, Steffen; Klemensø, Trine; Christensen, Bjarke H.

    2014-01-01

    Gadolinia-doped ceria (GDC) thin films are deposited by reactive magnetron sputtering in an industrial-scale setup and implemented as barrier layers between the cathode and electrolyte in metal-based solid oxide fuel cells consisting of a metal support, an electrolyte of ZrO2 co-doped with Sc2O3 ...

  1. Break-down of Losses in High Performing Metal-Supported Solid Oxide Fuel Cells

    DEFF Research Database (Denmark)

    Kromp, Alexander; Nielsen, Jimmi; Blennow Tullmar, Peter

    2012-01-01

    Metal supported SOFC designs offer competitive advantages such as reduced material costs and improved mechanical robustness. On the other hand, disadvantages might arise due to possible corrosion of the porous metal parts during processing and operation at high fuel utilization. In this paper we...

  2. Electrometallurgical treatment of metallic spent nuclear fuel stored at the Hanford Site

    International Nuclear Information System (INIS)

    Laidler, J.J.; Gay, E.C.

    1996-01-01

    The major component of the DOE spent nuclear fuel inventory is the metallic fuel stored at the Hanford site in the southeastern part of the state of Washington. Most of this fuel was discharged from the N-Reactor; a small part of the inventory is fuel from the early Hanford production reactors. The U.S. Department of Energy (DOE) plans to remove these fuels from the spent fuel storage pools in which they are presently stored, dry them, and place them in interim storage at a location at the Hanford site that is far removed from the Columbia River. It is not yet certain that these fuels will be acceptable for disposal in a mined geologic repository without further treatment, due to their potential pyrophoric character. A practical method for treatment of the Hanford metallic spent fuel, based on an electrorefining process, has been developed and has been demonstrated with unirradiated N-Reactor fuel and with simulated single-pass reactor (SPR) spent fuel. The process can be operated with any desired throughput rates; being a batch process, it is simply a matter of setting the size of the electrorefiner modules and the number of such modules. A single module, prototypic of a production-scale module, has been fabricated and testing is in progress at a throughput rate of 150 kg (heavy metal) per day. The envisioned production version would incorporate additional anode baskets and cathode tubes and provide a throughput rate of 333 kgHM/day. A system with four of these modules would permit treatment of Hanford metallic fuels at a rate of at least 250 metric tons per year

  3. Comments on US LMFBR steam generator base technology

    International Nuclear Information System (INIS)

    Simmons, W.R.

    1984-01-01

    The development of steam generators for the LMFBR was recognized from the onset by the AEC, now DOE, as a difficult, challenging, and high-priority task. The highly reactive nature of sodium with water/steam requires that the sodium-water/steam boundaries of LMFBR steam generators possess a degree of leak-tightness reliability not normally attempted on a commercial scale. In addition, the LMFBR steam generator is subjected to high fluid temperatures and severe thermal transients. These requirements place great demand on materials, fabrication processes, and inspection methods; and even greater demands on the designer to provide steam generators that can meet these demanding requirements, be fabricated without unreasonable shop requirements, and tolerate off-normal effects

  4. Hydrogen formation and control under postulated LMFBR accident conditions

    International Nuclear Information System (INIS)

    Armstrong, G.R.; Wierman, R.W.

    1976-09-01

    The objective of this study is to experimentally investigate the potential for autoignition and combustion of hydrogen-sodium mixtures which may be produced in LMFBR accidents. The purpose and ultimate usefulness of this work is to provide data that will establish the validity and acceptability of mechanisms inherent to the LMFBR that could either prevent or delay the accumulation of hydrogen gas to less than 4 percent (V) in the Reactor Containment Building (RCB) under accident conditions. The results to date indicate that sodium and sodium-hydrogen mixtures such as may be expected during LMFBR postulated accidents will ignite upon entering an air atmosphere and that the hydrogen present will be essentially all consumed until such time that the oxygen concentration is depleted

  5. A direct borohydride fuel cell with a polymer fiber membrane and non-noble metal catalysts.

    Science.gov (United States)

    Yang, Xiaodong; Liu, Yongning; Li, Sai; Wei, Xiaozhu; Wang, Li; Chen, Yuanzhen

    2012-01-01

    Polymer electrolyte membranes (PEM) and Pt-based catalysts are two crucial components which determine the properties and price of fuel cells. Even though, PEM faces problem of fuel crossover in liquid fuel cells such as direct methanol fuel cell (DMFC) and direct borohydride fuel cell (DBFC), which lowers power output greatly. Here, we report a DBFC in which a polymer fiber membrane (PFM) was used, and metal oxides, such as LaNiO₃ and MnO₂, were used as cathode catalysts, meanwhile CoO was used as anode catalyst. Peak power density of 663 mW·cm⁻² has been achieved at 65°C, which increases by a factor of 1.7-3.7 compared with classic DBFCs. This fuel cell structure can also be extended to other liquid fuel cells, such as DMFC.

  6. A direct borohydride fuel cell with a polymer fiber membrane and non-noble metal catalysts

    Science.gov (United States)

    Yang, Xiaodong; Liu, Yongning; Li, Sai; Wei, Xiaozhu; Wang, Li; Chen, Yuanzhen

    2012-08-01

    Polymer electrolyte membranes (PEM) and Pt-based catalysts are two crucial components which determine the properties and price of fuel cells. Even though, PEM faces problem of fuel crossover in liquid fuel cells such as direct methanol fuel cell (DMFC) and direct borohydride fuel cell (DBFC), which lowers power output greatly. Here, we report a DBFC in which a polymer fiber membrane (PFM) was used, and metal oxides, such as LaNiO3 and MnO2, were used as cathode catalysts, meanwhile CoO was used as anode catalyst. Peak power density of 663 mW.cm-2 has been achieved at 65°C, which increases by a factor of 1.7-3.7 compared with classic DBFCs. This fuel cell structure can also be extended to other liquid fuel cells, such as DMFC.

  7. A direct borohydride fuel cell with a polymer fiber membrane and non-noble metal catalysts

    Science.gov (United States)

    Yang, Xiaodong; Liu, Yongning; Li, Sai; Wei, Xiaozhu; Wang, Li; Chen, Yuanzhen

    2012-01-01

    Polymer electrolyte membranes (PEM) and Pt-based catalysts are two crucial components which determine the properties and price of fuel cells. Even though, PEM faces problem of fuel crossover in liquid fuel cells such as direct methanol fuel cell (DMFC) and direct borohydride fuel cell (DBFC), which lowers power output greatly. Here, we report a DBFC in which a polymer fiber membrane (PFM) was used, and metal oxides, such as LaNiO3 and MnO2, were used as cathode catalysts, meanwhile CoO was used as anode catalyst. Peak power density of 663 mW·cm−2 has been achieved at 65°C, which increases by a factor of 1.7–3.7 compared with classic DBFCs. This fuel cell structure can also be extended to other liquid fuel cells, such as DMFC. PMID:22880160

  8. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    International Nuclear Information System (INIS)

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

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

  9. Study on uranium metallization yield of spent Pressurized Water Reactor fuels and oxidation behavior of fission products in uranium metals

    International Nuclear Information System (INIS)

    Choi, Ke Chon; Lee, Chang Heon; Kim, Won Ho

    2003-01-01

    Metallization yield of uranium oxide to uranium metal from lithium reduction process of spent Pressurized Water Reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metallization yield was measured. Metallization yield of the solid part was 90.7∼95.9 wt%, and the powder being 77.8∼71.5 wt% individually. Oxidation behaviour of the quarternary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At 600∼700 .deg. C, weight increments of allow of No, Ru, Rh and Pd was 0.40∼0.55 wt%. Phase change on the surface of the allow was started at 750 .deg. C. In particular, Mo was rapidly oxidized and then the alloy lost 0.76∼25.22 wt% in weight

  10. Corrosion of metal bipolar plates for PEM fuel cells: A review

    Energy Technology Data Exchange (ETDEWEB)

    Antunes, Renato A. [Engenharia de Materiais, Universidade Federal do ABC (UFABC), 09210-170 Santo Andre, SP (Brazil); Oliveira, Mara Cristina L.; Ett, Gerhard; Ett, Volkmar [Electrocell Ind. Com. Equip. Elet. LTDA, Centro de Inovacao, Empreendedorismo e Tecnologia (CIETEC), 05508-000 Sao Paulo, SP (Brazil)

    2010-04-15

    PEM fuel cells are of prime interest in transportation applications due to their relatively high efficiency and low pollutant emissions. Bipolar plates are the key components of these devices as they account for significant fractions of their weight and cost. Metallic materials have advantages over graphite-based ones because of their higher mechanical strength and better electrical conductivity. However, corrosion resistance is a major concern that remains to be solved as metals may develop oxide layers that increase electrical resistivity, thus lowering the fuel cell efficiency. This paper aims to present the main results found in recent literature about the corrosion performance of metallic bipolar plates. (author)

  11. Method for preparing metal powder, device for preparing metal powder, method for processing spent nuclear fuel

    Science.gov (United States)

    Park, Jong-Hee [Clarendon Hills, IL

    2011-11-29

    A method for producing metal powder is provided the comprising supplying a molten bath containing a reducing agent, contacting a metal oxide with the molten bath for a time and at a temperature sufficient to reduce the metal in the metal oxide to elemental metal and produce free oxygen; and isolating the elemental metal from the molten bath.

  12. Sensitivity Analysis of FEAST-Metal Fuel Performance Code: Initial Results

    Energy Technology Data Exchange (ETDEWEB)

    Edelmann, Paul Guy [Los Alamos National Laboratory; Williams, Brian J. [Los Alamos National Laboratory; Unal, Cetin [Los Alamos National Laboratory; Yacout, Abdellatif [Argonne National Laboratories

    2012-06-27

    This memo documents the completion of the LANL milestone, M3FT-12LA0202041, describing methodologies and initial results using FEAST-Metal. The FEAST-Metal code calculations for this work are being conducted at LANL in support of on-going activities related to sensitivity analysis of fuel performance codes. The objective is to identify important macroscopic parameters of interest to modeling and simulation of metallic fuel performance. This report summarizes our preliminary results for the sensitivity analysis using 6 calibration datasets for metallic fuel developed at ANL for EBR-II experiments. Sensitivity ranking methodology was deployed to narrow down the selected parameters for the current study. There are approximately 84 calibration parameters in the FEAST-Metal code, of which 32 were ultimately used in Phase II of this study. Preliminary results of this sensitivity analysis led to the following ranking of FEAST models for future calibration and improvements: fuel conductivity, fission gas transport/release, fuel creep, and precipitation kinetics. More validation data is needed to validate calibrated parameter distributions for future uncertainty quantification studies with FEAST-Metal. Results of this study also served to point out some code deficiencies and possible errors, and these are being investigated in order to determine root causes and to improve upon the existing code models.

  13. Validation of the metal fuel version of the SAS4A accident analysis code

    International Nuclear Information System (INIS)

    Tentner, A.M.

    1991-01-01

    This paper describes recent work directed towards the validation of the metal fuel version of the SAS4A accident analysis code. The SAS4A code system has been developed at Argonne National Laboratory for the simulation of hypothetical severe accidents in Liquid Metal-Cooled Reactors (LMR), designed to operate in a fast neutron spectrum. SAS4A was initially developed for the analysis of oxide-fueled liquid metal-cooled reactors and has played an important role in the simulation and assessment of the energetics potential for postulated severe accidents in these reactors. Due to the current interest in the metal-fueled liquid metal-cooled reactors, a metal fuel version of the SAS4A accident analysis code is being developed in the Integral Fast Reactor program at Argonne. During such postulated accident scenarios as the unprotected (i.e. without scram) loss-of-flow and transient overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling, and fuel and cladding melting and relocation. Due to strong neutronic feedbacks these events can significantly influence the reactor power history in the accident progression. The paper presents the results of a recent SAS4A simulation of the M7 TREAT experiment. 6 refs., 5 figs

  14. Modified Fabrication Method of Metal Fuel Slug for Preventing Evaporation of Volatile Elements

    International Nuclear Information System (INIS)

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

    2014-01-01

    In order to develop innovative fabrication method of metal fuel slugs 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, U-10wt.%Zr-Mn fuel slugs for SFR have been fabricated by modified casting method and characterized to evaluate the feasibility of the alternative fabrication method. Alternative casting such as modified casting has been applied to develop fabrication method of fuel slugs for preventing the evaporation of volatile elements such as Am. U-10wt.%Zr-Mn containing a volatile surrogate Mn fuel slug was soundly fabricated under inert atmosphere with dimensions of L250mm. Mass fraction of fuel loss was so low, upto 0.2%. Mn element was most recovered with prevention in evaporation of Mn. It was seen that the losses of volatile Am can be effectively controlled to below detectable levels using modest pressure

  15. Applications of simulation experiments in LMFBR core materials technology

    International Nuclear Information System (INIS)

    Appleby, W.K.

    1976-01-01

    The development of charged particle bombardment experiments to simulate neutron irradiation induced swelling in austenitic alloys is briefly described. The applications of these techniques in LMFBR core materials technology are discussed. It is shown that use of the techniques to study the behavior of cold-worked Type-316 was instrumental in demonstrating at an early date the need for advanced materials. The simulation techniques then were used to identify alloying elements which can markedly decrease swelling and thus a focused reactor irradiation program is now in place to allow the future use of a lower swelling alloy for LMFBR core components

  16. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kurosawa, Katsutoshi.

    1984-01-01

    Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

  17. Intelligent type sodium instrumentations for LMFBR

    International Nuclear Information System (INIS)

    Chen Daolong

    1996-07-01

    The constructions and performances of lots of newly developed intelligent type sodium instrumentations are described. The graduation characteristic equations for corresponding transducer using the medium temperature as the parameter are given. These intelligent type sodium instrumentations are possessed of good linearity. The accurate measurement data of sodium process parameters (flowrate, pressure and level) can be obtained by means of their on-line compensation function of the temperature effect. Moreover, these intelligent type sodium instrumentations are possessed of the self-inspection, the electric shutoff protection, the setting of full-scale, the setting of alarm limits (two upper limits and two lower limits alarms), the thermocouple breaking alarm, mutual isolative the 0∼10 V direct-current analogue output and the CENTRONICS standard digital output, and the alarm relay contact output. Theses intelligent type sodium instrumentations are suitable particularly for the instrument, control and protective systems of LMFBR by means of these excellent functions based on microprocessor. The basic errors of the intelligent type sodium flowmeter, immersed sodium flowmeter, sodium manometer and sodium level gauge are +-2%, +-2.3%, +-0.3% and +-1.9% of measuring ranges respectively. (9 figs.)

  18. Coolant mixing in the LMFBR outlet plenum

    International Nuclear Information System (INIS)

    Chen, Y.B.; Golay, M.W.

    1977-06-01

    Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds Number (Re) values of 33000 and 70000 in a 1/15-scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet velocity field, upon the degree of inlet turbulence, and upon the turbulence momentum exchange model used in the calculations. It is found in the FFTF geometry that the TEACH-T predictions are better than that of VARR-II, and in the CRBR geometry neither code provides a good prediction of the observed behavior. From the sensitivity analysis, it is found that the production and dissipation of turbulence are the dominant terms in the transport equations for turbulent kinetic energy and turbulent energy dissipation rate, and the diffusion terms are relatively small. From the same study a new set of empirical constants for the turbulence model is evolved for the prediction of plenum flows

  19. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    Energy Technology Data Exchange (ETDEWEB)

    TOFFER, H.

    2006-07-18

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel

  20. Overview of the fast reactors fuels program

    International Nuclear Information System (INIS)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

  1. HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

    Energy Technology Data Exchange (ETDEWEB)

    Sell, D. A.; Baily, C. E.; Malewitz, T. J.; Medvedev, P. G.; Porter, D. L.; Hilton, B. A.

    2016-09-01

    A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium after the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.

  2. Measuring the noble metal and iodine composition of extracted noble metal phase from spent nuclear fuel using instrumental neutron activation analysis

    International Nuclear Information System (INIS)

    Palomares, R.I.; Dayman, K.J.; Landsberger, S.; Biegalski, S.R.; Soderquist, C.Z.; Casella, A.J.; Brady Raap, M.C.; Schwantes, J.M.

    2015-01-01

    Masses of noble metal and iodine nuclides in the metallic noble metal phase extracted from spent fuel are measured using instrumental neutron activation analysis. Nuclide presence is predicted using fission yield analysis, and radionuclides are identified and the masses quantified using neutron activation analysis. The nuclide compositions of noble metal phase derived from two dissolution methods, UO 2 fuel dissolved in nitric acid and UO 2 fuel dissolved in ammonium-carbonate and hydrogen-peroxide solution, are compared. - Highlights: • The noble metal phase was chemically extracted from spent nuclear fuel and analyzed non-destructively. • Noble metal phase nuclides and long-lived iodine were identified and quantified using neutron activation analysis. • Activation to shorter-lived radionuclides allowed rapid analysis of long-lived fission products in spent fuel using gamma spectrometry

  3. PROCESSING OF URANIUM-METAL-CONTAINING FUEL ELEMENTS

    Science.gov (United States)

    Moore, R.H.

    1962-10-01

    A process is given for recovering uranium from neutronbombarded uranium- aluminum alloys. The alloy is dissolved in an aluminum halide--alkali metal halide mixture in which the halide is a mixture of chloride and bromide, the aluminum halide is present in about stoichiometric quantity as to uranium and fission products and the alkali metal halide in a predominant quantity; the uranium- and electropositive fission-products-containing salt phase is separated from the electronegative-containing metal phase; more aluminum halide is added to the salt phase to obtain equimolarity as to the alkali metal halide; adding an excess of aluminum metal whereby uranium metal is formed and alloyed with the excess aluminum; and separating the uranium-aluminum alloy from the fission- productscontaining salt phase. (AEC)

  4. Liquid Metal Fast Breeder Reactor program. Volume IV. Environmental statement

    International Nuclear Information System (INIS)

    1974-12-01

    A broad overview is presented of the many implications of LMFBR program implementation, up to and encompassing a fully developed LMFBR power plant economy, including the secondary impacts, the unavoidable adverse environmental impacts, cumulative environmental impacts, and cost-benefit analyses, and alternative energy strategies. Under the heading of secondary impacts, the national implications of the availability of electricity from LMFBRs, and the specific economic impacts of the LMFBR program are examined. The currently feasible alternatives and potential future alternatives for mitigating adverse environmental impacts of the LMFBR fuel cycle are described. The problems of safeguarding special nuclear material from potential diversion to unauthorized purposes are analyzed. The cumulative environmental effects of LMFBR operation to the Year 2020, the decommissioning of LMFBRs and fuel cycle facilities upon the completion of their useful life, the irreversible and irretrievable commitments of resources that will accompany implementation of an LMFBR economy, and an analysis of the costs and benefits of implementing the LMFBR Program are included. (U.S.)

  5. Safety criteria for the future LMFBR's in France and main safety issues for the rapide 1500 project

    International Nuclear Information System (INIS)

    Justin, F.; Natta, M.; Orzoni, G.

    1985-04-01

    The main safety criteria for future LMFBR in France and the related issues for the RAPIDE 1500 project are presented and discussed. The evolutions with respect to SUPERPHENIX options and requirements are emphasized, in particular for the concerns of the prevention of core melt accidents, fuel damage limits and related required performances of the protection system, since one main option is not to consider whole core melt accidents in the containment design. One shall also point out the advantages of some mitigating features which were nevertheless added in the containment design, although without any explicit consideration for core melt accidents

  6. Characterization of Irradiated Metal Waste from the Pyrometallurgical Treatment of Used EBR-II Fuel

    Energy Technology Data Exchange (ETDEWEB)

    B.R. Westphal; K.C. Marsden; W.M. McCartin; S.M. Frank; D.D. Keiser, Jr.; T.S. Yoo; D. Vaden; D.G. Cummings; K.J. Bateman; J. J. Giglio; T. P. O' Holleran; P. A. Hahn; M. N. Patterson

    2013-03-01

    As part of the pyrometallurgical treatment of used Experimental Breeder Reactor-II fuel, a metal waste stream is generated consisting primarily of cladding hulls laden with fission products noble to the electrorefining process. Consolidation by melting at high temperature [1873 K (1600 degrees C)] has been developed to sequester the noble metal fission products (Zr, Mo, Tc, Ru, Rh, Te, and Pd) which remain in the iron-based cladding hulls. Zirconium from the uranium fuel alloy (U-10Zr) is also deposited on the hulls and forms Fe-Zr intermetallics which incorporate the noble metals as well as residual actinides during processing. Hence, Zr has been chosen as the primary indicator for consistency of the metal waste. Recently, the first production-scale metal waste ingot was generated and sampled to monitor Zr content for Fe-Zr intermetallic phase formation and validation of processing conditions. Chemical assay of the metal waste ingot revealed a homogeneous distribution of the noble metal fission products as well as the primary fuel constituents U and Zr. Microstructural characterization of the ingot confirmed the immobilization of the noble metals in the Fe-Zr intermetallic phase.

  7. Experimental specifications for eutectic reaction between metallic fuel and HT-9

    International Nuclear Information System (INIS)

    Hwang, Woan; Nam, Cheol; Lee, Byoung Oon; Ryu, Woo Seog

    1998-10-01

    The chemical interaction between metallic fuel and cladding is important in designing the fuel pin of the KALIMER. When metal fuel and cladding are contacted, the elements in fuel and cladding are inter-diffuse each other, forming the reaction layers at interface. The reaction layers may cause two important factors in aspects of fuel pin integrity. Firstly, it degrades cladding strength by reducing effective cladding thickness. Secondly, these layers accelerate eutectic reaction at transient conditions. To evaluate these phenomena, the diffusion couple experiment is planned by using metal fuels with various zirconium contents and HT-9 steel. The U-Zr fuel alloys will be used for the experiment with the different zirconium contents, these are 8, 10 and 12 weight %. This experiment aims to evaluate the effects of zirconium content on the chemical reaction. Furthermore, the reaction rate and threshold temperature of the eutectic melting will be determined as a function of the zirconium content. This document describes the detail experimental specifications for the eutectic reaction such as test setup, test requirements and test procedure. (author). 10 refs

  8. Comparative analysis of a hypothetical loss-of-flow accident in an irradiated LMFBR core using different computer models for a common benchmark problem

    International Nuclear Information System (INIS)

    Wider, H.U.; Devos, J.; Nguyen, H.; Goethem, G. Van.; Miles, K.J.; Tentner, A.M.; Pizzica, P.

    1989-01-01

    This report summarizes the results of an international exercise to compare whole-core accident calculations of the initiation phase of an unprotected LOF accident in a large irradiated LMFBR. The results for the accident phase before pin failure are in rather good agreement except for the fuel pin mechanics predictions. There are also some differences in the sodium boiling calculations but the voiding rates which are of key importance are very similar. The post - failure fuel motion and sodium voiding predictions show significant differences. However, the majority of these calculations agree that temporary fuel accumulations occur which increase the power beyond that caused by sodium voiding alone

  9. Volatile Elements Retention During Injection Casting of Metallic Fuel Slug for a Recycling Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Jong-Hwan; Song, Hoon; Kim, Hyung-Tae; Oh, Seok-Jin; Kuk, Seoung-Woo; Keum, Chang-Woon; Lee, Jung-Won; Kim, Ki-Hwan; Lee, Chan-Bock

    2015-01-01

    The as-cast fuels prepared by injection casting were sound and the internal integrities were found to be satisfactory through gamma-ray radiography. U and Zr were uniform throughout the matrix of the slug, and the impurities, i.e., oxygen, carbon, and nitrogen, satisfied the specification of the total impurities of less than 2000 ppm. The losses of the volatile Mn were effectively controlled using argon over pressures, and dynamic pumping for a period of time before injection showed no detrimental effect on the Mn loss by vaporization. This result suggests that volatile minor actinide-bearing fuels for SFRs can be prepared by improved injection methods. 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, 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, several injection casting methods were applied in order to prepare metallic fuel for an fast reactor that control the transport of volatile elements during fuel melting and casting. Mn was selected as a surrogate alloy since it possesses a total vapor pressure equivalent to that of a volatile minor actinide-bearing fuel. U.10Zr and U.10Zr.5Mn (wt%) metallic fuels were injection cast under various casting conditions and their soundness was characterized

  10. Constituent Redistribution in U-Zr Metallic Fuel Using the Advanced Fuel Performance Code BISON

    Energy Technology Data Exchange (ETDEWEB)

    Galloway, Jack D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Unal, Cetin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Matthews, Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-30

    Previous work done by Galloway, et. al. on EBR-II ternary (U-Pu-Zr) fuel constituent redistribution yielded accurate simulation data for the limited data sets of Zr redistribution. The data sets included EPMA scans of two different irradiated rods. First, T179, which was irradiated to 1.9 at% burnup, was analyzed. Second, DP16, which was irradiated to 11 at% burnup, was analyzed. One set of parameters that most accurately represented the zirconium profiles for both experiments was determined. Since the binary fuel (U-Zr) has previously been used as the driver fuel for sodium fast reactors (SFR) as well as being the likely driver fuel if a new SFR is constructed, this same process has been initiated on the binary fuel form. From limited binary EPMA scans as well as other fuel characterization techniques, it has been observed that zirconium redistribution also occurs in the binary fuel, albeit at a reduced rate compared to observation in the ternary fuel, as noted by Kim et. al. While the rate of redistribution has been observed to be slower, numerous metallographs of U-Zr fuel show distinct zone formations.

  11. Nano-structured noble metal catalysts based on hexametallate architecture for the reforming of hydrocarbon fuels

    Science.gov (United States)

    Gardner, Todd H.

    2015-09-15

    Nano-structured noble metal catalysts based on hexametallate lattices, of a spinel block type, and which are resistant to carbon deposition and metal sulfide formation are provided. The catalysts are designed for the reforming of hydrocarbon fuels to synthesis gas. The hexametallate lattices are doped with noble metals (Au, Pt, Rh, Ru) which are atomically dispersed as isolated sites throughout the lattice and take the place of hexametallate metal ions such as Cr, Ga, In, and/or Nb. Mirror cations in the crystal lattice are selected from alkali metals, alkaline earth metals, and the lanthanide metals, so as to reduce the acidity of the catalyst crystal lattice and enhance the desorption of carbon deposit forming moieties such as aromatics. The catalysts can be used at temperatures as high as 1000.degree. C. and pressures up to 30 atmospheres. A method for producing these catalysts and applications of their use also is provided.

  12. German position paper on structural analysis for LMFBR applications

    International Nuclear Information System (INIS)

    Angerbauer, A.; Link, F.

    1983-01-01

    During the design period of the German LMFBR, the SNR-300, extensive work had been done in the field of elastic and inelastic analysis. Furthermore, special design rules have been developed. A review of these activities and their state-of-the art is outlined in this paper

  13. Technical considerations relative to removal of sodium from LMFBR components

    International Nuclear Information System (INIS)

    McDonald, J.S.; Asquith, J.G.

    1975-01-01

    Reviewed in this paper are technical considerations which are of importance in choosing between an alcohol process and a moist nitrogen process for the removal of sodium from LMFBR components. Results observed in laboratory tests and in the cleaning of large scale components (e.g. a 28 MWt Modular Steam Generator Test Unit) are presented and discussed. (author)

  14. Studies of LMFBR: method of analysis and some results

    International Nuclear Information System (INIS)

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

    1983-01-01

    Some results of recent studies of LMFBR characteristics are summarized. A two-dimensional model of the LMFBR is taken from a publication and used as the base model for the analysis. Axial structures are added to the base model and a three-dimensional (Δ - Z) calculation has been done. Two dimensional (Δ and RZ) calculations are compared with the three-dimensional and published results. The eigenvalue, flux and power distributions, breeding characteristics, control rod worth, sodium-void and Doppler reactivities are analysed. Calculations are done by CITATION using six-group cross sections collapsed regionwise by EXPANDA in one-dimensional geometries from the 70-group JFS library. Burnup calculations of a simplified thorium-cycle LMFBR have also been done in the RZ geometry. Principal results of the studies are: (1) the JFS library appears adequate for predicting overall characteristics of an LMFBR, (2) the sodium void reactivity is negative within - 25 cm from the outer boundary of the core, (3) the halflife of Pa-233 must be considered explicitly in burnup analyses, and (4) two-dimensional (RZ and Δ) calculations can be used iteratively to analyze three-dimensional reactor systems. (Author) [pt

  15. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Cheng, S.K.; Basehore, K.

    1984-08-01

    This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.

  16. Process and equipment qualification of the ceramic and metal waste forms for spent fuel treatment

    International Nuclear Information System (INIS)

    Marsden, Ken; Knight, Collin; Bateman, Kenneth; Westphal, Brian; Lind, Paul

    2005-01-01

    The electrometallurgical process for treating sodium-bonded spent metallic fuel at the Materials and Fuels Complex of the Idaho National Laboratory separates actinides and partitions fission products into two waste forms. The first is the metal waste form, which is primarily composed of stainless steel from the fuel cladding. This stainless steel is alloyed with 15w% zirconium to produce a very corrosion-resistant metal which binds noble metal fission products and residual actinides. The second is the ceramic waste form which stabilizes fission product-loaded chloride salts in a sodalite and glass composite. These two waste forms will be packaged together for disposal at the Yucca Mountain repository. Two production-scale metal waste furnaces have been constructed. The first is in a large argon-atmosphere glovebox and has been used for equipment qualification, process development, and process qualification - the demonstration of process reliability for production of the DOE-qualified metal waste form. The second furnace will be transferred into a hot cell for production of metal waste. Prototype production-scale ceramic waste equipment has been constructed or procured; some equipment has been qualified with fission product-loaded salt in the hot cell. Qualification of the remaining equipment with surrogate materials is underway. (author)

  17. Method and device for removing nuclear fuel pellets from a metal fuel can

    International Nuclear Information System (INIS)

    John, C.D. Jr.

    1978-01-01

    There are to be removed the pellets from a nuclear fuel rod without being destroyed by the process. For this purpose at one end of the fuel rod an opening is produced, through which a fluid is pressed at high pressure. Thereby the can is expanded and subsequently cut open by means of a cutting device. The pellets then may slip out. (TK) [de

  18. 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.)

  19. Hydrogen production from bio-fuels using precious metal catalysts

    Directory of Open Access Journals (Sweden)

    Pasel Joachim

    2017-01-01

    Full Text Available Fuel cell systems with integrated autothermal reforming unit require active and robust catalysts for H2 production. Thus, an experimental screening of catalysts for autothermal reforming of commercial biodiesel fuel was performed. Catalysts consisted of a monolithic cordierite substrate, an oxide support (γ-Al2O3 and Pt, Ru, Ni, PtRh and PtRu as active phase. Experiments were run by widely varying the O2/C and H2O/C molar ratios at different gas hourly space velocities. Fresh and aged catalysts were characterized by temperature programmed methods and thermogravimetry to find correlations with catalytic activity and stability.

  20. Hydrogen production from bio-fuels using precious metal catalysts

    Science.gov (United States)

    Pasel, Joachim; Wohlrab, Sebastian; Rotov, Mikhail; Löhken, Katrin; Peters, Ralf; Stolten, Detlef

    2017-11-01

    Fuel cell systems with integrated autothermal reforming unit require active and robust catalysts for H2 production. Thus, an experimental screening of catalysts for autothermal reforming of commercial biodiesel fuel was performed. Catalysts consisted of a monolithic cordierite substrate, an oxide support (γ-Al2O3) and Pt, Ru, Ni, PtRh and PtRu as active phase. Experiments were run by widely varying the O2/C and H2O/C molar ratios at different gas hourly space velocities. Fresh and aged catalysts were characterized by temperature programmed methods and thermogravimetry to find correlations with catalytic activity and stability.

  1. Economic Viability of Metallic Sodium-Cooled Fast Reactor Fuel in Korea

    Directory of Open Access Journals (Sweden)

    S. K. Kim

    2013-01-01

    Full Text Available This paper evaluates whether SFR metallic nuclear fuel can be economical. To make this determination, the cost of SFCF (SFR fuel cycle facilities was estimated, and the break-even point of the manufacturing cost of SFR metallic nuclear fuel for direct disposal option was then calculated. As a result of the cost estimation, the levelized unit cost (LUC for SFCF was calculated to be 5,311 $/kgHM, and the break-even point was calculated to be $5,267/kgHM. Therefore, the cost difference between LUC and the break-even point is not only small but is also within the relevant range of the uncertainty level of Class 3 in accordance with a generic cost estimate classification matrix of AACE (the Association for the Advancement of Cost Engineering. This means it is very difficult to judge the economical feasibility of SFR metallic nuclear fuel because as of today there are no commercial facilities in Korea or the world. The economic feasibility of SFR metallic nuclear fuel, however, will be enhanced if the mass production of SFCF becomes possible in the future.

  2. Validation of models for the analysis of the transient behavior of metallic fast reactor fuel

    International Nuclear Information System (INIS)

    Kramer, J.M.; Hughes, T.H.; Gruber, E.E.

    1989-01-01

    The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in U-Pu-Zr metal alloys as a fuel for sodium-cooled fast reactors. Part of the attractiveness of the IFR concept is the improvement in reactor safety margins through inherent features of a metal-fueled LMR core. In order to demonstrate these safety margins it is necessary to have computer codes available to analyze the detailed response of metallic fuel to a wide range of accident initiators. Two of the codes that play a key role in assessing this response are the STARS fission gas behavior code and the FPIN2 fuel pin mechanics code. Verification and validation are two important components in the development of models and computer codes. Verification demonstrates through comparison of calculations with analytical solutions that the methodology and algorithms correctly solve the equations that govern the phenomena being modeled. Validation, on the other hand, demonstrates through comparison with data that the phenomena are being modeled correctly. Both components are necessary in order to have the confidence to extrapolate the calculations to reactor accident conditions. This paper presents the results of recent progress in the validation of models for the analysis of the behavior of metallic fast reactor fuel. 9 refs., 7 figs

  3. Metal hydride and pyrophoric fuel additives for dicyclopentadiene based hybrid propellants

    Science.gov (United States)

    Shark, Steven C.

    The purpose of this study is to investigate the use of reactive energetic fuel additives that have the potential to increase the combustion performance of hybrid rocket propellants in terms of solid fuel regression rate and combustion efficiency. Additives that can augment the combustion flame zone in a hybrid rocket motor by means of increased energy feedback to the fuel grain surface are of great interest. Metal hydrides have large volumetric hydrogen densities, which gives these materials high performance potential as fuel additives in terms of specifc impulse. The excess hydrogen and corresponding base metal may also cause an increase in the hybrid rocket solid fuel regression rate. Pyrophoric additives also have potential to increase the solid fuel regression rate by reacting more readily near the burning fuel surface providing rapid energy feedback. An experimental performance evaluation of metal hydride fuel additives for hybrid rocket motor propulsion systems is examined in this study. Hypergolic ignition droplet tests and an accelerated aging study revealed the protection capabilities of Dicyclopentadiene (DCPD) as a fuel binder, and the ability for unaided ignition. Static hybrid rocket motor experiments were conducted using DCPD as the fuel. Sodium borohydride (NabH4) and aluminum hydride (AlH3) were examined as fuel additives. Ninety percent rocket grade hydrogen peroxide (RGHP) was used as the oxidizer. In this study, the sensitivity of solid fuel regression rate and characteristic velocity (C*) efficiency to total fuel grain port mass flux and particle loading is examined. These results were compared to HTPB combustion performance as a baseline. Chamber pressure histories revealed steady motor operation in most tests, with reduced ignition delays when using NabH4 as a fuel additive. The addition of NabH4 and AlH3 produced up to a 47% and 85% increase in regression rate over neat DCPD, respectively. For all test conditions examined C* efficiency ranges

  4. Theory and use of GIRAFFE for analysis of decay characteristics of delayed-neutron precursors in an LMFBR

    International Nuclear Information System (INIS)

    Gross, K.C.

    1980-07-01

    The application of the computer code GIRAFFE (General Isotope Release Analysis For Failed Elements) written in FORTRAN IV is described. GIRAFFE was designed to provide parameter estimates of the nonlinear discrete-measurement models that govern the transport and decay of delayed-neutron precursors in a liquid-metal fast breeder reactor (LMFBR). The code has been organized into a set of small, relatively independent and well-defined modules to facilitate modification and maintenance. The program logic, the numerical techniques, and the methods of solution used by the code are presented, and the functions of the MAIN program and of each subroutine are discussed

  5. Metal waste forms from treatment of EBR-II spent fuel

    International Nuclear Information System (INIS)

    Abraham, D. P.

    1998-01-01

    Demonstration of Argonne National Laboratory's electrometallurgical treatment of spent nuclear fuel is currently being conducted on irradiated, metallic driver fuel and blanket fuel elements from the Experimental Breeder Reactor-II (EBR-II) in Idaho. The residual metallic material from the electrometallurgical treatment process is consolidated into an ingot, the metal waste form (MWF), by employing an induction furnace in a hot cell. Scanning electron microscopy (SEM) and chemical analyses have been performed on irradiated cladding hulls from the driver fuel, and on samples from the alloy ingots. This paper presents the microstructures of the radioactive ingots and compares them with observations on simulated waste forms prepared using non-irradiated material. These simulated waste forms have the baseline composition of stainless steel - 15 wt % zirconium (SS-15Zr). Additions of noble metal elements, which serve as surrogates for fission products, and actinides are made to that baseline composition. The partitioning of noble metal and actinide elements into alloy phases and the role of zirconium for incorporating these elements is discussed in this paper

  6. Understanding metal-organic frameworks for solar fuel production

    NARCIS (Netherlands)

    Garcia Santaclara, J.

    2017-01-01

    The fascinating chemical and physical properties of MOFs have recently stimulated exploration of their application for photocatalysis. Despite the intense research effort, the efficiencies of most photocatalytic MOFs for solar fuel generation are still very modest. In this introduction we analyse

  7. Metal plutonium conversion to components of nuclear reactor fuel

    International Nuclear Information System (INIS)

    Subbotin, V.G.; Panov, A.V.; Mashirev, V.P.

    2000-01-01

    Capabilities of different technologies for plutonium conversion to the fuel components of nuclear reactors are studied. Advantages and shortcomings of aqueous and nonaqueous methods of plutonium treatment are shown. Proposals to combine and coordinate efforts of world scientific and technological community in solving problems concerning plutonium of energetic and weapon origin treatment were put forward. (authors)

  8. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Jeong, C. J.; Yang, M. S

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs.

  9. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    International Nuclear Information System (INIS)

    Park, Jee Won; Jeong, C. J.; Yang, M. S.

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs

  10. A fuel freezing model for liquid-metal fast breeder reactor hypothetical core disruptive accidents

    Energy Technology Data Exchange (ETDEWEB)

    Best, F.R.; Erdman, C.; Wayne, D.

    1985-01-01

    A proposed fuel freezing mechanism for molten UO2 fuel penetrating a steel channel was investigated in the course of liquid-metal-cooled fast breeder reactor hypothetical core disruptiv accident safety studies. The fuel crust deposited on an underlying melting steel wall was analyzed as being subjected to two stresses one due to the pressure difference between the flowing fuel and the stagnant molten steel layer, and the other resulting from the temperature variation through the crust thickness. Analyses based on the proposed freezing mechanism and comparisons with fuel freezing experiments confirmed that fuel freezing occurs in three modes. For initially low steel wall temperatures, the fuel crust was stable and grew to occlude the channel. At high steel wall temperatures (above 1070 K), instantaneous wall melting leading to steel entrainment was calculated to occur with final penetration depending on the refreezing of the entrained steel. Between these two extremes, the stress developed within the crust at the steel melting front exceeds the critical buckling value, the crust ruptures, and steel is injected into the fuel flow. Freezing is dominated by the fuel/steel mixture. The theoretical penetration distances and freezing times were in good agreement with the experimental results with no more than 20% error involved.

  11. Material properties requirements for LMFBR structural design: general considerations and data needs

    International Nuclear Information System (INIS)

    Pugh, C.E.; Purdy, C.M.

    1977-01-01

    A statement is given of material properties information needed in connection with the structural design technology for liquid-metal fast breeder reactor (LMFBR) primary circuit components. Implementation of current analysis methods and criteria is considered with an emphasis on data and data correlations for performing elastic-plastic and creep analyses, for establishing allowable stress limits, and for computing creep-fatigue damage. Further development of the technology is discussed in relation to properties information. Emphasis is placed on improved constitutive equations for representing inelastic material behavior, on procedures for treating time-dependent fatigue, and on criteria for creep rupture. The properties are generally discussed without regard to specific alloys, since most categories of information are needed for each major structural material. Some sample experimental results are given for type 304 stainless steel and 2 1 / 4 Cr-1 Mo steel

  12. Sodium-NaK engineering handbook. Volume III. Sodium systems, safety, handling, and instrumentation. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Foust, O J [ed.

    1978-01-01

    The handbook is intended for use by present and future designers in the Liquid Metals Fast Breeder Reactor (LMFBR) Program and by the engineering and scientific community performing other type investigation and exprimentation requiring high-temperature sodium and NaK technology. The arrangement of subject matter progresses from a technological discussion of sodium and sodium--potassium alloy (NaK) to discussions of varius categories and uses of hardware in sodium and NaK systems. Emphasis is placed on sodium and NaK as heat-transport media. Sufficient detail is included for basic understanding of sodium and NaK technology and of technical aspects of sodium and NaK components and instrument systems. Information presented is considered adequate for use in feasibility studies and conceptual design, sizing components and systems, developing preliminary component and system descriptions, identifying technological limitations and problem areas, and defining basic constraints and parameters.

  13. Fast reactors with axial arrangement of oxide and metal fuels in the core

    International Nuclear Information System (INIS)

    Troyanov, M.F.; Ilyunin, V.G.; Matveev, V.I.; Murogov, V.M.; Proshkin, A.A.; Rudneva, V.Ya.; Shmelev, A.N.

    1980-01-01

    Problems of using metal fuel in fast reactor (FR) core are discussed Results are given of the calculation of two-dimentional (R-Z) FR version having a composed core with the combined usage of oxide and metal fuels having parameters close to optimal from the point of view of fuel breeding rate, an oxide subzone having increased enrichment and a decreased proper conversion ratio. A reactor is considered where metallic fuel elements are placed from the side of ''cold'' coolant inlet (400-480 deg C), and oxide fuel elements - in the region where the coolant has a higher temperature (500-560 deg C). It is shown that the new fuel breeding rate in such a reactor can be increased by 20-30% as compared with an oxide fuel reactor. Growth of the total conversion ratio is mainly stipulated with the increase of the inner conversion ratio of the core (CRC) which is important not only from the point of view of nuclear fuel breeding rate but also the optimization of the mode of powerful fast reactor operation with provision for the change in reactivity in the process of its continuous operation. The fact, that the core version under investigation has a CRC value slightly exceeding unit, stipulates considerably less reactivity change as compared with the oxide version in the process of the reactor operation and permits at a constant reactor control system power to significantly increase the time between reloadings and, therefore, to increase the NPP load factor which is of great importance both from the point of view of economy and the improvement of operation conditions as well as of reactor operation reliability. It is concluded on the base of the analysis of the results obtained that FRs with the combined usage of oxide and metal fuels having an increased specific load and increased conversion ratio as compared with the oxide fuel FRs provide a higher rate of development of the whole nuclear power balanced with respect to the fuel [ru

  14. LMFBR steam generators in the United Kingdom

    International Nuclear Information System (INIS)

    Anderson, R.; Hayden, O.

    2002-01-01

    Experience has been gained in the UK on the operation of LMFBR Steam Generator Units (SGU) over a period of 20 years from the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR). The DFR steam generator featured a double barrier and therefore did not represent a commercial design. PFR, however, faced the challenge of a single wall design and it is experience from this which is most valuable. The PFR reactor went critical in March 1974 and the plant operating history since then has been dominated by experience with leaks in the tube to tube plate welds of the high performance U-tubes SGU's. Operation at high power using the full complement of three secondary sodium circuits was delayed until July 1976 by the occurrence of leaks in the tube to tube plate welds of the superheater and reheater units which are fabricated in stainless steel. Repairs were carried out to the two superheaters and they were returned to service. The reheater tube bundle was removed from circuit after sodium was found to have entered the steam side. When the sodium had been removed and inspection carried out it was decided not to recover the unit. Since 1976 the remaining five stainless steel units have operated satisfactorily. This year a replacement reheater unit has been installed. This is of a new design in 9-Cr-Mo ferritic steel using a sleeve through which the steam tube passes to eliminate the tube to tube plate weld. Despite a few early leaks in evaporator tube to tube plate welds up to 1979, these failures did not initially present a major problem. However, in 1980 the rate of evaporator weld failures increased and despite the successful application of a shot peening process to eliminate stress corrosion failures from the water side of the weld, failures traced to the sodium side continued. A sleeving process was developed for application to complete evaporator units on a production basis with the objective of bypassing the welds at each end of the 500 tubes. The decision

  15. French studies on local blockages in LMFBR fuel subassemblies

    International Nuclear Information System (INIS)

    Girard, C.; Jolas, P.; Seiler, J.M.

    1979-08-01

    This paper reviews experimental and theoretical studies done in FRANCE on the problem of partial subassembly blockages. The priorities are defined and the development of the French program in the European context is presented. Results of the out of pile experiments performed at CEA and EDF in single and two phases flow are given. A description of the main codes used to interpret these experiments is then shortly reviewed. It is found that the thermal behavior in single phase may be calculated with good precision, and that a simple semi-empirical formula can predict with good accuracy the number of channels blocked that lead to sodium boiling

  16. Nuclear fuel cycle head-end enriched uranium purification and conversion into metal

    International Nuclear Information System (INIS)

    Bonini, A.; Cabrejas, J.; Lio, L. de; Dell'Occhio, L.; Devida, C.; Dupetit, G.; Falcon, M.; Gauna, A.; Gil, D.; Guzman, G.; Neuringer, P.; Pascale, A.; Stankevicius, A.

    1998-01-01

    The CNEA (Comision Nacional de Energia Atomica - Argentina) operated two facilities at the Ezeiza Atomic Center which supply purified enriched uranium employed in the production of nuclear fuels. At one of those facilities, the Triple Height Laboratory scraps from the production of MTR type fuel elements (mainly out of specification U 3 O 8 plates or powder) are purified to nuclear grade. The purification is accomplished by a solvent extraction process. The other facility, the Enriched Uranium Laboratory produces 90% enriched uranium metal to be used in Mo 99 production (originally the uranium was used for the manufacture of MTR fuel elements made of aluminium-uranium alloy). This laboratory also provided metallic uranium with a lower enrichment (20%) for a first uranium-silicon testing fuel element, and in the near future it is going to recommence 20% enriched uranium related activities in order to provide the metal for the silicon-based fuel elements production (according to the policy of enrichment reduction for MTR reactors). (author)

  17. Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Hartmann C.

    2018-01-01

    Full Text Available Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc. during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE. Two sets of floating linear voltage differential transformer (LVDT pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has

  18. Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

    Science.gov (United States)

    Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

    2018-01-01

    Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been

  19. Review of the literature for dry reprocessing oxide, metal, and carbide fuel: The AIROX, RAHYD, and CARBOX pyrochemical processes

    Energy Technology Data Exchange (ETDEWEB)

    Hoyt, R.C.; Rhee, B.W. [Rockwell International Corp., Canoga Park, CA (United States). Energy Systems Group

    1979-09-30

    The state of the art of dry processing oxide, carbide, and metal fuel has been determined through an extensive literature review. Dry processing in one of the most proliferation resistant fuel reprocessing technologies available to date, and is one of the few which can be exported to other countries. Feasibility has been established for oxide, carbide, and metal fuel on a laboratory scale, and large-scale experiments on oxide and carbide fuel have shown viability of the dry processing concept. A complete dry processing cycle has been demonstrated by multicycle processing-refabrication-reirradiation experiments on oxide fuel. Additional experimental work is necessary to: (1) demonstrate the complete fuel cycle for carbide and metal fuel, (2) optimize dry processing conditions, and (3) establish fission product behavior. Dry process waste management is easier than for an aqueous processing facility since wastes are primarily solids and gases. Waste treatment can be accomplished by techniques which have been, or are being, developed for aqueous plants.

  20. Gas phase deposition of oxide and metal-oxide coatings on fuel particles

    International Nuclear Information System (INIS)

    Patokin, A.P.; Khrebtov, V.L.; Shirokov, B.M.

    2008-01-01

    Production processes and properties of oxide (Al 2 O 3 , ZrO 2 ) and metal-oxide (Mo-Al 2 O 3 , Mo-ZrO 2 , W-Al 2 O 3 , W-ZrO 2 ) coatings on molybdenum substrates and uranium dioxide fuel particles were investigated. It is shown that the main factors that have an effect on the deposition rate, density, microstructure and other properties of coatings are the deposition temperature, the ratio of H 2 and CO 2 flow rates, the total reactor pressure and the ratio of partial pressures of corresponding metal chlorides during formation of metal-oxide coatings

  1. Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

    Science.gov (United States)

    Brehm, Jr., William F.; Colburn, Richard P.

    1982-01-01

    An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.

  2. Investigation of Novel Electrocatalysts for Metal Supported Solid Oxide Fuel Cells - Ru:GDC

    DEFF Research Database (Denmark)

    Sudireddy, Bhaskar Reddy; Nielsen, Jimmy; Thydén, Karl Tor Sune

    2015-01-01

    The electrochemical performance and stability of the planar metal supported solid oxide fuel cells (MS-SOFC) with two different electrocatalytically active materials, namely, Ni:GDC and Ru:GDC were investigated. Ru:GDC with an ASR of 0.322 Ωcm2 performed better than Ni:GDC with an ASR of 0.453 Ωc...

  3. Corrosion resistance of metallic materials for use in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Legry, J.P.; Pelras, M.; Turluer, G.

    1989-01-01

    This paper reviews the corrosion resistance properties required from metallic materials to be used in the various developments of the PUREX process for nuclear fuel reprocessing. Stainless steels, zirconium or titanium base alloys are considered for the various plant components, where nitric acid is the main electrolyte with differing acid and nitrate concentrations, temperature and oxidizing species. (author)

  4. Determination of uranium metal concentration in irradiated fuel storage basin sludge using selective dissolution

    International Nuclear Information System (INIS)

    Delegard, C.H.; Sinkov, S.I.; Chenault, J.W.; Schmidt, A.J.; Pool, K.N.; Welsh, T.L.

    2014-01-01

    Irradiated uranium metal fuel was stored underwater in the K East and K West storage basins at the US Department of Energy Hanford Site. The uranium metal under damaged cladding reacted with water to generate hydrogen gas, uranium oxides, and spalled uranium metal particles which intermingled with other particulates to form sludge. While the fuel has been removed, uranium metal in the sludge remains hazardous. An expeditious routine method to analyze 0.03 wt% uranium metal in the presence of >30 wt% total uranium was needed to support safe sludge management and processing. A selective dissolution method was designed based on the rapid uranium oxide dissolution but very low uranium metal corrosion rates in hot concentrated phosphoric acid. The uranium metal-bearing heel from the phosphoric acid step then is rinsed before the uranium metal is dissolved in hot concentrated nitric acid for analysis. Technical underpinnings of the selective dissolution method, including the influence of sludge components, were investigated to design the steps and define the reagents, quantities, concentrations, temperatures, and times within the selective dissolution analysis. Tests with simulant sludge proved the technique feasible. Tests with genuine sludge showed a 0.0028 ± 0.0037 wt% (at one standard deviation) uranium metal analytical background, a 0.011 wt% detection limit, and a 0.030 wt% quantitation limit in settled (wet) sludge. In tests using genuine K Basin sludge spiked with uranium metal at concentrations above the 0.030 wt% ± 25 % (relative) quantitation limit, uranium metal recoveries averaged 99.5 % with a relative standard deviation of 3.5 %. (author)

  5. Investigation of Reactivity Feedback Mechanism of Axial and Radial Expansion Effect of Metal-Fueled Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Seong, Seung-Hwan; Choi, Chi-Woong; Jeong, Tae-Kyung; Ha, Gi-Seok

    2015-01-01

    The major inherent reactivity feedback models for a ceramic fuel used in a conventional light water reactor are Doppler feedback and moderator feedback. The metal fuel has these two reactivity feedback mechanisms previously mentioned. In addition, the metal fuel has two more reactivity feedback models related to the thermal expansion phenomena of the metal fuel. Since the metal fuel has a good capability to expand according to the temperature changes of the core, two more feedback mechanisms exist. These additional two feedback mechanism are important to the inherent safety of metal fuel and can make metal-fueled SFR safer than oxide-fueled SFR. These phenomena have already been applied to safety analysis on design extended condition. In this study, the effect of these characteristics on power control capability was examined through a simple load change operation. The axial expansion mechanism is induced from the change of the fuel temperature according to the change of the power level of PGSFR. When the power increases, the fuel temperatures in the metal fuel will increase and then the reactivity will decrease due to the axial elongation of the metal fuel. To evaluate the expansion effect, 2 cases were simulated with the same scenario by using MMS-LMR code developed at KAERI. The first simulation was to analyze the change of the reactor power according to the change of BOP power without the reactivity feedback model of the axial and radial expansion of the core during the power transient event. That is to say, the core had only two reactivity feedback mechanism of Doppler and coolant temperature

  6. Mixing requirements for the limiting fuel-coolant interactions in liquid metal fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lenz, Jr, W F

    1976-11-01

    An estimation of the mixing requirements for the limiting fuel-coolant interactions in two specific liquid metal cooled fast reactors, the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor (CRBR), has been undertaken. The mixing requirements were represented in terms of the limiting mixing time constants. These constants were determined with the Argonne parametric FCI Computer Model for a range of core involvements. Specifically, fuel masses used ranged from as low as one-seventh of the core to a full core involvement. In general, conservative values for additional FCI input parameters were assumed such that the results would be conservative. With the results in hand, several mechanisms were investigated to determine what limiting effects they could have on the mixing rates of the fuel and coolant during an FCI. The energy requirements for mixing were investigated. The results, however, provided no limiting effects. A solidification limited fragmentation model was also investigated. Although this model provided no absolute limiting effects, it did show that fuel particle sizes of a certain size could indeed limit the fuel-coolant mixing rates. Additionally, the limiting effects were found to be much less significant for UC fuel. The third mechanism that was investigated concerned the limiting effects of the finite fuel release rates as a result of TOP accidents in the FFTF. Equivalent mixing time constants based on the fuel release rates were shown to be greater than the limiting values. Thus, this mechanism was shown to be limiting for the particular accident sequence investigated.

  7. Influence of oxygen-metal ratio on mixed-oxide fuel performance

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Leggett, R.D.

    1979-04-01

    The fuel oxygen-to-metal ratio (O/M) is recognized as an important consideration for performance of uranium--plutonium oxide fuels. An overview of the effects of differing O/M's on the irradiation performance of reference design mixed-oxide fuel in the areas of chemical and mechanical behavior, thermal performance, and fission gas behavior is presented. The pellet fuel has a nominal composition of 75 wt% UO 2 + 25 wt% PuO 2 at a pellet density of approx. 90% TD. for nominal conditions this results in a smeared density of approx. 85%. The cladding in all cases is 20% CW type 316 stainless steel with an outer diameter of 5.84 to 6.35 mm. O/M has been found to significantly influence fuel pin chemistry, mainly FCCI and fission product and fuel migration. It has little effect on thermal performance and overall mechanical behavior or fission gas release. The effects of O/M (ranging from 1.938 to 1.984) in the areas of fuel pin chemistry, to date, have not resulted in any reduction in fuel pin performance capability to goal burnups of approx. 8 atom% or more

  8. Gas Generation from K East Basin Sludges and Irradiated Metallic Uranium Fuel Particles Series III Testing

    International Nuclear Information System (INIS)

    Schmidt, Andrew J.; Delegard, Calvin H.; Bryan, Samuel A.; Elmore, Monte R.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

    2003-01-01

    The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer

  9. 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

  10. Transition phase in LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Ostensen, R.W.; Henninger, R.J.; Jackson, J.F.

    1976-01-01

    Mechanistic analyses of transient-under-cooling accidents have led in some cases to a mild initiating phase instead of a direct hydrodynamic disassembly of the core. The fuel is then trapped in the core by the strong mechanical surroundings and blockages formed by refrozen cladding steel and/or fuel. The formation of fuel blockages has been verified experimentally. The bottled-up core will boil on fission and decay heat, with steel as the working fluid. Boil-up in a churn turbulent flow regime may prevent recriticality due to fuel recompaction. Ultimate fuel removal from the core is probably by a two-phase blow-down after permanent leakage paths are opened. However, a vigorous recriticality can not be precluded. Reactors with void coefficients larger than that in CRBR are more likely to disassemble in the initiating phase, so the transition phase may be unique to small cores

  11. New Nuclear Materials Including Non Metallic Fuel Elements. Vol. II. Proceedings of the Conference on New Nuclear Materials Technology, Including Non Metallic Fuel Elements

    International Nuclear Information System (INIS)

    1963-01-01

    One of the major aims of the International Atomic Energy Agency in furthering the peaceful uses of atomic energy is to encourage the development of economical nuclear power. Certainly, one of the more obvious methods of producing economical nuclear power is the development of economical fuels that can be used at high temperatures for long periods of time, and which have sufficient strength and integrity to operate under these conditions without permitting the release of fission products. In addition it is desirable that after irradiation these new fuels be economically reprocessed to reduce further the cost of the fuel cycle. As nuclear power becomes more and more competitive with conventional power the interest in new and more efficient higher-temperature fuels naturally increases rapidly. For these reasons, the Agency organized a Conference on New Nuclear Materials Technology, Including Non-Metallic Fuel Elements, which was held from 1 to 5 July 1963 at the International Hotel, Prague, with the assistance and co-operation of the Government of the Czechoslovak Socialist Republic. A total of 151 scientists attended, from 23 countries and 4 international organizations. The participants heard and discussed more than 60 scientific papers. The Agency wishes to thank the scientists who attended this Conference for their papers and for many spirited discussions that truly mark a successful meeting. The Agency wishes also to record its gratitude for the assistance and generous hospitality accorded the Conference, the participants and the Agency's staff by the Government of the Czechoslovak Socialist Republic and by the people of Prague. The scientific information contained in these Proceedings should help to quicken the pace of progress in the fabrication of new and m ore economical fuels, and it is hoped that these proceedings will be found useful to all workers in this and related fields

  12. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

    Forbes, I.A.; Driscoll, M.J.; Rasmussen, N.C.; Lanning, D.D.; Kaplan, I.

    1971-01-01

    This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238 U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics

  13. Hydrogen jet recombination under postulated LMFBR accident conditions

    International Nuclear Information System (INIS)

    Wierman, R.W.

    1977-01-01

    Certain conditions may be postulated in LMFBR risk assessments for which the potential of hydrogen release to the reactor containment building needs to be evaluated. The inherent self-ignition characteristics of hydrogen jets entering the air atmosphere of the reactor containment building should be understood for such analyses. If hydrogen jets were to self-ignite (recombine) at the source where they enter the reactor containment building, then undesirable hydrogen accumulation would not occur. Therefore, experiments have been conducted investigating the phenomena associated with the recombination of hydrogen jets under conditions similar to those postulated for LMFBR studies. The data presented define the conditions required for self-ignition of the hydrogen jets

  14. Metal Nanoshells for Plasmonically Enhanced Solar to Fuel Photocatalytic Conversion

    Science.gov (United States)

    2016-05-18

    facilitate charge separation of photo-excited carriers. Zinc indium sulfide (ZIS), a visible light- active photocatalyst, was used as the photocatalytic matrix...then coated with a thin layer of silica (SiO2), followed by a zinc indium sulfide (ZnIn2S4; ZIS) semiconductor shell. The blended-metal GS-NS cores...of pH values (ACS Appl. Mater. Interfaces 2013, 5, 2479). Such an improvement in stability not only offers alternative routes to the synthesis of

  15. Development of acidic processes for decontaminating LMFBR components

    International Nuclear Information System (INIS)

    Hill, E.F.; Colburn, R.P.; Lutton, J.M.; Maffei, H.P.

    1978-01-01

    The objective of the DOE decontamination program is to develop a well characterized chemical decontamination process for application to LMFBR primary system components that subsequently permits contact maintenance and allows requalification of the components for reuse in reactors. The paper describes the subtasks of deposit characterization, development of requalification and process acceptance criteria, development of process evaluation techniques and studies which led to a new acidic process for decontaminating 304 stainless steel hot leg components

  16. LMFBR operational safety: the EBR-II experience

    International Nuclear Information System (INIS)

    Sackett, J.I.; Allen, N.L.; Dean, E.M.; Fryer, R.M.; Larson, H.A.; Lehto, W.K.

    1978-01-01

    The mission of the Experimental Breeder Reactor II (EBR-II) has evolved from that of a small LMFBR demonstration plant to a major irradiation-test facility. Because of that evolution, many operational-safety issues have been encountered. The paper describes the EBR-II operational-safety experience in four areas: protection-system design, safety-document preparation, tests of off-normal reactor conditions, and tests of elements with breached cladding

  17. Thermal analysis methods for LMFBR wire wrapped bundles

    International Nuclear Information System (INIS)

    Todreas, N.E.

    1976-11-01

    A note is presented which was written to stimulate an awareness and discussion of the fundamental differences in the formulation of certain existing analysis codes for LMFBR wire wrap bundles. The contention of the note is that for those array types where data exists (one wire per pin, equal start angles), the ENERGY method results for coolant temperature under forced convection conditions provide benchmarks of reliability equal to the results of codes COBRA and TH1-3D

  18. Large pool LMFBR design. Final report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Wett, J. F.; Churchill, J. R.

    1979-03-01

    The design effort reported is an extension on past design effort and continuous concentration on those parts of the nuclear island unique to a commercial size pool type LMFBR. In particular, the work covers the reactor vessel, deck, rotating plugs, upper and lower internals, internal plenum separator system, IHX, pumps, cold traps, intermediate system layout, containment/confinement system, plot plan, and residual heat removal systems. Preliminary thermal, hydraulic, stress, and system analyses are also presented.

  19. Shielding design method for LMFBR validation on the Phenix factor

    International Nuclear Information System (INIS)

    Cabrillat, J.C.; Crouzet, J.; Misrakis, J.; Salvatores, M.; Rado, V.; Palmiotti, G.

    1983-05-01

    Shielding design methods, developed at CEA for shielding calculations find a global validation by the means of Phenix power reactor (250 MWe) measurements. Particularly, the secondary sodium activation of pool type LMFBR such as Super Phenix (1200 MWe) which is subject to strict safety limitation is well calculated by the adapted scheme, i.e. a two dimension transport calculation of shielding coupled to a Monte-Carlo calculation of secondary sodium activation

  20. A new approach to the design of LMFBR liners

    International Nuclear Information System (INIS)

    Polentz, L.M.

    1980-01-01

    An advance in the state-of-the-art of LMFBR liners which permits notable savings in construction costs without any sacrifice of safety is described. The application of the new design concept to the rework of the upper reactor vault liner of the FFTF is discussed. Factors which affect the application of the new design approach to other LMFBRs are delineated and discussed. (author)

  1. Structural analysis for elevated temperature design of the LMFBR

    International Nuclear Information System (INIS)

    Griffin, D.S.

    1976-02-01

    In the structural design of LMFBR components for elevated temperature service it is necessary to take account of the time-dependent, creep behavior of materials. The accommodation of creep to assure design reliability has required (1) development of new design limits and criteria, (2) development of more detailed representations of material behavior, and (3) application of the most advanced analysis techniques. These developments are summarized and examples are given to illustrate the current state of technology in elevated temperature design

  2. Future development LMFBR-steam generators SNR2

    International Nuclear Information System (INIS)

    Essebaggers, J.; Pors, J.G.

    1975-01-01

    The development work for steam generators for large LMFBR plants by Neratoom will be reviewed consisting of: 1. Development engineering information. 2. Concept select studies followed by conceptual designs of selected models. 3. Development manufacturing techniques. 4. Detail design of a prototype unit. 5. Testing of sub-constructions for prototype steam generators. In this presentation item 1 and 2 above will be high lighted, identifying the development work for the SNR-2 steam generators on short term basis. (author)

  3. Transport-diffusion comparisons for small core LMFBR disruptive accidents

    International Nuclear Information System (INIS)

    Tomlinson, E.T.

    1977-11-01

    A number of numerical experiments were performed to assess the validity of diffusion theory for calculating the reactivity state of various small core LMFBR disrupted geometries. The disrupted configurations correspond, in general, to various configurations predicted by SAS3A for transient undercooling (TUC) and transient overpower (TOP) accidents for homogeneous cores and to the ZPPR-7 configurations for heterogeneous core. In all TUC cases diffusion theory was shown to be inadequate for the calculation of reactivity changes during core disassembly

  4. Scoping studies of vapor behavior during a severe accident in a metal-fueled reactor

    International Nuclear Information System (INIS)

    Spencer, B.W.; Marchaterre, J.F.

    1985-01-01

    Scoping calculations have been performed examining the consequences of fuel melting and pin failures for a reactivity-insertion type accident in a sodium-cooled, pool-type reactor fueled with a metal alloy fuel. The principal gas and vapor species released are shown to be Xe, Cs,and bond sodium contained within the fuel porosity. Fuel vapor pressure is insignificant, and there is no energetic fuel-coolant interaction for the conditions considered. Condensation of sodium vapor as it expands into the upper sodium pool in a jet mixing regime may occur as rapidly as the vapor emerges from the disrupted core (although reactor-material experiments are needed to confirm these high condensation rates). If the predictions of rapid direct-contact condensation can be verified experimentally for the sodium system, the implication is that the ability of vapor expansion to perform appreciable work on the system is largely eliminated. Furthermore, the ability of an expanding vapor bubble to transport fuel and fission product species to the cover gas region where they may be released to the containment is also largely eliminated. The radionuclide species except for fission gas are largely retained within the core and sodium pool

  5. Compendium of computer codes for the safety analysis of LMFBR's

    International Nuclear Information System (INIS)

    1975-06-01

    A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)

  6. 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

  7. The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

    International Nuclear Information System (INIS)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung

    2016-01-01

    Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask

  8. Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors

    International Nuclear Information System (INIS)

    Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.

    1978-01-01

    Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors

  9. Spent fuel shipping cask accident evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Fields, S.R.

    1975-12-01

    Mathematical models have been developed to simulate the dynamic behavior, following a hypothetical accident and fire, of typical casks designed for the rail shipment of spent fuel from nuclear reactors, and to determine the extent of radioactive releases under postulated conditions. The casks modeled were the IF-300, designed by the General Electric Company for the shipment of spent LWR fuel, and a cask designed by the Aerojet Manufacturing Company for the shipment of spent LMFBR fuel.

  10. Microstructure characterizaton of advanced oxide fuel

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Gerber, E.W.; McCord, R.B.

    1977-01-01

    Preirradiation porosity, grain size, and microcomposition characteristics are presented for selected advanced oxide (PuO 2 -UO 2 ) LMFBR developmental fuels fabricated for irradiation testing in EBR-II. Quantitative microscopy, electron microprobe analysis, and a recently developed quantitative autoradiographic technique are utilized to relate microstructure characteristics to fabrication parameters

  11. High performance metal-supported solid oxide fuel cells with Gd-doped ceria barrier layers

    DEFF Research Database (Denmark)

    Klemensø, Trine; Nielsen, Jimmi; Blennow Tullmar, Peter

    2011-01-01

    Metal-supported solid oxide fuel cells are believed to have commercial advantages compared to conventional anode (Ni–YSZ) supported cells, with the metal-supported cells having lower material costs, increased tolerance to mechanical and thermal stresses, and lower operational temperatures...... at 650 °C and 0.6 V, were obtained on cells with barrier layers fabricated by magnetron sputtering. The performance is dependent on the density of the barrier layer, indicating Sr2+ diffusion is occurring at the intermediate SOFC temperatures. The optimized design further demonstrate improved durability...

  12. Manufacturing and characterization of metal-supported solid oxide fuel cells

    DEFF Research Database (Denmark)

    Blennow Tullmar, Peter; Hjelm, Johan; Klemensø, Trine

    2011-01-01

    A metal-supported solid oxide fuel cell design offers competitive advantages, for example reduced material costs and improved robustness. This paper reports the performance and stability of a recently developed metal-supported cell design, based on a novel cermet anode, on a 25cm2 (1cm2/16cm2...... functional layer, oxygen reduction in the mixed ionic-electronic conducting cathode and an additional polarization process with a rather high relaxation frequency, which may be assigned to an insulating corrosion interlayer.The durability of the cells was investigated by means of galvanostatic operation...

  13. Identification of passive shutdown system parameters in a metal fueled LMR

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1992-01-01

    This document discusses periodic testing of the passive shutdown system in a metal fueled liquid metal reactor which has been proposed as a Technical Specification requirement. In the approach to testing considered in this paper, perturbation experiments performed at normal operation are used to predict an envelope that bounds reactor response to flowrate, inlet temperature and external reactivity forcing functions. When the envelope for specific upsets lies within safety limits, one concludes that the passive shutdown system is operation properly for those upsets. Simulation results for the EBR-II reactor show that the response envelope for loss of flow and rod reactivity insertion events does indeed bound these events

  14. Metal waste forms from the electrometallurgical treatment of spent nuclear fuel

    International Nuclear Information System (INIS)

    Abraham, D.P.; McDeavitt, S.M.; Park, J.

    1996-01-01

    Stainless steel-zirconium alloys are being developed for the disposal of radioactive metal isotopes isolated using an electrometallurgical treatment technique to treat spent nuclear fuel. The nominal waste forms are stainless steel-15 wt% zirconium alloy and zirconium-8 wt% stainless steel alloy. These alloys are generated in yttria crucibles by melting the starting materials at 1,600 C under an argon atmosphere. This paper discusses the microstructures, corrosion and mechanical test results, and thermophysical properties of the metal waste form alloys

  15. 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

  16. Plasmon-enhanced Solar Fuel Production with Gold-metal Oxide Hybrid Nanomaterials

    DEFF Research Database (Denmark)

    Engelbrekt, Christian; Law, Matt; Zhang, Jingdong

    Harnessing sunlight to drive chemical reactions for energy storage is an important element in the transitiontowards green and sustainable technologies. Solar fuel production using semiconductor nanoparticles (SNPs) are widely studied but suffer from poor utilization of the solar spectrum and....../or poor stability under operating conditions. A new avenue addressing these challenges involve plasmonic metal nanoparticles (PNPs). These stable materials have tunable optical properties and exciting catalytic behavior. Composite photocatalysts of SNPs and PNPs exploit the majority of the solar spectrum......, provide new catalytic routes and expands the scope of solar photocatalysis. We prepare metal oxide SNPs, gold PNPs and their hybrids through mild aqueous syntheses to develop efficient photocatalyst for solar fuel production. Focus is placed on the synergetic interplay between SNPs and PNPs, understanding...

  17. Modeling the behavior of metallic fast reactor fuels during extended transients

    International Nuclear Information System (INIS)

    Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

    1993-01-01

    Passive safety features in metal-fueled reactors utilizing the Integral Fast Reactor (IFR) fuel system make it possible to avoid core damage for extended time periods even when automatic scram system fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this intermediate time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements. (orig.)

  18. Contribution to the study of nuclear fuel materials with a metallic uranium base

    International Nuclear Information System (INIS)

    Englander, M.

    1957-11-01

    In a power reactor destined to supply industrially recoverable thermal energy, the most economical source of heat still consists of natural metallic uranium. However, the nuclear fuel material, most often employed in the form of rods of 20 to 40 mm diameter, is subjected to a series of stresses which lead to irreversible distortions usually incompatible with the substructure of the reactor. As a result the fuel material must possess at the outset a certain number of qualities which must be determined. Investigations have therefore been carried out, first on the technological characters peculiar to each of the three allotropic phases of pure uranium metal, and on their interactions on the stabilisation of the material which consists of either cast uranium or uranium pile-treated in the γ phase. (author) [fr

  19. Secondary creep of porous metal supports for solid oxide fuel cells by a CDM approach

    DEFF Research Database (Denmark)

    Esposito, L.; Boccaccini, D. N.; Pucillo, G. P.

    2017-01-01

    The creep behaviour of porous iron-chromium alloy used in solid oxide fuel cells (SOFCs) becomes relevant under SOFC operating temperatures. In this paper, the secondary creep stage of infiltrated and non-infiltrated porous metal supports (MS) was investigated and theoretically modelled by a cont......The creep behaviour of porous iron-chromium alloy used in solid oxide fuel cells (SOFCs) becomes relevant under SOFC operating temperatures. In this paper, the secondary creep stage of infiltrated and non-infiltrated porous metal supports (MS) was investigated and theoretically modelled...... as function of temperature, determined by the high temperature impulse excitation technique, was directly used to account for the porosity and the related effective stress acting during the creep tests. The proposed creep rate formulation was used to extend the Crofer® 22 APU Monkman-Grant diagram...

  20. Quality assurance in the manufacture of metallic uranium fuel for research reactors

    International Nuclear Information System (INIS)

    Shah, B.K.; Kumar, Arbind; Nanekar, P.P.; Vaidya, P.R.

    2009-01-01

    Two Research Reactors viz. CIRUS and DHRUVA are operating at Trombay since 1960 and 1985 respectively. Cirus is a 40 MWth reactor using heavy water as moderator and light water as coolant. Dhruva is a 100 MWth reactor using heavy water as moderator and coolant. The maximum neutron flux of these reactors are 6.7 x 10 13 n/cm 2 /s (Cirus) and 1.8 x 10 14 n/cm 2 /s (Dhruva). Both these reactors are used for basic research, R and D in reactor technology, isotope production and operator training. Fuel material for these reactors is natural uranium metallic rods claded in finned aluminium (99.5%) tubes. This presentation will discuss various issues related to fabrication quality assurance and reactor behavior of metallic uranium fuel used in research reactors

  1. In situ ceramic layer growth on coated fuel particles dispersed in a zirconium metal matrix

    Science.gov (United States)

    Terrani, K. A.; Silva, C. M.; Kiggans, J. O.; Cai, Z.; Shin, D.; Snead, L. L.

    2013-06-01

    The extent and nature of the chemical interaction between the outermost coating layer of coated fuel particles embedded in zirconium metal during fabrication of metal matrix microencapsulated fuels were examined. Various particles with outermost coating layers of pyrocarbon, SiC, and ZrC have been investigated in this study. ZrC-Zr interaction was the least substantial, while the PyC-Zr reaction can be exploited to produce a ZrC layer at the interface in an in situ manner. The thickness of the ZrC layer in the latter case can be controlled by adjusting the time and temperature during processing. The kinetics of ZrC layer growth is significantly faster from what is predicted using literature carbon diffusivity data in ZrC. SiC-Zr interaction is more complex and results in formation of various chemical phases in a layered aggregate morphology at the interface.

  2. Experimental program on fuel rod behaviour under off-normal conditions

    International Nuclear Information System (INIS)

    Languille, A.; Cecchi, P.

    1985-01-01

    During LMFBR plant operation, fuel developments are primarily concerned with the fuel pin irradiation behaviour under steady-state conditions up to high burn-up levels. But additional studies under off-normal conditions are necessary in order to assess fuel pin performance and to define operational limits. (author)

  3. Research on plant of metal fuel fabrication using casting process (2)

    International Nuclear Information System (INIS)

    Senda, Yasuhide; Yamada, Seiya

    2005-02-01

    In this research work for the metal fuel fabrication system (38 tHM/y), the studies of the concept of the main process equipments were performed based on the previous studies on the process design and the quality control system design. In this study the handling equipment of the products were also designed, according to these designs the handling periods were evaluated. Consequently the numbers of the equipments were assessed taking into account for the method of the blending the fuel composition. (1) Structural concept design of the major equipments, the fuel handling machine and the gravimetries in the main fabrication process. The structural concept were designed for the fuel composition blending equipment, the fuel pin assembling equipment, the sodium bonding equipment, the handling equipment for fuel slug palettes, the handling equipment for fuel pins and the gravimetries. (2) Re-assessment of the numbers of the equipments taking account of the handling periods. Based on the results of item (1) the periods were evaluated for the fuel slug and pin handling. Processing time of demolder is short, then the number of it is increased to two. Three vehicles are also added to transfer the slugs and a heel smoothly. (3) Design of the buffer storages. The buffer storages among the equipments were designed through the comparison of the process speed between the equipments taking into account for the handling periods. The required amount of the structural parts (for example cladding materials) was assessed for the buffer in the same manner and the amount of the buffer facilities were optimized. (author)

  4. A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

    International Nuclear Information System (INIS)

    Yang, W.S.; Kim, T.K.; Grandy, C.

    2007-01-01

    This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% Δk. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% Δk. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

  5. Low Emissions Burner Technology for Metal Processing Industry using Byproducts and Biomass Derived Liquid Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Agrawal, Ajay; Taylor, Robert

    2013-09-30

    path forward to utilize both fossil and alternative liquid fuels in the same combustion system. In particular, experiments show that straight VO can be cleanly combusted without the need for chemical processing or preheating steps, which can result in significant economic and environmental benefits. Next, low-emission combustion of glycerol/methane was achieved by utilizing FB injector to yield fine droplets of highly viscous glycerol. Heat released from methane combustion further improves glycerol pre-vaporization and thus its clean combustion. Methane addition results in an intensified reaction zone with locally high temperatures near the injector exit. Reduction in methane flow rate elongates the reaction zone, which leads to higher CO emissions and lower NOx emissions. Similarly, higher air to liquid (ALR) mass ratio improves atomization and fuel pre-vaporization and shifts the flame closer to the injector exit. In spite of these internal variations, all fuel mixes of glycerol with methane produced similar CO and NOx emissions at the combustor exit. Results show that FB concept provides low emissions with the flexibility to utilize gaseous and highly viscous liquid fuels, straight VO and glycerol, without preheating or preprocessing the fuels. Following these initial experiments in quartz combustor, we demonstrated that glycerol combustion can be stably sustained in a metal combustor. Phase Doppler Particle Analyzer (PDPA) measurements in glycerol/methane flames resulted in flow-weighted Sauter Mean Diameter (SMD) of 35 to 40 μm, depending upon the methane percentage. This study verified that lab-scale dual-fuel burner using FB injector can successfully atomize and combust glycerol and presumably other highly viscous liquid fuels at relatively low HRR (<10 kW). For industrial applications, a scaled-up glycerol burner design thus seemed feasible.

  6. System for determining and locating a defective fuel element in a nuclear reactor

    International Nuclear Information System (INIS)

    Crosgrove, R.O.; Moody, E.; Sletten, H.L.

    1980-01-01

    A system for detecting and locating defective fuel elements in a nuclear reactor, e.g. a liquid metal fast breeder reactor (LMFBR) is described. In such a reactor, there are a large number of fuel rods situated in a radio-active core. A continuous flow of a liquid heat exchange metal, e.g. liquid sodium flows past the fuel rods. The system according to the invention first obtains a combined sample of the heat exchanger liquid, which emerges from at least one selected group of containers containing a large number of fuel rods. This combined sample is examined regarding the presence or absence of a predetermined contaminating fission material, particularly fission material emitting neutrons, which gets into the heat exchange liquid due to a defect (break) in one or more fuel rods. If the presence of a contaminating fission material is detected in the combined sample, individual samples of heat exchange liquid are taken from the individual containers, one at a time. These individual samples are also examined for the presence or absence of the contaminating fission material. Thus the fuel rod container which is responsible for the presence of the contaminating fission material in the combined sample is identified. (orig.) [de

  7. Facile and Gram-scale Synthesis of Metal-free Catalysts: Toward Realistic Applications for Fuel Cells

    Science.gov (United States)

    Kim, Ok-Hee; Cho, Yong-Hun; Chung, Dong Young; Kim, Min Jeong; Yoo, Ji Mun; Park, Ji Eun; Choe, Heeman; Sung, Yung-Eun

    2015-03-01

    Although numerous reports on nonprecious metal catalysts for replacing expensive Pt-based catalysts have been published, few of these studies have demonstrated their practical application in fuel cells. In this work, we report graphitic carbon nitride and carbon nanofiber hybrid materials synthesized by a facile and gram-scale method via liquid-based reactions, without the use of toxic materials or a high pressure-high temperature reactor, for use as fuel cell cathodes. The resulting materials exhibited remarkable methanol tolerance, selectivity, and stability even without a metal dopant. Furthermore, these completely metal-free catalysts exhibited outstanding performance as cathode materials in an actual fuel cell device: a membrane electrode assembly with both acidic and alkaline polymer electrolytes. The fabrication method and remarkable performance of the single cell produced in this study represent progressive steps toward the realistic application of metal-free cathode electrocatalysts in fuel cells.

  8. A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

    Directory of Open Access Journals (Sweden)

    Tanju Sofu

    2015-04-01

    Full Text Available The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel–coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

  9. Heaters to simulate fuel pins for heat transfer tests in single-phase liquid-metal-flow

    International Nuclear Information System (INIS)

    Casal, V.; Graf, E.; Hartmann, W.

    1976-09-01

    The development of heaters for thermal simulation of the fuel elements of liquid metal cooled fast breeder reactors (SNR) is reported. Beginning with the experimental demands various heating methods are discussed for thermodynamic investigations of the heat transfer in liquid metals. Then a preferred heater rod is derived to simulate the fuel pins of a SNR. Finally it is reported on the fabrication and the operation practice. (orig.) [de

  10. Review of Phenomenological Models for the Initial Phase HCDA Analysis in a Metal-Fueled Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Ki Rim; Ha, Kwi Seok; Chang, Won Pyo; Suk, Soo Dong

    2009-03-01

    The safety aspects of the KALIMER design results from the advanced safety performance characteristics of its ternary alloy metallic fuel. The superior thermal, mechanical, and neutronic performance of the metal-fueled core assures inherent safety response to unprotected and multiple fault accidents which are HCDA initiating events. HCDA has received great attentions because of its significant consequence, leading to substantial core disruption, although its probability of occurrence is very low. The SAS4A code provides an integrated quantitative framework for examining the phenomenological behaviors under HCDA conditions. Various phenomenological models such as prefailure characterization, transient pin response, margins to cladding failure, axial in-pin fuel relocation prior to cladding breach, and molten fuel relocation after cladding breach are required for the HCDA analysis. The important mechanisms which introduce negative reactivity during HCDA are fuel extrusion and in-pin fuel relocation, and structural feedback through thermal-mechanical neutronic effects. This report describes the safety performance characteristics of the metal fuel as observed in ex-pile and in-pile tests, and describes associated theoretical models employed into the SAS4A HCDA analysis code. Most of such tests and experiments, and development of theoretical models have been performed for the IFR program by ANL. This report provides a phenomenological basis for gaining an understanding of the metal fuel performance characteristics that obtained from expile experiments and in-pile tests. This report will provide insight and direction for planning HCDA experiments and developing theoretical models in Korea later

  11. Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining

    International Nuclear Information System (INIS)

    Herrmann, S.D.; Li, S.X.

    2010-01-01

    A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl - 1 wt% Li2O at 650 C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

  12. Test design description Volume 2, Part 1. IFR-1 metal fuel irradiation test (AK-181) element as-built data

    Energy Technology Data Exchange (ETDEWEB)

    Dodds, N. E.

    1986-06-01

    The IFR-1 Test, designated as the AK-181 Test Assembly, will be the first irradiation test of wire wrapped, sodium-bonded metallic fuel elements in the Fast Flux Test Facility (FFTF). The test is part of the Integral Fast Reactor (IFR) fuels program conducted by Argonne National Laboratory (ANL) in support of the Innovative Reactor Concepts Program sponsored by the US Department of Energy (DOE). One subassembly, containing 169 fuel elements, will be irradiated for 600 full power days to achieve 10 at.% burnup. Three metal fuel alloys (U-10Zr, U-8Pu-10Zr) will be irradiated in D9 cladding tubes. The metal fuel elements have a fuel-smeared density of 75% and each contains five slugs. The enriched zone contains three slugs and is 36-in. long. One 6.5-in. long depleted uranium axial blanket slug (DU-10Zr) was loaded at each end of the enriched zone. the fuel elements were fabricated at ANL-W and delivered to Westinghouse-Hanford for wirewrapping and assembly into the test article. This Test Design Description contains relevant data on compositions, densities, dimensions and weights for the cast fuel slugs and completed fuel elements. The elements conform to the requirements in MG-22, "Users` Guide for the Irradiation of Experiments in the FTR."

  13. Sodium water reaction R and D for French LMFBR

    International Nuclear Information System (INIS)

    Cambillard, E.; Finck, P.; Lapicore, A.; Simeon, C.

    1985-01-01

    This paper presents the research and development which is underway for the French LMFBR steam generator safety study. The program comprises three major areas: (1) the analysis of realistic leaks, which includes the leak evolution and its consequences; (2) the response time of leak detection systems compared to leak propagation phenomena; and (3) the guillotine rupture (DBA) studies relative to source term evaluation by experimental/calculational approach and mechanical calculations. This program has provided information for the demonstrations of the steam generator safety in respect to a sodium-water reaction

  14. Immersed acoustical transducers and their potential uses in LMFBR

    International Nuclear Information System (INIS)

    Argous, J.P.; Brunet, M.; Baron, J.; Lhuillier, C.; Segui, J.L.

    1980-04-01

    Six years satisfactory operation in PHENIX has proved the reliability and effectivness of under-sodium viewing (VISUS) and Acoustic Detection. This fact has been strong incentive to maintain, on the future LMFBR the visus as well as the Acoustic Detection functions. These two functions are performed on SUPER PHENIX, by two sets of distinct systems using the well-known solution. Taking into account of recent improvements in sodium immersible acoustic transducers technology, CEA decided to undertake the development of a multi-functions instrument. This paper gives an outline of this new concept, which should be able to reduce the cost and the complexity of core instrumentation

  15. Assessment of transport effects in LMFBR safety neutronics

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Ott, K.O.; Ferguson, D.R.

    1976-01-01

    A qualitative and quantitative assessment of the significance of neutron transport effects in LMFBR core disruptive accident analysis is presented. Material relocations which might cause important neutron transport behavior are identified. A quantitative measure of the error in the neutron flux is obtained from a consistent numerical comparison of transport and diffusion theory eigenvalue solutions for models of disrupted cores. A numerical technique for the prediction of transport eigenvalues and eigenvectors is formulated and applied. The technique is based on a modified diffusion theory which is fully capable of reproducing transport theory solutions

  16. The Recovery of the Metal Insulation Cable in the Instrumentation of Nuclear Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Sim, Bong Sik; Lee, Chul Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Mineral-insulated (MI) cables are widely used to prolong the instrumentation cable of instruments such as a thermocouple (TC), linear variable differential transformer (LVDT) and self-powered neutron detector (SPND), which are used to measure various irradiation characteristics of nuclear fuels and materials. MI cables are expected to be helpful for instrumentation of nuclear fuel and material irradiation because of their high electrical insulation, heat resistance and mechanical strength. The MI cable used to extend thermocouple wires is classified as the following: 1) For common metal types of thermocouples, the thermocouple extension wire is of substantially the same composition as the corresponding thermocouple type and it can offer advantages in cost or mechanical properties when used for the connection between a thermocouple and instruments. 2) For noble metal types of thermocouples, the thermocouple compensation wire is an entirely different alloy formulated to match the noble metal characteristics, which is necessary due to the high cost of noble metals. During the installation of an instrument, an MI cable damaged by impact must be recovered because it is difficult to change the entire thermocouple. And for MI cable recovery, it is necessary to develop the instrumentation technology of FTL. This paper described the experimental results of MI cable recovery, which consists of a removal test of the MI cable sheath and a joining test of the compensation of the wire and MI cable/ wire/compensation wire and sheath of MI cable/bushing, for carrying out irradiation tests of nuclear fuel and materials in the FTL facility of HANARO

  17. Metal decoration of exfoliated graphite nanoplatelets (xGnP) for fuel cell application

    Science.gov (United States)

    Do, In-Hwan

    The synthesis and characterization of metal particles at nanometer length scale has been the object of much research in modern nanotechnology due to their great impact on new nanoscale scientific and technological applications. Nanoscale metal particles possess unique optical, thermal, electronic, magnetic properties and chemical reactivity since the size of the resulting materials is on the same order as the fundamental interaction distances that give rise to physical properties and thus shows the quantum size effect which is not observed in their bulky status. Therefore, an effective synthetic method is required to obtain uniform small metal powders with controlled size and a narrow size distribution and also to produce nanocomposites consisting of either metals or metal oxides supported on carbons or metals dispersed on metal oxides for a variety of applications in chemical industries, automobiles, energy and power generating devices, hydrogen economy as well as for sensors. On the other hand, although their excellent mechanical, thermal and electrical conductivity, excellent corrosion and oxidation resistance, and low impurity levels which are required as a breakthrough material to increase performance of next generation energy devices, exfoliated graphite nanoplatelet (xGnP) has not been studied as deeply as recent new nano structured carbon materials such as single wall carbon nanotubes (SWNT), multi-wall carbon nanotubes (MWNT), carbon nanohorn (CNH), graphite nanofiber (GNF), and fullerenes. In addition, xGnP is much cost-effective compared to other carbon nanostructures. Hence, it is interesting to evaluate the applicability of xGnP as a support material for fuel cell which is one of promising energy devices for the future. In this research, a new simple, efficient and economic way is presented for the synthesis of noble metal nanoparticles such as Pt, Ru, Pd, etc and their deposition on various carbon supports and metal oxides via microwave heating in the

  18. Fabrication experience of sodium bonded metallic test fuel with U-6%Zr in T91 clad tube for irradiation in FBTR

    International Nuclear Information System (INIS)

    Muralidaran, P.; Prabhu, T.V.; Manivannan, A.; Vinod, A.V.; Padmanabhan, R.; Ravisankar, G.; Ganesan, V.; Vasudeva Rao, P.R.

    2012-01-01

    Future fast reactors in India will be fuelled by metallic alloys to achieve the enhanced breeding necessary to ensure rapid growth of power generation. Since, experience in fabrication and irradiation of metallic alloy fuel does not exist in the country, it is necessary to generate fabrication and irradiation experience before launching commercial FBRs with metallic alloy fuel. As part of this development, sodium bonded metallic test fuel pin fabrication was taken up at IGCAR. This paper describes the fabrication experience of sodium bonded metallic test fuel pin using Nat. U-6%Zr and T91 clad tube towards test irradiation in FBTR at Chemistry Group

  19. Numerical evaluation of micro-structural parameters of porous supports in metal-supported solid oxide fuel cells

    DEFF Research Database (Denmark)

    Reiss, Georg; Frandsen, Henrik Lund; Brandstätter, Wilhelm

    2014-01-01

    Metallic supported Solid Oxide Fuel Cells (SOFCs) are considered as a durable and cost effective alternative to the state-of-the-art ceramic supported cell designs. In order to understand the mass and charge transport in the metal-support of this new type of cell a novel technique involving X...

  20. Application of ASA supported noble metal catalysts in the deep hydrodesulphurisation of diesel fuel

    Energy Technology Data Exchange (ETDEWEB)

    Reinhoudt, H.R.; Troost, R.; Van Schalkwijk, S.; Van Langeveld, A.D.; Sie, S.T.; Moulijn, J.A. [Delft University of Technology, Delft (Netherlands); Schulz, H. [Universitaet Karlsruhe, Engler Bunte Institut, Karlsruhe (Germany); Chadwick, D. [Imperial College of Science, Technology and Medicine, London (United Kingdom); Cambra, J. [Escuela de Ingenierios, Bilbao (Spain); De Beer, V.H.J.; Van Veen, J.A.R. [Eindhoven University of Technology, Eindhoven (Netherlands); Fierro, J.L.G. [C.S.I.C., Madrid (Spain)

    1997-07-01

    The potential of Amorphous Silica Alumina (ASA) supported Pt and Pd catalysts for deep hydrodesulphurisation (HDS) of diesel fuels was investigated. It appeared that the ASA supported catalysts exhibit an excellent activity for the conversion of 4-Ethyl, 6-Methyl Dibenzothiophene (4-E,6-M DBT) under model conditions as compared to conventional HDS catalysts and {gamma}-Al{sub 2}O{sub 3} supported noble metal catalysts. Pt/ASA was also tested under practical conditions using a diesel fuel feed. The Pt/ASA catalyst showed a comparable activity to the NiW/{gamma}-Al{sub 2}O{sub 3} catalyst which was higher than that of the conventional CoMo/{gamma}-Al{sub 2}O{sub 3} catalyst. The main difference of the catalyst was the better hydroconversion of the 4,6 di-alkylated DBT's. The better performance of Pt/ASA in the testing under model conditions as compared to the diesel fuel HDS can be attributed to poisoning of part of the active phase by basic nitrogen compounds like quinoline. It is concluded that ASA supported noble metal catalysts have a promising potential for deep HDS processing. 5 refs.

  1. 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

  2. Visual in-pile fuel disruption experiments

    International Nuclear Information System (INIS)

    Cano, G.L.; Ostensen, R.W.; Young, M.F.

    1978-01-01

    In a loss-of-flow (LOF) accident in an LMFBR, the mode of disruption of fuel may determine the probability of a subsequent energetic excursion. To investigate these phenomena, in-pile disruption of fission-heated irradiated fuel pellets was recorded by high speed cinematography. Instead of fuel frothing or dust-cloud breakup (as used in the SAS code) massive and very rapid fuel swelling, not predicted by analytical models, occurred. These tests support massive fuel swelling as the initial mode of fuel disruption in a LOF accident. (author)

  3. Supply and cost factors for metals in the Canadian nuclear fuel waste immobilization program

    International Nuclear Information System (INIS)

    McConnell, D.B.

    1982-11-01

    Estimates have been made of the demand for immobilization containers to accommodate the irradiated fuel bundles arising from Canadian nuclear generating stations to the year 2020. The resulting estimates for container shells and container-filling alloys were compared to estimates for Canadian and Western World production of the candiate metals. The results indicate that, among the container shell metals, supply difficulties might arise only for Grade 7 titanium. Among the filling metals, only lead-antimony alloy might present supply problems. Current cost figures for plate made of each shell metal, and bulk quantities of filling metals, were compared. Materials costs would be least for a supported shell of stainless steel, followed by copper, titanium alloys Grades 2, 12 and 7, and Inconel 625. Aluminum-silicon is the lowest-cost filling matrix, followed by zinc, lead, and lead-antimony. Container durability, vault conditions, groundwater composition and other factors may play an overriding role in the final selection of materials for container construction

  4. Metal ferrite oxygen carriers for chemical looping combustion of solid fuels

    Energy Technology Data Exchange (ETDEWEB)

    Siriwardane, Ranjani V.; Fan, Yueying

    2017-01-31

    The disclosure provides a metal ferrite oxygen carrier for the chemical looping combustion of solid carbonaceous fuels, such as coal, coke, coal and biomass char, and the like. The metal ferrite oxygen carrier comprises MFe.sub.xO.sub.y on an inert support, where MFe.sub.xO.sub.y is a chemical composition and M is one of Mg, Ca, Sr, Ba, Co, Mn, and combinations thereof. For example, MFe.sub.xO.sub.y may be one of MgFe.sub.2O.sub.4, CaFe.sub.2O.sub.4, SrFe.sub.2O.sub.4, BaFe.sub.2O.sub.4, CoFe.sub.2O.sub.4, MnFeO.sub.3, and combinations thereof. The MFe.sub.xO.sub.y is supported on an inert support. The inert support disperses the MFe.sub.xO.sub.y oxides to avoid agglomeration and improve performance stability. In an embodiment, the inert support comprises from about 5 wt. % to about 60 wt. % of the metal ferrite oxygen carrier and the MFe.sub.xO.sub.y comprises at least 30 wt. % of the metal ferrite oxygen carrier. The metal ferrite oxygen carriers disclosed display improved reduction rates over Fe.sub.2O.sub.3, and improved oxidation rates over CuO.

  5. Carbon nanotubes paste sensor modified with bismuth film for determination of metallic ions in ethanol fuel

    Directory of Open Access Journals (Sweden)

    Felipe Augusto Gorla

    2015-05-01

    Full Text Available In the present study an anodic stripping voltammetric method using a bismuth film modified carbon nanotubes paste electrode for simultaneous determination of metals Zn2+, Cd2+and Pb2+in ethanol fuel is described. The metallic ions were preconcentrated on the bismuth film in the time and deposition potential of 500 s and -1.2 V and the stripping step was carried out by square wave voltammetry (frequency of 15 Hz, pulse amplitude of 25 mV and potential step of 5 mV. Acetate buffer at 0.1 mol L-1concentration and pH 4.5 was used as support electrolyte. The method showed linearity including the analytical blank up to 48.39 ?g L-1 for the metals and the obtained limits of detection were 3.36, 0.32 and 0.47 ?g L-1for Zn2+, Cd2+and Pb2+, respectively. The proposed method was applied in ethanol fuel samples.

  6. CRAB-II: a computer program to predict hydraulics and scram dynamics of LMFBR control assemblies and its validation

    International Nuclear Information System (INIS)

    Carelli, M.D.; Baker, L.A.; Willis, J.M.; Engel, F.C.; Nee, D.Y.

    1982-01-01

    This paper presents an analytical method, the computer code CRAB-II, which calculates the hydraulics and scram dynamics of LMFBR control assemblies of the rod bundle type and its validation against prototypic data obtained for the Clinch River Breeder Reactor (CRBR) primary control assemblies. The physical-mathematical model of the code is presented, followed by a description of the testing of prototypic CRBR control assemblies in water and sodium to characterize, respectively, their hydraulic and scram dynamics behavior. Comparison of code predictions against the experimental data are presened in detail; excellent agreement was found. Also reported are experimental data and empirical correlations for the friction factor of the absorber bundle in the entire flow range (laminar to turbulent) which represent an extension of the state-of-the-art, since only fuel and blanket assemblies friction factor correlations were previously reported in the open literature

  7. Frequency interpretation of hold-time experiments on high temperature low-cycle fatigue of steels for LMFBR

    International Nuclear Information System (INIS)

    Udoguchi, T.; Asada, Y.; Ichino, I.

    1975-01-01

    The effect of frequency or hold-time on the low-cycle fatigue strength of AISI 316 stainless steel and SCM 3 Cr--Mo steel for fuel cladding, piping, and other structural members of LMFBR is investigated under high temperature conditions. Push-pull fatigue tests are conducted in air under conditions of fully reversed axial strain-control with a tensile strain hold-time ranging fromm 0 to 120 min for AISI 316, and with a tensile and an equal compressive strain hold-time ranging from 0 to 995 s for SCM 3. In these tests, a decrease of fatigue life is observed as the hold-time is increased. An empirical formula is presented which can predict well the effect of hold-time on high temperature low-cycle fatigue life in terms of frequency. The formula is a little different from those in the literature

  8. A frequency interpretation of hold-time experiments on high temperature low-cycle fatigue of steels for LMFBR

    International Nuclear Information System (INIS)

    Udoguchi, T.; Asada, Y.; Ichino, I.

    1975-01-01

    The effect of frequency or hold-time on the low-cycle fatigue strength of AISI 316 stainless steel and SCM 3 Cr-Mo steel for fuel cladding, piping and other structural members of LMFBR is investigated under high temperature conditions. Push-pull fatigue tests are conducted in air under conditions of fully reversed axial strain-control with a tensile strain hold-time ranging from 0 to 120 min for AISI 316, and with a tensile and an equal compressive strain hold-time ranging from 0 to 995 s for SCM 3. In these tests, a considerable decrease of fatigue life is observed as the hold-time is increased. An empirical formula is presented which can predict well the effect of hold-time on high temperature low-cycle fatigue life in terms of frequency. The formula is a little different from those in the literature. (author)

  9. An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

    Directory of Open Access Journals (Sweden)

    Wuseong You

    2017-12-01

    Full Text Available In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

  10. Manufacture of hypoeutectic Al-Si metal powders for dispersion matriz in nuclear fuels

    International Nuclear Information System (INIS)

    Raffaeli, H A; Harri, S; Acosta, M; Castillo Guerra, R; Rossi, G; Fabro, J O; Rubiolo, G H

    2012-01-01

    Within the framework of the development of low enriched nuclear fuels for research reactors, U.Mo/Al is the most promising option that has however to be optimized. Indeed at the U.Mo/Al interfaces between U.Mo particles and the Al matrix, an interaction layer grows under irradiation inducing an unacceptable fuel swelling. Adding silicon in limited content into the Al matrix has clearly improved the in-pile fuel behaviour. This breakthrough is attributed to an U.Mo/Al.Si protective layer around U.Mo particles appeared during fuel manufacturing. The present work deals with three techniques applied to produce metal powders of hypoeutectic Al-Si alloys: ball milling, centrifugal atomization and gas atomization. Size and microstructure of the particles are analyzed in the three techniques. The best result is found with the gas atomization system, flakes and rods morphology predominates in the produced powders, with particle sizes below 150 microns and the greater mass population (65%) is between 150 and 125 microns. The particle surface is smooth and the high solidification rate provides a good distribution of the α-Al primary and eutectic phase within each particle (author)

  11. Contribution to the study of the fission-gas release in metallic nuclear fuels

    International Nuclear Information System (INIS)

    Kryger, B.

    1969-10-01

    In order to study the effect of an external pressure on the limitation of swelling due to fission-gas precipitation, some irradiations have been carried out at burn-ups of about 35.000 MWd/ton, and at average sample temperatures of 575 Celsius degrees, of non-alloyed uranium and uranium 8 per cent molybdenum gained in a thick stainless steel can. A cylindrical central hole allows a fuel swelling from 20 to 33 per cent according to the experiment. After irradiation, the uranium samples showed two types of can rupture: one is due to the fuel swelling, and the other, to the pressure of the fission gases, released through a network of microcracks. The cans of the uranium-molybdenum samples are all undamaged and it is shown that the gas release occurs by interconnection of the bubbles for swelling values higher than those obtained in the case of uranium. For each type of fuel, a swelling-fission gas release relationship is established. The results suggest that good performances with a metallic fuel intended for use in fast reactor conditions can be obtained. (author) [fr

  12. Function of all-metal separators for waste fuels. Phase 1; Funktion av allmetallseparatorer foer avfallsbraenslen. Etapp 1

    Energy Technology Data Exchange (ETDEWEB)

    Jacoby, Juergen; Wrangensten, Lars

    2004-08-01

    Various waste incineration facilities, which use different types of waste fuels, have difficulties with a high content of non-magnetic metal, especially aluminum in their fuels. Aluminum may melt on the grate and can lead to corrosion or fouling in the furnace. Additionally, a high content of aluminum in the flyash may cause difficulties in terms of storage or further use of the ash as e.g. construction material. The industrial demand for efficient separators for non-magnetic metals from a fuel stream is rather large. There is however some uncertainty in the performance and efficiency of metal separators. Two types of separators can be found, the first type is called eddy current separator, the other type is based upon a metal detector with a sorting unit in the form of a chute or similar afterwards. An eddy current separator consists of a fast rotating drum containing several permanent magnets with alternating polarity. Due to the rotation, the change in the magnetic field induces eddy currents in conducting materials. The eddy currents cause a force in non-magnetic metal, the Lorentz force, which repels the material away from the rotating drum while all other material follows the systems flow direction. Systems equipped with a metal detector activate a mechanical sorting device, separate chute or air nozzles, when a metal particle is detected. In contrast to eddy current separators all types of metals can be detected and sorted out by systems based on metal detector. Several technical solutions for metal separation supplied by various manufacturers are described in the report. The companies have been asked to supply product information on the working principle, technical data, efficiency and limits for different types of metals. Two reference power plants have been visited and their experiences with all-metal separators are described. Haendeloeverket in Norrkoeping uses eddy current separators for separation of non-magnetic metals from household waste

  13. A combinatorial study on catalytic synergism in supported metal catalysts for fuel cell technology

    Science.gov (United States)

    Kobayashi, Tetsuhiko; Ueda, Atsushi; Yamada, Yusuke; Shioyama, Hiroshi

    2004-02-01

    In order to accelerate the catalyst development for the increasing demand on the fuel cell technology, it has been attempted to adopt a combinatorial approach. The catalytic synergism, often observed on the supported metal catalysts for the fuel cell utilization, has been subjected to study. It is proposed herein that not only a comparison of catalysts in one reaction, but also the comparison of interrelated reactions by use of a common catalyst library brings about important information to elucidate the catalytic synergism. Preliminary results of the comparison between the water-gas shift reaction and the steam reforming of MeOH on a given set of catalyst library are presented. An important indicator to predict the serendipitous synergism is expected to be obtained from such information by use of artificial intelligence.

  14. Method to measure the oxygen-to-metal ratio in oxidic nuclear fuels

    International Nuclear Information System (INIS)

    Vollath, D.

    1976-01-01

    A method to measure the oxygen-to-metal ratio in oxidic nuclear fuels is proposed which is very exact, requires only a short measuring time, and can be automatized. At temperatures between 650 0 C and 1,300 0 C, the oxygen potential of the nuclear fuel is measured electrochemically using a solid electrolyte and a reference substance with known oxygen potential. The result is derived by computer from the measured value and a calibration curve. A suitable solid electrolyte is a thorium oxide doped with yttrium oxide and calcium oxide. The arrangement to carry out the measuring process is described in detail; a very suitable measuring chain is Pt - Fe/FeO - ThO 2 (Y 2 O 3 , CaO) - ZrO 2 (CaO) - ThO 2 (Y 2 O 3 , CaO) - sample - Pt. (UWI) [de

  15. Towards High Power Density Metal Supported Solid Oxide Fuel Cell for Mobile Applications

    DEFF Research Database (Denmark)

    Nielsen, Jimmi; Persson, Åsa H.; Muhl, Thuy Thanh

    2018-01-01

    For use of metal supported solid oxide fuel cell (MS-SOFC) in mobile applications it is important to reduce the thermal mass to enable fast startup, increase stack power density in terms of weight and volume and reduce costs. In the present study, we report on the effect of reducing the Technical...... loss, two different routes for increasing the porosity of the support layer and thus performance were explored. The first route is the introduction of gas channels by puncturing of the green tape casted support layer. The second route is modification of the co-sintering profile. In summary, the cell...... thickness and thus weight and volume was reduced and the cell power density at 0.7 V at 700°C was increased by 46% to 1.01 Wcm−2 at a fuel utilization of 48%. All modifications were performed on a stack technological relevant cell size of 12 cm × 12 cm....

  16. Proposal for computer investigation of LMFBR core meltdown accidents

    International Nuclear Information System (INIS)

    Boudreau, J.E.; Harlow, F.H.; Reed, W.H.; Barnes, J.F.

    1974-01-01

    The environmental consequences of an LMFBR accident involving breach of containment are so severe that such accidents must not be allowed to happen. Present methods for analyzing hypothetical core disruptive accidents like a loss of flow with failure to scram cannot show conclusively that such accidents do not lead to a rupture of the pressure vessel. A major deficiency of present methods is their inability to follow large motions of a molten LMFBR core. Such motions may lead to a secondary supercritical configuration with a subsequent energy release that is sufficient to rupture the pressure vessel. The Los Alamos Scientific Laboratory proposes to develop a computer program for describing the dynamics of hypothetical accidents. This computer program will utilize implicit Eulerian fluid dynamics methods coupled with a time-dependent transport theory description of the neutronic behavior. This program will be capable of following core motions until a stable coolable configuration is reached. Survey calculations of reactor accidents with a variety of initiating events will be performed for reactors under current design to assess the safety of such reactors

  17. Analytical approach for confirming the achievement of LMFBR reliability goals

    International Nuclear Information System (INIS)

    Ingram, G.E.; Elerath, J.G.; Wood, A.P.

    1981-01-01

    The approach, recommended by GE-ARSD, for confirming the achievement of LMFBR reliability goals relies upon a comprehensive understanding of the physical and operational characteristics of the system and the environments to which the system will be subjected during its operational life. This kind of understanding is required for an approach based on system hardware testing or analyses, as recommended in this report. However, for a system as complex and expensive as the LMFBR, an approach which relies primarily on system hardware testing would be prohibitive both in cost and time to obtain the required system reliability test information. By using an analytical approach, results of tests (reliability and functional) at a low level within the specific system of interest, as well as results from other similar systems can be used to form the data base for confirming the achievement of the system reliability goals. This data, along with information relating to the design characteristics and operating environments of the specific system, will be used in the assessment of the system's reliability

  18. Transient analysis of LMFBR reinforced/prestressed concrete containment

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Belytschko, T.B.; Bazant, Z.P.

    1979-01-01

    The use of prestressed concrete reactor vessels (PCRVs) for LMFBR containment creates a need for analytical methods for treating the transient response of such structures, for LMFBR containments must be capable of sustaining the dynamic effects which arise in a hypothetical core disruptive accident (HCDA). These analyses require several unique features: a model of concrete which includes tensile cracking, a methodology for representing the prestressing tendons and for simulating the prestressing operation, and an efficient computational tool for treating the transient response. Furthermore, for the sake of convenience, all of these features should be available in a single computer code. For the purpose of treating the transient response, a finite element program with explicit time integration was chosen. The use of explicit time integration has the advantage that it can easily treat the complicated constitutive model which arises from the considerations of concrete cracking and it can handle the slip between reinforcing tendons and the concrete through the use of the well known sliding interface options. However, explicit time integration programs are usually not well suited to the simulation of static processes such as prestressing. Nevertheless, explicit time integration programs can handle static processes through the introduction of damping by what is known as a dynamic relaxation procedure. For this reason, the dynamic relaxation procedure was refined through the introduction of lumped mass, viscous damping. This provision made the prestressing operation of the concrete structures by means of the explicit formulation rather convenient. (orig.)

  19. Fuel and core design study of the sodium-cooled fast reactors. Studies on metallic fuel cores in the JFY2002

    International Nuclear Information System (INIS)

    Sugino, Kazuteru; Mizuno, Tomoyasu

    2003-06-01

    Based on the results obtained in the former feasibility study, the metallic fueled core of ordinary-type, that is, 2-region homogeneous core, has been established aiming at the improvement in the core performance, and subsequent comparison has been performed with the mixed oxide fueled core. Further, the attractive concept of the metallic fueled core of high outlet temperature has been constructed which has good nuclear features as a metallic fueled core and has identical outlet temperature to mixed oxide fuelled core. Following items have been found as a result of the investigation on the ordinary-type core. The metallic fueled core whose maximum fast neutron fluence (En>0.1MeV) is set identical (5x10 23 n/cm 2 ) to the mixed oxide fueled cores with core discharge burnup 150GWd/t has sufficient core performances as a metallic fueled core, e.g. higher breeding ratio and longer operation period compared with mixed oxide fueled cores, but the core discharge burnup is limited up to 100GWd/t. However effective discharge burnup including the contribution of the blanket region is comparative to mixed oxide cores under the same breeding ratio condition. In order to enlarge the core discharge burnup to 150GWd/t keeping the core performance identical to above mentioned core's, the irradiation deformation of structural material should be reduced to that of mixed oxide fueled cores. Further the maximum fast neutron fluence reaches to 7-8x10 23 n/cm 2 (En>0.1MeV). The investigations on the core of high outlet temperature have clarified following items. Even in the change of core regions by pin-diameter form 3-region to 2-region and in the limited maximum fuel pin diameter 8.5 mm, realization of the identical outlet/inlet temperatures to the mixed oxide cores (550/395degC) is feasible under the criteria of the maximum temperature 650degC at the inner surface of the cladding. The constructed core accommodates the targets of breeding ratio from about 1.0 to 1.2 only by adjusting

  20. Application of 2-1/4 Cr-1 Mo as a structural material in saturated steam cycle LMFBR systems. Final report

    International Nuclear Information System (INIS)

    Licina, G.J.; Busboom, H.J.; Ring, P.J.; Roy, P.; Schmidt, C.G.; Spalaris, C.N.

    1982-02-01

    The suitability and incentives were examined for using 2-1/4Cr-1Mo steel as a structural material for the entire primary and secondary sodium systems in a 1000 MWe pool-type Liquid Metal Fast Breeder Reactor. The critical properties, advantages and disadvantages of 2-1/4Cr-1Mo, and data needed for design were described for each major component in the reactor. The relative importance of alloy properties to the successful use of ferritics in LMFBR was identified. Licensing issues, likely to surface if ferritic alloys were to be used for critical reactor components, were discussed

  1. LMFBR safety testing needs and the conceptual design of a new safety research experiment facility

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Matlock, R.G.; Goldman, A.J.

    1975-09-01

    Experiment needs for the LMFBR safety program are reviewed. The screening of reactor concepts which would meet the needs is described and a conceptual design for a new safety research experiment facility is presented

  2. State of the art review of degradation processes in LMFBR materials. Volume II. Corrosion behavior

    International Nuclear Information System (INIS)

    Dillon, R.D.

    1975-01-01

    Degradation of materials exposed to Na in LMFBR service is reviewed. The degradation processes are discussed in sections on corrosion and mass transfer, erosion, wear and self welding, sodium--water reactions, and external corrosion. (JRD)

  3. Metal Phosphates as Proton Conducting Materials for Intermediate Temperature Fuel Cell and Electrolyser Applications

    DEFF Research Database (Denmark)

    Anfimova, Tatiana

    of phosphates were systematically reviewed including solid acids or alkali hydrogen phosphates, pyrophosphates, and rare earth metal phosphates. Demonstration of the fuel cell technology based on solid acid proton conductor CsH2PO4 has inspired the active research in the area. Based on the literature survey....... The conductivity and its stability are studied and correlated with the phosphate morphologies. The additional solid state NMR studies have been performed in collaboration with Southern Denmark University (SDU). Chapter 8 presents the result obtained for a novel proton conductor based on cerium ultraphosphate...

  4. Modeling the behavior of metallic fast reactor fuels during extended transients

    International Nuclear Information System (INIS)

    Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

    1992-01-01

    Passive safety features in the metal-fueled Integral Fast Reactor (IFR) make it possible to avoid core damage for extended time periods even when automatic scram systems fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements

  5. A Review of Metallic Bipolar Plates for Proton Exchange Membrane Fuel Cells: Materials and Fabrication Methods

    Directory of Open Access Journals (Sweden)

    Shahram Karimi

    2012-01-01

    Full Text Available The proton exchange membrane fuel cell offers an exceptional potential for a clean, efficient, and reliable power source. The bipolar plate is a key component in this device, as it connects each cell electrically, supplies reactant gases to both anode and cathode, and removes reaction products from the cell. Bipolar plates have been fabricated primarily from high-density graphite, but in recent years, much attention has been paid to developing cost-effective and feasible alternative materials. Two different classes of materials have attracted attention: metals and composites. This paper offers a comprehensive review of the current research being carried out on metallic bipolar plates, covering materials and fabrication methods.

  6. Method of improving fuel cell performance by removing at least one metal oxide contaminant from a fuel cell electrode

    Science.gov (United States)

    Kim, Yu Seung [Los Alamos, NM; Choi, Jong-Ho [Los Alamos, NM; Zelenay, Piotr [Los Alamos, NM

    2009-08-18

    A method of removing contaminants from a fuel cell catalyst electrode. The method includes providing a getter electrode and a fuel cell catalyst electrode having at least one contaminant to a bath and applying a voltage sufficient to drive the contaminant from the fuel cell catalyst electrode to the getter electrode. Methods of removing contaminants from a membrane electrode assembly of a fuel cell and of improving performance of a fuel cell are also provided.

  7. Electron probe microanalysis of a METAPHIX UPuZr metallic alloy fuel irradiated to 7.0 at.% burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Brémier, S., E-mail: stephan.bremier@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Inagaki, K. [Central Research Institute of Electric Power Industry, Nuclear Technology Research Laboratory, 2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511 (Japan); Capriotti, L.; Poeml, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Ogata, T.; Ohta, H. [Central Research Institute of Electric Power Industry, Nuclear Technology Research Laboratory, 2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511 (Japan); Rondinella, V.V. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany)

    2016-11-15

    The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15–20 μm and migration of cladding elements to the fuel. - Highlights: • Electron Probe MicroAnalysis of a UPuZr metallic fuel alloy irradiated to 7.0 at.% burn-up. • Significant redistribution of the fuel components along the fuel radius, nearly complete depletion of Zr in the central part of the fuel. • Interactions between the fuel and the cladding with occurrence of a number of intermediary layers and migration of cladding elements to the fuel. • Safe irradiation behaviour of the base alloy fuel.

  8. LMFBR self-activated shutdown systems

    International Nuclear Information System (INIS)

    Sowa, E.S.; Barthold, W.P.; Eggen, D.T.; Huebotter, P.R.; Josephson, J.; Pizzica, P.A.; Turski, R.B.; van Erp, J.B.

    1976-01-01

    Self-actuated shutdown systems (SASSs), fully contained within the dimensions of a fuel subassembly and installed in the core in judiciously chosen locations, can provide an important additional safety feature for LMFBRs. If actuated by phenomena inherent to the system and its immediate environment, these systems can contribute considerably to the total reliability of the overall plant protection system, in particular as regards protection against human error. It was shown that this type of shutdown system is capable of inserting a substantial amount of negative reactivity into the core with a relatively small impact on plant performance. Furthermore, it was shown that a coolable geometry can be maintained in LMFBRs of current design for a wide spectrum of accident initiators, and for a range of response times and insertion rates which appear to be achievable within practical design limits. Experiments showed that Curie-point-operated devices have considerable promise for application in self-actuated shutdown systems, in particular as regards meeting the requirements of testability and resettability

  9. Research on treatment of wastewater containing heavy metal by microbial fuel cell

    Science.gov (United States)

    Chen, Zixuan; Lu, Xun; Yin, Ruixia; Luo, Yunyi; Mai, Hanjian; Zhang, Nan; Xiong, Jingfang; Zhang, Hongguo; Tang, Jinfeng; Luo, Dinggui

    2018-02-01

    With rapid development of social economy, serious problem has been caused by wastewater containing heavy metals, which was difficult to be treated by many kinds of traditional treatment methods, such as complex processes, high cost or easy to cause secondary pollution. As a novel biological treatment technology, microbial fuel cells (MFC) can generate electric energy while dealing with wastewater, which was proposed and extensively studied. This paper introduced the working principle of MFC, the classification of cathode, and the research progress on the treatment of wastewater containing Cr(VI), Cu(II), Ag(I), Mn(II) and Cd(II) by MFC. The study found that different cathode, different heavy metals anddifferent hybrid systems would affect the performance of the system and removal effect for heavy metal in MFC. MFC was a highly potential pollution control technology. Until now, the research was still in the laboratory stage. Its industrial application for recovery of heavy metal ion, improving the energy recovery rate and improvement or innovation of system were worthy of further research.

  10. Monte-Carlo Modeling of Parameters of a Subcritical Cascade Reactor Based on MSBR and LMFBR Technologies

    CERN Document Server

    Bznuni, S A; Zhamkochyan, V M; Polanski, A; Sosnin, A N; Khudaverdyan, A H

    2001-01-01

    Parameters of a subcritical cascade reactor driven by a proton accelerator and based on a primary lead-bismuth target, main reactor constructed analogously to the molten salt breeder (MSBR) reactor core and a booster-reactor analogous to the core of the BN-350 liquid metal cooled fast breeder reactor (LMFBR). It is shown by means of Monte-Carlo modeling that the reactor under study provides safe operation modes (k_{eff}=0.94-0.98), is apable to transmute effectively radioactive nuclear waste and reduces by an order of magnitude the requirements on the accelerator beam current. Calculations show that the maximal neutron flux in the thermal zone is 10^{14} cm^{12}\\cdot s^_{-1}, in the fast booster zone is 5.12\\cdot10^{15} cm^{12}\\cdot s{-1} at k_{eff}=0.98 and proton beam current I=2.1 mA.

  11. Program for the development of design data: LMFBR steam generator materials

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-01-01

    Progress in fiscal year 1976 is reported for two tasks in the LMFBR steam generator materials development program. The primary objective of Task 18 is characterization of materials, mainly 2-1/4 Cr-1 Mo and austenitic stainless steels, to assure that satisfactory materials compatibility is achieved with LMFBR steam generator environments. Studies described include: the kinetics and magnitude of decarburization of 2-1/4 Cr-1 Mo in high temperature liquid sodium; decarburization effects on mechanical properties; stress corrosion susceptibility of 2-1/4 Cr-1 Mo in water/steam environments of LMFBR steam generators; and stainless steel mechanical properties within a carburizing sodium environment at high temperatures. In addition, two activities are in progress that directly support LMFBR steam generator design and fabrication. Under Subtask E, General Electric heads the working committee on steam generator materials for the Nuclear Systems Materials Handbook. Subtask F, an experimental effort, is investigating thermal degradation effects produced by particulate deposition (in sodium) on heat transfer surfaces. The Steam Generator Materials Qualification Task (Task 10-G) was initiated in support of the development of materials and processes used in Clinch River Breeder Reactor steam generators. The topics discussed are: LMFBR steam generator tubing; LMFBR steam generator tubesheet forgings (characterization of VAR/ESR melted 2-1/4 Cr-1 Mo); friction, wear and self-welding; nondestructive examination development; water chemistry studies; and transition joint development.

  12. Do trace metals (chromium, copper, and nickel) influence toxicity of diesel fuel for free-living marine nematodes?

    Science.gov (United States)

    Hedfi, Amor; Boufahja, Fehmi; Ben Ali, Manel; Aïssa, Patricia; Mahmoudi, Ezzeddine; Beyrem, Hamouda

    2013-06-01

    The objective of this study was to test the hypotheses that (1) free-living marine nematodes respond in a differential way to diesel fuel if it is combined with three trace metals (chromium, copper, and nickel) used as smoke suppressants and that (2) the magnitude of toxicity of diesel fuel differs according to the level of trace metal mixture added. Nematodes from Sidi Salem beach (Tunisia) were subjected separately for 30 days to three doses of diesel fuel and three others of a trace metals mixture. Simultaneously, low-dose diesel was combined with three amounts of a trace metal mixture. Results from univariate and multivariate methods of data evaluation generally support our initial hypothesis that nematode assemblages exhibit various characteristic changes when exposed to different types of disturbances; the low dose of diesel fuel, discernibly non-toxic alone, became toxic when trace metals were added. For all types of treatments, biological disturbance caused severe specific changes in assemblage structure. For diesel fuel-treated microcosms, Marylynnia bellula and Chromaspirinia pontica were the best positive indicative species; their remarkable presence in given ecosystem may predict unsafe seafood. The powerful toxicity of the combination between diesel fuel and trace metals was expressed with only negative bioindicators, namely Trichotheristus mirabilis, Pomponema multipapillatum, Ditlevsenella murmanica, Desmodora longiseta, and Bathylaimus capacosus. Assemblages with high abundances of these species should be an index of healthy seafood. When nematodes were exposed to only trace metals, their response looks special with a distinction of a different list of indicative species; the high presence of seven species (T. mirabilis, P. multipapillatum, Leptonemella aphanothecae, D. murmanica, Viscosia cobbi, Gammanema conicauda, and Viscosia glabra) could indicate a good quality of seafood and that of another species (Oncholaimellus mediterraneus) appeared an

  13. An ultrasonic fuel identification system for liquid metal cooled reactors resilient against multiple transducer failures

    International Nuclear Information System (INIS)

    Van-Dyck, Dries; Dierckx, Marc

    2013-06-01

    We describe a fuel assembly identification system developed for the MYRRHA reactor - a new multi-purpose flexible irradiation facility to replace the aging BR2. MYRRHA is a fast spectrum research reactor cooled with lead-bismuth eutectic (LBE) and conceived as an accelerator driven system capable of operating in sub-critical and critical modes. As liquid metal is opaque to visual light, the conventional optical fuel assembly identification system, as used by water cooled reactors, has to be replaced by a system not hindered by the opacity of the coolant. As already suggested in the late sixties, we use ultrasound for this purpose and present an encoding especially designed to enhance the robustness of the ultrasonic read-out. The encoding is based on notches of varying depth on the inflow nozzle of a fuel assembly. The depth of each notch is used to encode two bits and is measured by a dedicated transducer aligned over the notch. To increase the reliability of the fuel identification process, the identification number is protected by an error correcting code based on Hamming codes. We describe the ultrasonic system used to read out the vector of depths which is subsequently converted to a vector of bits. We explain the encoding of the twelve bit fuel identification numbers to a 22-bit error correcting code and discuss how Hamming decoding can be used to correct single bit errors, detect two bit errors or fill in the missing bits of a failing transducer. We also present a method based on solving a linear system over Boolean variables to (partially) reconstruct the fuel identification number in case multiple transducers fail. We show that the probability on full reconstruction is 100% for up to two transducer failures, 98% for three, 79% for four and 20% for five failing transducers. Finally, we present validation results in water and lead-bismuth eutectic for the differential measurement method used to measure the depth of the notches which form the basis for the

  14. Rare metal fission products in nuclear spent fuel as catalysts for hydrogen production by water electrolysis

    International Nuclear Information System (INIS)

    Ozawa, Masaki

    2004-01-01

    Separation and utilization of rare metal fission products (RMFP) in nuclear spent fuel were studied to apply them as a catalyst for hydrogen generation by water electrolysis. The RMFP, namely Pd, Ru, Rh and Tc, etc, are abundant, more than ca. 30kg per metric ton of a typical fast reactor spent fuel. The RMFP can be selectively separated from high level liquid waste (HLLW) by catalytic electrolytic extraction (CEE) method. Specific metallic cations such as Pd 2+ , which originate in the solutions, may act as promoters (i.e., Pd adatom ) or mediators, thereby accelerating electrochemical deposition of RuNO 3+ , Rh 3+ and ReO 4 - (simulator TcO 4 - ). In utilizing CEE method, electrodeposited electrodes were prepared, and successively dedicated to the water (alkaline or artificial sea water) electrolysis tests. Among the RMFP deposited electrodes, maximum potential shifting for hydrogen evolution to noble side was observed for the quaternary, Pd-Ru-Rh-Re (3.5:4:1:1), deposit Pt electrode, with suggesting the highest cathodic currents for hydrogen evolution both in alkaline solution and artificial sea water. The electro analytic activity of quaternary, Pd-Ru-Rh-Re (3.5:4:1:1), deposit Pt electrode exceeded that of Pt electrode by ca. twice both in alkaline solution and artificial sea water. The paper conclusively proposes RMFP generated by nuclear fission to utilize as an alternative material for hydrogen production with a novel vision to bridge nuclear and hydrogen energy systems. (author)

  15. Boreside rotating ultrasonic tester for wastage determination of LMFBR-type steam generator tubes

    International Nuclear Information System (INIS)

    Neely, H.H.; Renger, H.L.

    1979-01-01

    Large sodium-water reaction (SWR) leak tests are being run in near-prototypic steam generators at prototypic plant conditions of the Liquid Metal Fast Breeder Reactor (LMFBR). These tests simulate various types of steam tube failure at predetermined locations. A SWR results in a highly energetic-exothermic-caustic reaction which erodes neighboring tubes. A boreside-rotating ultrasonic inspection device was developed to measure wall thickness and inside diameter of the 2/one quarter/Cr-1 Mo, 10.1 mm I.D. steam tubes. Rotation of the UT beam yields a complimentary scan of the full tube in a single pass. The UT system was designed with a 15 MHz transducer in pulse-echo compression-wave mode at a pulse rate of 10,000/second. The UT beam is rotated at 20 r/s on a 1.27 mm pitch. System outputs are diameter, wall thickness, attitude, and axial position. Measurements are processed, then fed to a CRT and computer for later retrieval and plotting

  16. Evaluation of the structural integrity of LMFBR equipment cell liners: results of preliminary investigations

    International Nuclear Information System (INIS)

    McAfee, W.J.; Sartory, W.K.

    1976-01-01

    The behavior of a plane wall segment of a prototype liquid-metal-cooled fast breeder reactor (LMFBR) cell under conditions of a postulated massive sodium spill was studied. Sodium-concrete reaction calculations were performed assuming an initial flaw existed in the liner such that high-temperature sodium could penetrate to the concrete underneath. Based on existing sodium-concrete reaction rate data, bounding values were established for the maximum energy release per unit volume of concrete. The potential effect of this energy release on the deformation of the liner material was determined. The temperature buildup in the liner and the pressure differential across the flaw in the liner were also studied. The transient thermal and structural responses of the steel liner and backup concrete were analyzed in detail using the inelastic computer code ANSYS. The literature on the mechanical, physical, and general behavior properties of concrete at high temperature was reviewed. This review emphasized the structural behavior of concrete and did not cover the sodium-concrete reaction

  17. An appreciation of the events, models and data used for LMFBR radiological source term estimations

    International Nuclear Information System (INIS)

    Keir, D.; Clough, P.N.

    1989-01-01

    In this report, the events, models and data currently available for analysis of accident source terms in liquid metal cooled fast neutron reactors are reviewed. The types of hypothetical accidents considered are the low probability, more extreme types of severe accident, involving significant degradation of the core and which may lead to the release of radionuclides. The base case reactor design considered is a commercial scale sodium pool reactor of the CDFR type. The feasibility of an integrated calculational approach to radionuclide transport and speciation (such as is used for LWR accident analysis) is explored. It is concluded that there is no fundamental obstacle, in terms of scientific data or understanding of the phenomena involved, to such an approach. However this must be regarded as a long-term goal because of the large amount of effort still required to advance development to a stage comparable with LWR studies. Particular aspects of LMFBR severe accident phenomenology which require attention are the behaviour of radionuclides during core disruptive accident bubble formation and evolution, and during the less rapid sequences of core melt under sodium. The basic requirement for improved thermal hydraulic modelling of core, coolant and structural materials, in these and other scenarios, is highlighted as fundamental to the accuracy and realism of source term estimations. The coupling of such modelling to that of radionuclide behaviour is seen as the key to future development in this area

  18. Nanoparticle Precipitation in Irradiated and Annealed Ceria Doped with Metals for Emulation of Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Weilin [Pacific Northwest National Laboratory, P.O. Box; Conroy, Michele A. [Pacific Northwest National Laboratory, P.O. Box; Kruska, Karen [Pacific Northwest National Laboratory, P.O. Box; Overman, Nicole R. [Pacific Northwest National Laboratory, P.O. Box; Droubay, Timothy C. [Pacific Northwest National Laboratory, P.O. Box; Gigax, Jonathan [Texas A& amp,M University, 3380; Shao, Lin [Texas A& amp,M University, 3380; Devanathan, Ram [Pacific Northwest National Laboratory, P.O. Box

    2017-09-29

    Epsilon alloy precipitates have been observed with varied compositions and sizes in spent nuclear fuels, such as UO2. Presence of the inclusions, along with other oxide precipitates, gas bubbles and irradiation-induced structural defects, can significantly degrade the physical properties of the fuel. To predict fuel performance, a fundamental study of the precipitation processes is needed. This study uses ceria (CeO2) as a surrogate for UO2. Polycrystalline CeO2 films doped with Mo, Ru, Rh, Pd and Re (surrogate for Tc) were grown at 823 K using pulsed laser deposition, irradiated at 673 K with He+ ions, and subsequently annealed at higher temperatures. A number of methods, including transmission electron microscopy and atom probe tomography, were applied to characterize the samples. The results indicate that there is a uniform distribution of the doped metals in the as-grown CeO2 film. Pd particles of ~3 nm in size appear near dislocation edges after He+ ion irradiation to ~13 dpa. Thermal annealing at 1073 K in air leads to formation of precipitates with Mo and Pd around grain boundaries. Further annealing at 1373 K produces 70 nm sized precipitates with small grains at cavities.

  19. Electrochemical investigation of mixed metal oxide nanocomposite electrode for low temperature solid oxide fuel cell

    Science.gov (United States)

    Abbas, Ghazanfar; Raza, Rizwan; Ashfaq Ahmad, M.; Ajmal Khan, M.; Jafar Hussain, M.; Ahmad, Mukhtar; Aziz, Hammad; Ahmad, Imran; Batool, Rida; Altaf, Faizah; Zhu, Bin

    2017-10-01

    Zinc-based nanostructured nickel (Ni) free metal oxide electrode material Zn0.60/Cu0.20Mn0.20 oxide (CMZO) was synthesized by solid state reaction and investigated for low temperature solid oxide fuel cell (LTSOFC) applications. The crystal structure and surface morphology of the synthesized electrode material were examined by XRD and SEM techniques respectively. The particle size of ZnO phase estimated by Scherer’s equation was 31.50 nm. The maximum electrical conductivity was found to be 12.567 S/cm and 5.846 S/cm in hydrogen and air atmosphere, respectively at 600∘C. The activation energy of the CMZO material was also calculated from the DC conductivity data using Arrhenius plots and it was found to be 0.060 and 0.075 eV in hydrogen and air atmosphere, respectively. The CMZO electrode-based fuel cell was tested using carbonated samarium doped ceria composite (NSDC) electrolyte. The three layers 13 mm in diameter and 1 mm thickness of the symmetric fuel cell were fabricated by dry pressing. The maximum power density of 728.86 mW/cm2 was measured at 550∘C.

  20. Preliminary Compatibility Assessment of Metallic Dispenser Materials for Service in Ethanol Fuel Blends

    Energy Technology Data Exchange (ETDEWEB)

    Pawel, Steven J [ORNL; Kass, Michael D [ORNL; Janke, Christopher James [ORNL

    2009-11-01

    The compatibility of selected metals representative of those commonly used in dispensing systems was evaluated in an aggressive E20 formulation (CE20a) and in synthetic gasoline (Reference Fuel C) in identical testing to facilitate comparison of results. The testing was performed at modestly elevated temperature (nominally 60 C) and with constant fluid flow in an effort to accelerate potential interactions in the screening test. Based on weight change, the general corrosion of all individual coupons exposed in the vapor phase above Reference Fuel C and CE20a as well as all coupons immersed in Reference Fuel C was essentially nil (<0.3 {micro}m/y), with no evidence of localized corrosion such as pitting/crevice corrosion or selective leaching at any location. Modest discoloration was observed on the copper-based alloys (cartridge brass and phosphor bronze), but the associated corrosion films were quite thin and apparently protective. For coupons immersed in CE20a, four different materials exhibited net weight loss over the entire course of the experiment: cartridge brass, phosphor bronze, galvanized steel, and terne-plated steel. None of these exhibited substantial incompatibility with the test fluid, with the largest general corrosion rate calculated from coupon weight loss to be approximately 4 {micro}m/y for the cartridge brass specimens. Selective leaching of zinc (from brass) and tin (from bronze) was observed, as well as the presence of sulfide surface films rich in these elements, suggesting the importance of the role of sulfuric acid in the CE20a formulation. Analysis of weight loss data for the slightly corroded metals indicated that the corrosivity of the test environment decreased with exposure time for brass and bronze and increased for galvanized and terne-plated steel. Other materials immersed in CE20a - type 1020 mild steel, type 1100 aluminum, type 201 nickel, and type 304 stainless steel - each appeared essentially immune to corrosion at the test

  1. Tradeoff of sodium void worth and burnup reactivity swing: Impacts on balance safety position in metallic-fueled cores

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

    1994-01-01

    A study has been conducted to investigate the effect of a lower sodium void worth on the consequences of severe accidents in metallic-fueled sodium-cooled reactors. Four 900 MWth designs were used for the study, where all of the reactor cores were designed based on the metallic fuel of the Integral Fast Reactor (IFR) concept. The four core designs each have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation was to determine the differences in severe accident response for the four core designs, in order to estimate the improvement in overall safety that could be achieved from a reduction in the sodium void worth for reactor cores which use a metallic fuel form

  2. Fabrication details for wire wrapped fuel assembly components

    International Nuclear Information System (INIS)

    Bosy, B.J.

    1978-09-01

    Extensive hydraulic testing of simulated LMFBR blanket and fuel assemblies is being carried out under this MIT program. The fabrication of these test assemblies has involved development of manufacturing procedures involving the wire wrapped pins and the flow housing. The procedures are described in detail in the report

  3. Use of fuel failure correlations in accident analysis

    International Nuclear Information System (INIS)

    O'Dell, L.D.; Baars, R.E.; Waltar, A.E.

    1975-05-01

    The MELT-III code for analysis of a Transient Overpower (TOP) accident in an LMFBR is briefly described, including failure criteria currently applied in the code. Preliminary results of calculations exploring failure patterns in time and space in the reactor core are reported and compared for the two empirical fuel failure correlations employed in the code. (U.S.)

  4. 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

  5. Metal cask RT-5000 for the dry storage and transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vorobyov, A.I.; Kazeev, V.G.; Krayev, V.S.; Shcherbina, A.N.; Churikov, Y.I. [All-Russian Research Inst. of Technical Physics, Snezhinsk (Russian Federation)

    2003-05-01

    Presentation of new-type cask, developed at RFNC-VNIITF, is in the article. The prototype model of the shipping cask was subjected to tests imitating normal shipment conditions (free fall, pressing, and impact) and to tests imitating emergency situation during shipment (a drop from the 9-m height onto a pin is replaced by acceleration of the shipping cask at a guide rail of the rocket-catapult installation (RCI), a 1-m drop onto a pin, heat tests a 30-minutes fire at the temperature of for 8500 C, submergence to the depth of 15 and 200 meters). After each test the hermeticity preservation is examined. Parallel with the real testing, a mathematical simulation of physical processes induced by the corresponding tests was conducted at the RFNC-VNIITF. The required parameters obtained from the tests are used to calibrate the calculation methods. As a result it has been possible to obtain a good agreement between the results of calculations and experiments; this will allow the mathematic simulation to be used wider. The advantage of the RT-5000 metal cask in comparison with metal-concrete analogs are as follows: SFA are placed into the RT-5000 entirely without cutting into two bunches of fuel elements; the expensive hot doom is not required for automatic cutting the SFA and for loading the bunches of fuel elements into intermediate cases; the possibility remains to transport the RT-5000 without reloading SFA after 50-year storage, although this is a problem for the metal-concrete casks.

  6. Gravitational agglomeration of post-HCDA LMFBR aerosols: nonspherical particles

    International Nuclear Information System (INIS)

    Tuttle, R.F.; Loyalka, S.K.

    1982-12-01

    Aerosol behavior analysis computer programs have shown that temporal aerosol size distributions in nuclear reactor containments are sensitive to shape factors. This research investigates shape factors by a detailed theoretical analysis of hydrodynamic interactions between a nonspherical particle and a spherical particle undergoing gravitational collisions in an LMFBR environment. First, basic definitions and expressions for settling speeds and collisional efficiencies of nonspherical particles are developed. These are then related to corresponding quantities for spherical particles through shape factors. Using volume equivalent diameter as the defining length in the gravitational collision kernel, the aerodynamic shape factor, the density correction factor, and the gravitational collision shape factor, are introduced to describe the collision kernel for collisions between aerosol agglomerates. The Navier-Stokes equation in oblate spheroidal coordinates is solved to model a nonspherical particle and then the dynamic equations for two particle motions are developed. A computer program (NGCEFF) is constructed, and the dynamical equations are solved by Gear's method

  7. Buoyancy effects on turbulent mixing in the LMFBR outlet plenum

    International Nuclear Information System (INIS)

    Chang, S.H.

    1983-01-01

    The effect of flow stratification is of particular concern during transient after scram in the outlet plenum of LMFBR. In this case, buoyancy effects on turbulent mixing are the importance to designers. An investigation has been made to identify the appropriate change in the available turbulence models which are necessary to include the effects of buoyancy on turbulence transport equations. The developed physical model of the buoyant turbulent flow are solved through SMAC method. Testing of the developed numerical model was undertaken and compared with experimental results. The results show that the buoyant turbulent effects account for the significant increase in the stability of the stratification, with a strong suppression of turbulence in the outlet plenum. (Author)

  8. Computational technique for estimating corrosion product release in an LMFBR

    International Nuclear Information System (INIS)

    Kuhn, W.L.

    1976-01-01

    A model is developed to predict the release of activated corrosion products in an LMFBR. Model parameters are obtained from literature correlations and from fitting the model to release rates observed in experimental sodium loops. The model is simplified to consider only two cases: release controlled by corrosion only or controlled by simultaneous corrosion and diffusion only. The release rates, radionuclide inventories, and resulting radiation fields near primary system components are predicted for the Fast Flux Test Facility as a function of operating temperature, oxygen concentration in the sodium, and time of operation. Saturation inventories are reported for 54 Mn, 60 Co, 58 Co, 51 Cr, and 59 Fe. The resulting radiation fields are due mainly to 54 Mn, and are found to reach the 5 R/hr range for operation at a core outlet temperature of 593 0 C

  9. Neutron and gamma transport effects by heterogeneous core designs. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Lam, S.K.

    1977-01-01

    The use of diffusion theory for the prediction of power production near a reactor core-blanket interface and the assumption that gammas are absorbed in situ can lead to substantial errors. This is primarily due to the breakdown of Fick's law for neutron diffusion near the core-blanket boundary, and the gamma leakage from the core into the blanket. These considerations are more pronounced in a situation where a large number of internal blanket assemblies are present, such as in the large heterogeneous core designs. The power distribution is studied for both fission and gamma heating in a large heterogeneous LMFBR with 3 core zones separated by 2 internal blanket zones. Comparisons are made between diffusion and transport theory for neutronics calculations, and between in-situ absorption and rigorous transport theory calculation for gamma heating.

  10. A technique for computing bowing reactivity feedback in LMFBR's

    International Nuclear Information System (INIS)

    Finck, P.J.

    1987-01-01

    During normal or accidental transients occurring in a LMFBR core, the assemblies and their support structure are subjected to important thermal gradients which induce differential thermal expansions of the walls of the hexcans and differential displacement of the assembly support structure. These displacements, combined with the creep and swelling of structural materials, remain quite small, but the resulting reactivity changes constitute a significant component of the reactivity feedback coefficients used in safety analyses. It would be prohibitive to compute the reactivity changes due to all transients. Thus, the usual practice is to generate reactivity gradient tables. The purpose of the work presented here is twofold: develop and validate an efficient and accurate scheme for computing these reactivity tables; and to qualify this scheme

  11. Research and development of bellows for LMFBR in Japan

    International Nuclear Information System (INIS)

    Takahashi, Tadao; Mukai, Kazuo; Yamamoto, Ken.

    1979-11-01

    The bellows is employed as a useful mechanical element with its flexibility and imperviousness to liquid and gas in the system in which such chemically active substance as sodium is handled. Since the early time of development of Japanese LMFBR, bellows have been used e.g. for the shaft seal of small sodium valves. Bellows are employed in the fast experimental reactor JOYO which is now in operation and the fast prototype reactor MONJU whose design program is in final stage at the following parts: - control rod drive mechanism, - intermediate heat exchanger, - small valve, - mechanical penetration assembly of the containment boundary, - outer piping of the double-walled primary system (for JOYO only). In addition, the application of bellows as thermal expansion joint to the main piping system is under consideration for future FBRs. This paper outlines the research and development work on bellows for the FBR use in Japan. (author)

  12. Research and development of bellows for LMFBR in Japan

    International Nuclear Information System (INIS)

    Takahashi, T.; Mukai, K.; Yamamoto, K.

    1980-01-01

    Bellows are employed as useful mechanical elements with their flexibility and imperviousness to liquid and gas in the system in which such chemically active substances as sodium are handled. Since the early time of development of Japanese LMFBR, bellows have been used e.g. for the shaft seal of small sodium valves. Bellows are employed in the fast experimental reactor JOYO which is now in operation and the fast prototype reactor MONJU whose design program is in final stage at the following parts: control rod drive mechanism; intermediate heat exchanger; small valve; mechanical penetration assembly of the containment boundary; outer piping of the double-walled primary system (for JOYO only). In addition, the application of bellows as thermal expansion joint to the main piping system is under consideration for future FBRs. This paper outlines the research and development work on bellows for FBR use in Japan. (author)

  13. Fatigue of LMFBR piping due to flow stratification

    International Nuclear Information System (INIS)

    Woodward, W.S.

    1983-01-01

    Flow stratification due to reverse flow was simulated in a 1/5-scale water model of a LMFBR primary pipe loop. The stratified flow was observed to have a dynamic interface region which oscillated in a wave pattern. The behavior of the interface was characterized in terms of location, local temperature fluctuation and duration for various reverse flow conditions. A structural assessment was performed to determine the effects of stratified flow on the fatigue life of the pipe. Both the static and dynamic aspects of flow stratification were examined. The dynamic interface produces thermal striping on the inside of the pipe wall which is shown to have the most deleterious effect on the pipe wall and produce significant fatigue damage relative to a static interface

  14. MIT LMFBR blanket research project. Final summary report

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J.

    1983-08-01

    This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record.

  15. Numerical evaluation of oxide growth in metallic support microstructures of Solid Oxide Fuel Cells and its influence on mass transport

    DEFF Research Database (Denmark)

    Reiss, Georg; Frandsen, Henrik Lund; Persson, Åsa Helen

    2015-01-01

    Metal-supported Solid Oxide Fuel Cells (SOFCs) are developed as a durable and cost-effective alternative to the state-of-the-art cermet SOFCs. This novel technology offers new opportunities but also new challenges. One of them is corrosion of the metallic support, which will decrease the long......-temperature corrosion theory, and the required model parameters can be retrieved by standard corrosion weight gain measurements. The microstructure is reconstructed from X-ray computed tomography, and converted into a computational grid. The influence of the changing microstructure on the fuel cell performance...

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

    Science.gov (United States)

    Johnson, Irving

    1988-06-01

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

  17. Specialists meeting on LMFBR flow induced vibrations. Summary report

    International Nuclear Information System (INIS)

    1977-12-01

    A Specialists' Meeting on LMFBR Flow-Induced Vibrations was held at ANL in the United States which was sponsored by the International Atomic Energy Agency (IAEA) on the recommendations of the International Working Group on Fast Reactors (IWGFR). It was attended by participants from France, the Federal Republic of Germany, Italy, Japan, Netherlands, the United Kingdom, the Union of Soviet Socialist Republics, the United States and the IAEA. The purpose of the meeting was to provide, for the first time, a common forum for the exchange of information on flow-induced vibration programs of the member countries. As this was a first meeting, information was sought in the broad areas of: 1. Design Criteria and Problem Areas in LMFBR Design; 2. Current Design Procedures; and 3. Ongoing Research. A session was devoted to each of the above topics wherein papers were presented and discussed followed by open discussions on the session topic. The objective of the open discussions was to identify, from a review of specific reactor designs, (a) flow induced vibration problem areas (expected and observed) and their potential for occurrence; (b) failure modes and associated design criteria; (c) specific components that are susceptible to flow induced vibration; and (d) probable excitation mechanisms. It was aimed to assess the current state-of-the-art in designing to avoid flow induced vibration with consideration of licensing requirements; to evaluate existing methods of analysis, testing, and surveillance, along with their limitations and to identify areas requiring research and review ongoing research programmes relative to these research needs

  18. 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.

  19. 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

  20. Evaluation of an EMITEC resistively heated metal monolith catalytic converter on two M100 neat methanol-fueled vehicles

    Science.gov (United States)

    Piotrowski, Gregory K.; Schaefer, Ronald M.

    1992-12-01

    The report describes the evaluation of a resistively heated catalyst system on two different methanol fueled vehicles. The EMITEC catalyst consisted of a compact resistively heated metal monolith in front of a larger conventional main converter. The EMITEC catalyst was evaluated on two neat methanol-fueled vehicles, a 1981 Volkswagen Rabbit and a 1988 Toyota Corolla. Emission testing was conducted over the Federal Test Procedure (FTP) CVS-75 test cycle. The emissions of primary interest were cold start methanol (unburned fuel), carbon monoxide, and formaldehyde.

  1. CSER 94-014: Storage of metal-fuel loaded EBR-II casks in concrete vault on PFP grounds

    International Nuclear Information System (INIS)

    Hess, A.L.

    1994-01-01

    A criticality safety evaluation is presented to permit EBR-2 spent fuel casks loaded with metallic fuel rods to be stored in an 8-ft diameter, cylindrical concrete vault inside the PFP security perimeter. The specific transfer of three casks with Pu alloy fuel from the Los Alamos Molten Plutonium Reactor Experiment from the burial grounds to the vault is thus covered. Up to seven casks may be emplaced in the casing with 30 inches center to center spacing. Criticality safety is assured by definitive packaging rules which keep the fissile medium dry and at a low effective volumetric density

  2. The possibility of prediction of the lifetime of metallic nuclear fuel elements in a radiation field of thermal nuclear reactors

    International Nuclear Information System (INIS)

    Livne, Z.; Ramon, P.

    1979-01-01

    An attempt is made to clarify the possible causes of failure of irradiated nuclear fuel cartridges, in order to determine the parameters which govern the lifetime of the fuel and a way to predict it. Measurements of mechanical properties of irradiated uranium metal and cladding, can serve as a basis for failure prediction. Testing irradiated fuel elements by bending till fracture enables to evaluate the integral character of the fuel element, along the cross-section, taking into account the difference in brittleness of several zones. It is likely that the bending test, which indicates the behaviour of a stress-strain function, is a faster and more reliable way to determine the mechanical properties of irradiated nuclear fuel. Since the stresses applied to the cladding during irradiation are locally hydrostatic, its postirradiation blow-up provide information on strength and elasticity variations of the irradiated cladding material. (B.G.)

  3. Interaction of coal-derived synthesis gas impurities with solid oxide fuel cell metallic components

    Energy Technology Data Exchange (ETDEWEB)

    Marina, Olga A.; Coyle, Christopher A.; Edwards, Danny J.; Chou, Yeong-Shyung; Cramer, Carolyn N. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Pederson, Larry R. [North Dakota State University, Fargo, ND 58102 (United States)

    2011-01-15

    Oxidation-resistant alloys find use as interconnect materials, heat exchangers, and gas supply tubing in solid oxide fuel cell (SOFC) systems, especially when operated at temperatures below {proportional_to}800 C. If fueled with synthesis gas derived from coal or biomass, such metallic components could be exposed to impurities contained in those fuel sources. In this study, coupons of ferritic stainless steels Crofer 22 APU and SS 441, austenitic nickel-chromium superalloy Inconel 600, and an alumina-forming high nickel alloy alumel were exposed to synthesis gas containing {<=}2 ppm phosphorus, arsenic and antimony, and reaction products were tested. Crofer 22 APU coupons coated with a (Mn,Co){sub 3}O{sub 4} protective layer were also evaluated. Phosphorus was found to be the most reactive. On Crofer 22 APU, the (Mn,Cr){sub 3}O{sub 4} passivation layer reacted to form an Mn-P-O product, predicted to be manganese phosphate from thermochemical calculations, and Cr{sub 2}O{sub 3}. On SS 441, reaction of phosphorus with (Mn,Cr){sub 3}O{sub 4} led to the formation of manganese phosphate as well as an Fe-P product, predicted from thermochemical calculations to be Fe{sub 3}P. Minimal interactions with antimony or arsenic in synthesis gas were limited to Fe-Sb and Fe-As solid solution formation. Though not intended for use on the anode side, a (Mn,Co){sub 3}O{sub 4} spinel coating on Crofer 22 APU reacted with phosphorus in synthesis gas to produce products consistent with Mn{sub 3}(PO{sub 4}){sub 2} and Co{sub 2}P. A thin Cr{sub 2}O{sub 3} passivation layer on Inconel 600 did not prevent the formation of nickel phosphides and arsenides and of iron phosphides and arsenides, though no reaction with Cr{sub 2}O{sub 3} was apparent. On alumel, an Al{sub 2}O{sub 3} passivation layer rich in Ni did not prevent the formation of nickel phosphides, arsenides, and antimonides, though no reaction with Al{sub 2}O{sub 3} occurred. This work shows that unprotected metallic components of

  4. Influence of carbonation under oxy-fuel combustion flue gas on the leachability of heavy metals in MSWI fly ash.

    Science.gov (United States)

    Ni, Peng; Xiong, Zhuo; Tian, Chong; Li, Hailong; Zhao, Yongchun; Zhang, Junying; Zheng, Chuguang

    2017-09-01

    Due to the high cost of pure CO 2 , carbonation of MSWI fly ash has not been fully developed. It is essential to select a kind of reaction gas with rich CO 2 instead of pure CO 2 . The CO 2 uptake and leaching toxicity of heavy metals in three typical types of municipal solid waste incinerator (MSWI) fly ash were investigated with simulated oxy-fuel combustion flue gas under different reaction temperatures, which was compared with both pure CO 2 and simulated air combustion flue gas. The CO 2 uptake under simulated oxy-fuel combustion flue gas were similar to that of pure CO 2 . The leaching concentration of heavy metals in all MSWI fly ash samples, especially in ash from Changzhou, China (CZ), decreased after carbonation. Specifically, the leached Pb concentration of the CZ MSWI fly ash decreased 92% under oxy-fuel combustion flue gas, 95% under pure CO 2 atmosphere and 84% under the air combustion flue gas. After carbonation, the leaching concentration of Pb was below the Chinese legal limit. The leaching concentration of Zn from CZ sample decreased 69% under oxy-fuel combustion flue gas, which of Cu, As, Cr and Hg decreased 25%, 33%, 11% and 21%, respectively. In the other two samples of Xuzhou, China (XZ) and Wuhan, China (WH), the leaching characteristics of heavy metals were similar to the CZ sample. The speciation of heavy metals was largely changed from the exchangeable to carbonated fraction because of the carbonation reaction under simulated oxy-fuel combustion flue gas. After carbonation reaction, most of heavy metals bound in carbonates became more stable and leached less. Therefore, oxy-fuel combustion flue gas could be a low-cost source for carbonation of MSWI fly ash. Copyright © 2017 Elsevier Ltd. All rights reserved.

  5. PBDOWN - a computer code for simulating core material discharge and thermal to mechanical energy conversion in LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Royl, P.

    1981-01-01

    PBDOWN is a computer code that simulates the blowdown of confined boiling materials ('pools') into a colder upper coolant plenum as time dependent ejection and expansion with consideration of a few selected exchange processes. Its application is restricted to situations resulting from hypothetical loss of flow (LOF) accidents in LMFBR's, where enough voiding has occured, that in core sodium vapor pressures become negligible. PBDOWN considers one working fluid for the discharge process (either fuel or steel) and a maximum of two working fluids (either fuel and sodium or steel and sodium) for the expansion process in the upper coolant plenum. Entrainment of sodium at the accelerated bubble liquid interfaces is mechanistically calculated by a Taylor instability entrainment model. Simulation of a hemispherical expansion form together with this mechanistic entrainment model gives a new integrated calculation of the time dependent sodium mass in the bubble. The paper summarizes the basic equations and assumptions of this computer model. Sample results compare different heat transfer and Na entrainment models during steel and fuel driven discharge processes. Mechanistic sodium entrainment simulation for SNR-type reactors coupled with a realistic heat transfer model is shown to reduce the integral mechanical work potential by a factor of 1.3 to 2.0 over the isentropic energy of the discharge working fluids. (orig.)

  6. High temperature corrosion of metallic interconnects in solid oxide fuel cells

    Directory of Open Access Journals (Sweden)

    Bastidas, D. M.

    2006-12-01

    Full Text Available Research and development has made it possible to use metallic interconnects in solid oxide fuel cells (SOFC instead of ceramic materials. The use of metallic interconnects was formerly hindered by the high operating temperature, which made the interconnect degrade too much and too fast to be an efficient alternative. When the operating temperature was lowered, the use of metallic interconnects proved to be favourable since they are easier and cheaper to produce than ceramic interconnects. However, metallic interconnects continue to be degraded despite the lowered temperature, and their corrosion products contribute to electrical degradation in the fuel cell. Coatings of nickel, chromium, aluminium, zinc, manganese, yttrium or lanthanum between the interconnect and the electrodes reduce this degradation during operation

    El uso de interconectores metálicos en pilas de combustible de óxido sólido (SOFC en sustitución de materiales cerámicos ha sido posible gracias a la investigación y desarrollo de nuevos materiales metálicos. Inicialmente, el uso de interconectores metálicos fue limitado, debido a la elevada temperatura de trabajo, ocasionando de forma rápida la degradación del material, lo que impedía que fuesen una alternativa. A medida que la temperatura de trabajo de las SOFC descendió, el uso de interconectores metálicos demostró ser una buena alternativa, dado que son más fáciles de fabricar y más baratos que los interconectores cerámicos. Sin embargo, los interconectores metálicos continúan degradándose a pesar de descender la temperatura a la que operan las SOFC y, asimismo, los productos de corrosión favorecen las pérdidas eléctricas de la pila de combustible. Recubrimientos de níquel, cromo, aluminio, zinc, manganeso, itrio y lantano entre el interconector y los electrodos reduce dichas pérdidas eléctricas.

  7. Synthesis and characterization of metallic nuclear fuels; Sintese e caracterizacao de combustiveis nucleares metalicos

    Energy Technology Data Exchange (ETDEWEB)

    Longen, F.R., E-mail: frlongen@utfpr.edu.br [Universidade Tecnologica Federal do Parana (UTFPR), Medianeira, PR (Brazil); Barco, R.; Paesano Junior, A. [Universidade Estadual de Maringa (UEM), PR (Brazil); Pagano Junior, L. [Centro Tecnologico da Marinha (CETEM), Sao Paulo, SP (Brazil)

    2014-07-01

    U-Zr-Mo and U-Zr-Gd ternary alloys, potentially useful as metallic nuclear fuel, were prepared at different concentrations by arc-melting and characterized by X-ray diffraction. Those alloys containing molybdenum were submitted to thermal annealing in inert atmosphere, followed by quenching in water. These samples were measured before and after the thermal treatment. The diffractometric results evidenced that the as-cast alloys solidified mostly with a body centered cubic structure (γphase) and that for the uranium richest samples a second phase formed, with an orthorhombic structure (α phase). For the U-Zr-Gd alloys the X-ray diffractometry revealed the retention of a hexagonal structure (δ phase) and gadolinium traces in the poorest uranium samples. The U{sub 57}(Zr{sub 92}Gd{sub 8}){sub 43} sample resulted monophasic becoming, according to literature, the first time that a solid solution combining uranium and gadolinium is identified. (author)

  8. A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Jerden, James

    2017-06-26

    A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gaps in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.

  9. Power generation using an activated carbon and metal mesh cathode in a microbial fuel cell

    KAUST Repository

    Zhang, Fang

    2009-11-01

    An inexpensive activated carbon (AC) air cathode was developed as an alternative to a platinum-catalyzed electrode for oxygen reduction in a microbial fuel cell (MFC). AC was cold-pressed with a polytetrafluoroethylene (PTFE) binder to form the cathode around a Ni mesh current collector. This cathode construction avoided the need for carbon cloth or a metal catalyst, and produced a cathode with high activity for oxygen reduction at typical MFC current densities. Tests with the AC cathode produced a maximum power density of 1220 mW/m2 (normalized to cathode projected surface area; 36 W/m3 based on liquid volume) compared to 1060 mW/m2 obtained by Pt catalyzed carbon cloth cathode. The Coulombic efficiency ranged from 15% to 55%. These findings show that AC is a cost-effective material for achieving useful rates of oxygen reduction in air cathode MFCs. © 2009 Elsevier B.V. All rights reserved.

  10. Fossil fuel consumption and heavy metal emissions into the atmosphere in Russia

    International Nuclear Information System (INIS)

    Ginzburg, V.; Gromov, S.

    1999-01-01

    In recent decades more and more attention has been paid to the problem of ecosystem pollution by heavy metals. Many trace elements are registered now as a global pollutant due to their toxic nature. Their negative influence on the environment is caused by accumulation in different ecosystem components and increased involvement in biochemical cycles. The atmosphere is the main medium through which pollutants transported from emission sources to background territories where heavy metals are deposited into water and on plants. Heavy metal emissions into the atmosphere cause certain global environmental problems due to their long lifetime and the long-term transport of these elements in the atmosphere, as well as the increasing rate of their accumulation in the environment even at most remote territories. Moreover, heavy metals have evidently entered human food chains. The influence of global ecosystem pollution by heavy metals on human health is not well known as yet. Most trace elements comes into the atmosphere with natural and man-made aerosols. The main sources of natural aerosols in the atmosphere are soil erosion and weathering of mountain rocks, volcanic and space dust, forest firing smoke, and others. Major anthropogenic sources of toxic elements are fossil fuel combustion, mining, industrial processes, and waste incineration. The anthropogenic flow of heavy metals to the atmosphere is about 94-97 per cent of the total. An inventory of emission sources should be the first step in developing a control strategy and modelling global and regional cycles of trace elements. In this article the situation with lead, cadmium and mercury emissions from coal combustion of power plants and gasoline combustion by road transport is discussed. Pollutant amounts released into the atmosphere in industrial regions induce not only local deterioration of air, but they also affect on remote areas, and areas sensitive to contamination, such as the Arctic region. Problems on the

  11. A review of ANL base technology studies in support of the U.S. LMFBR vibration program

    International Nuclear Information System (INIS)

    Wambsganss, M.W.; Chen, S.S.; Mulcahy, T.M.; Shin, Y.S.

    1977-01-01

    Argonne National Laboratory (ANL) is the center for base technology studies of flow induced vibration for the U.S. LMFBR Program. This paper reviews and summarizes published results, reports on the status of ongoing programs, and discusses future needs as outlined in the U.S. LMFBR Vibrations Program Plan. (author)

  12. Advanced methods for the fabrication of mixed uranium plutonium oxide, monocarbide and mononitride fuels for fast breeder reactors

    International Nuclear Information System (INIS)

    Ganguly, C.

    1988-01-01

    Mixed uranium plutonium monocarbide (MC) and mononitride (MN) are considered as advanced LMFBR fuels. 'Powder-pellet (POP) is the conventional method for production of MOX, MC and MN fuel pellets starting from UO 2 and PuO 2 powders. The POP route involves generation and handling of plutonium bearing fine powder or dust particles which has the problem of radiotoxic dust hazard. Further, fine powders have poor flowability which makes automation and remote fabrication difficult. The combination of ammonium uranyl plutonyl carbonate (AUPuC) process and low temperature oxidative sintering (LTS) minimises the radiotoxic dust hazard and the fabrication cost of MOX fuel and is considered as advanced POP route. Vibro-sol (or sphere-pac) and sol-gel microsphere pelletisation (SGMP) are advanced methods for fabrication of MOX, MC and MN fuels. Here, dust-free and free-flowing gel microspheres of the oxide or oxide-carbon are prepared from heavy metal nitrate feed solutions by 'internal' or 'external' gelation processes. The gel microspheres are subjected to controlled calcination for MOX or carbothermic reduction for MC and MN. Thereafter, the microspheres are either vibropacked in fuel pins or directly pelletised and sintered. These processes minimise radiotoxic dust hazard, facilitate automation and remote processing and ensure excellent microhomogeneity. The present paper summarises the state-of-art of the POP, vibrosol and SGMP processes for the fabrication of MOX, MC and MN fuels, highlighting the author's experience in SGMP process. (author). 21 refs., 11 figs

  13. A Mechanistic Reliability Assessment of RVACS and Metal Fuel Inherent Reactivity Feedbacks

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Brunett, Acacia J.; Passerini, Stefano; Grelle, Austin

    2017-09-24

    GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory (Argonne) participated in a two year collaboration to modernize and update the probabilistic risk assessment (PRA) for the PRISM sodium fast reactor. At a high level, the primary outcome of the project was the development of a next-generation PRA that is intended to enable risk-informed prioritization of safety- and reliability-focused research and development. A central Argonne task during this project was a reliability assessment of passive safety systems, which included the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedbacks of the metal fuel core. Both systems were examined utilizing a methodology derived from the Reliability Method for Passive Safety Functions (RMPS), with an emphasis on developing success criteria based on mechanistic system modeling while also maintaining consistency with the Fuel Damage Categories (FDCs) of the mechanistic source term assessment. This paper provides an overview of the reliability analyses of both systems, including highlights of the FMEAs, the construction of best-estimate models, uncertain parameter screening and propagation, and the quantification of system failure probability. In particular, special focus is given to the methodologies to perform the analysis of uncertainty propagation and the determination of the likelihood of violating FDC limits. Additionally, important lessons learned are also reviewed, such as optimal sampling methodologies for the discovery of low likelihood failure events and strategies for the combined treatment of aleatory and epistemic uncertainties.

  14. Validation of KENO, ANISN and Hansen-Roach cross-section set on plutonium oxide and metal fuel system

    International Nuclear Information System (INIS)

    Matsumoto, Tadakuni; Yumoto, Ryozo; Nakano, Koh.

    1980-01-01

    In the previous report, the authors discussed the validity of KENO, ANISN and Hansen-Roach 16 group cross-section set on the critical plutonium nitrate solution systems with various geometries, absorbers and neutron interactions. The purpose of the present report is to examine the validity of the same calculation systems on the homogeneous plutonium oxide and plutonium-uranium mixed oxide fuels with various density values. Eleven experiments adopted for validation are summarized. First six experiments were performed at Pacific Northwest Laboratory of Battelle Memorial Institute, and the remaining five at Los Alamos Scientific Laboratory. The characteristics of core fuel are given, and the isotopic composition of plutonium, the relation between H/(Pu + U) atomic ratio and fuel density as compared with the atomic ratios of PuO 2 and mixed oxides in powder storage and pellet fabrication processes, and critical core dimensions and reflector conditions are shown. The effective multiplication factors were calculated with the KENO code. In case of the metal fuels with simple sphere geometry, additional calculations with the ANISN code were performed. The criticality calculation system composed of KENO, ANISN and Hansen-Roach cross-section set was found to be valid for calculating the criticality on plutonium oxide, plutonium-uranium mixed oxide, plutonium metal and uranium metal fuel systems as well as on plutonium solution systems with various geometries, absorbers and neutron interactions. There seems to remain some problems in the method for evaluating experimental correction. Some discussions foloow. (Wakatsuki, Y.)

  15. Performance Factors and Sulfur Tolerance of Metal Supported Solid Oxide Fuel Cells with Nanostructured Ni:GDC Infiltrated Anodes

    DEFF Research Database (Denmark)

    Nielsen, Jimmi; Sudireddy, Bhaskar Reddy; Hagen, Anke

    2016-01-01

    Two metal supported solid oxide fuel cells (active area 16 cm2) with nanostructured Ni:GDC infiltrated anodes, possessing different anode and support microstructures were studied in respect to sulfur tolerance at an operating temperature of 650°C. The studied MS-SOFCs are based on ferretic...

  16. Performance Factors and Sulfur Tolerance of Metal Supported Solid Oxide Fuel Cells with Nanostructured Ni:GDC Infiltrated Anodes

    DEFF Research Database (Denmark)

    Nielsen, Jimmi; Sudireddy, Bhaskar Reddy; Hagen, Anke

    2015-01-01

    Two metal supported solid oxide fuel cells (active area 16 cm2) with nanostructured Ni:GDC infiltrated anodes, but different anode and support microstructures were studied in respect to sulfur tolerance at the aimed operating temperature of 650ºC. The studied MS-SOFCs are based on ferretic...

  17. Metal organic chemical vapor deposition of environmental barrier coatings for the inhibition of solid deposit formation from heated jet fuel

    Science.gov (United States)

    Mohan, Arun Ram

    Solid deposit formation from jet fuel compromises the fuel handling system of an aviation turbine engine and increases the maintenance downtime of an aircraft. The deposit formation process depends upon the composition of the fuel, the nature of metal surfaces that come in contact with the heated fuel and the operating conditions of the engine. The objective of the study is to investigate the effect of substrate surfaces on the amount and nature of solid deposits in the intermediate regime where both autoxidation and pyrolysis play an important role in deposit formation. A particular focus has been directed to examining the effectiveness of barrier coatings produced by metal organic chemical vapor deposition (MOCVD) on metal surfaces for inhibiting the solid deposit formation from jet fuel degradation. In the first part of the experimental study, a commercial Jet-A sample was stressed in a flow reactor on seven different metal surfaces: AISI316, AISI 321, AISI 304, AISI 347, Inconel 600, Inconel 718, Inconel 750X and FecrAlloy. Examination of deposits by thermal and microscopic analysis shows that the solid deposit formation is influenced by the interaction of organosulfur compounds and autoxidation products with the metal surfaces. The nature of metal sulfides was predicted by Fe-Ni-S ternary phase diagram. Thermal stressing on uncoated surfaces produced coke deposits with varying degree of structural order. They are hydrogen-rich and structurally disordered deposits, spherulitic deposits, small carbon particles with relatively ordered structures and large platelets of ordered carbon structures formed by metal catalysis. In the second part of the study, environmental barrier coatings were deposited on tube surfaces to inhibit solid deposit formation from the heated fuel. A new CVD system was configured by the proper choice of components for mass flow, pressure and temperature control in the reactor. A bubbler was designed to deliver the precursor into the reactor

  18. Biogenic metallic nanoparticles as catalyst for bioelectricity production: A novel approach in microbial fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Saravanakumar, Kandasamy, E-mail: saravana732@gmail.com [School of Agriculture and Biology, Shanghai Jiao Tong University, Shanghai (China); State Key Laboratory of Microbial Metabolism, Shanghai Jiao Tong University, Shanghai (China); Key Laboratory of Urban Agriculture (South), Ministry of Agriculture, Shanghai (China); MubarakAli, Davoodbasha [Microbial Genetic Engineering Laboratory, Division of Bioengineering, College of Life Science and Bioengineering, Incheon National University, Songdo 406772, Incheon (Korea, Republic of); Department of Microbiology, School of Lifesciences, Bharathidasan University, Tiruchirappalli 620024 (India); Kathiresan, Kandasamy [Centre of Advanced Study in Marine Biology, Faculty of Marine Sciences, Annamalai University, Parangipettai 608 502, Tamil Nadu (India); Thajuddin, Nooruddin [Department of Microbiology, School of Lifesciences, Bharathidasan University, Tiruchirappalli 620024 (India); Department of Botany and Microbiology, College of Science, King Saud University, Riyadh 11451 (Saudi Arabia); Alharbi, Naiyf S. [Department of Botany and Microbiology, College of Science, King Saud University, Riyadh 11451 (Saudi Arabia); Chen, Jie, E-mail: jiechen59@sjtu.edu.cn [School of Agriculture and Biology, Shanghai Jiao Tong University, Shanghai (China); State Key Laboratory of Microbial Metabolism, Shanghai Jiao Tong University, Shanghai (China); Key Laboratory of Urban Agriculture (South), Ministry of Agriculture, Shanghai (China)

    2016-01-15

    Highlights: • Trichoderma sp., showed an abilities to synthesis of AgNPs and AuNPs with an excellent stability. • AuNPs significantly enhanced the bioelectricity production by MFC of anaerobic fermentation as catalyst. • Maximum bioelectricity production was optimized and obtained the voltage of 432.80 mA using RSM. - Abstract: The present work aimed to use the biogenic metallic nanoparticles as catalyst for bioelectricity production in microbial fuel cell (MFC) approach under anaerobic condition. Silver and gold nanoparticles (AuNPs) were synthesized using Trichoderma sp. Particle size and cystallinity were measured by X-ray diffraction revealed the crystalline structure with average size of 36.17 nm. Electron microscopic studies showed spherical shaped silver nanoparticles (AgNPs) and cubical shaped AuNPs with size ranges from 50 to 150 nm. The concentration of biogenic metallic nanoparticles as catalyst for enhanced bioelectricity generations and estimated by response surface methodology (RSM) and found at the greatest of 342.80 mA under optimized conditions are time interval, temperature, nanoparticles used as 63 h, 28 ± 2.0 °C, 22.54 mg l{sup −1} (AgNPs) and 25.62 mg l{sup −1} (AuNPs) in a batch reactor. AuNPs acted as an excellent catalyst to enhance the bioelectricity production. This novel technique could be used for eco-friendly, economically feasible and facile electricity production.

  19. Study on Doppler coefficient for metallic fuel fast reactor added hydrogeneous moderator

    Energy Technology Data Exchange (ETDEWEB)

    Hirakawa, Naohiro; Iwasaki, Tomohiko; Tsujimoto, Kazuhumi [Tohoku Univ., Sendai (Japan). Faculty of Engineering; Osugi, Toshitaka; Okajima, Shigeaki; Andoh, Masaki; Nemoto, Tatsuo; Mukaiyama, Takehiko

    1998-01-01

    A series of mock-up experiments for moderator added metallic fast reactor core was carried out at FCA to obtain the experimental verification for improvement of reactivity coefficients. Softened neutron spectrum increases Doppler effect by a factor of 2, and flatter adjoint neutron spectrum decreases Na void effect by a factor of 0.6 when hydrogen to heavy metal atomic number ratio is increased from 0.02 to 0.13. The experimental results are analyzed with SLALOM and CITATION-FBR, which is the standard design code system for a fast reactor at JAERI, and SRAC95 and CITATION-FBR. The present code system gives generally good agreement with the experimental results, especially by the use of the latter, the dependence of the Doppler effect to the hydrogen to fuel element atomic number density ratio is disappeared. Therefore, it looks possible to use the present code system for the conceptual design of a fast reactor system with hydrogeneous materials. (author)

  20. A novel Ni/ceria-based anode for metal-supported solid oxide fuel cells

    Science.gov (United States)

    Rojek-Wöckner, Veronika A.; Opitz, Alexander K.; Brandner, Marco; Mathé, Jörg; Bram, Martin

    2016-10-01

    For optimization of ageing behavior, electrochemical performance, and sulfur tolerance of metal-supported solid oxide fuel cells a new anode concept is introduced, which is based on a Ni/GDC cermet replacing the established Ni/YSZ anodes. In the present work optimized processing parameters compatible with MSC substrates are specified by doing sintering studies on pressed bulk specimen and on real porous anode structures. The electrochemical performance of the Ni/GDC anodes was characterized by means of symmetrical electrolyte supported model-type cells. In this study, three main objectives are pursued. Firstly, the effective technical realization of the Ni/GDC concept is demonstrated. Secondly, the electrochemical behavior of Ni/GDC porous anodes is characterized by impedance spectroscopy and compared with the current standard Ni/YSZ anode. Further, a qualitative comparison of the sulfur poisoning behavior of both anode types is presented. Thirdly, preliminary results of a successful implementation of the Ni/GDC cermet into a metal-supported single cell are presented.

  1. Influence of platinum group metal-free catalyst synthesis on microbial fuel cell performance

    Science.gov (United States)

    Santoro, Carlo; Rojas-Carbonell, Santiago; Awais, Roxanne; Gokhale, Rohan; Kodali, Mounika; Serov, Alexey; Artyushkova, Kateryna; Atanassov, Plamen

    2018-01-01

    Platinum group metal-free (PGM-free) ORR catalysts from the Fe-N-C family were synthesized using sacrificial support method (SSM) technique. Six experimental steps were used during the synthesis: 1) mixing the precursor, the metal salt, and the silica template; 2) first pyrolysis in hydrogen rich atmosphere; 3) ball milling; 4) etching the silica template using harsh acids environment; 5) the second pyrolysis in ammonia rich atmosphere; 6) final ball milling. Three independent batches were fabricated following the same procedure. The effect of each synthetic parameters on the surface chemistry and the electrocatalytic performance in neutral media was studied. Rotating ring disk electrode (RRDE) experiment showed an increase in half wave potential and limiting current after the pyrolysis steps. The additional improvement was observed after etching and performing the second pyrolysis. A similar trend was seen in microbial fuel cells (MFCs), in which the power output increased from 167 ± 2 μW cm-2 to 214 ± 5 μW cm-2. X-ray Photoelectron Spectroscopy (XPS) was used to evaluate surface chemistry of catalysts obtained after each synthetic step. The changes in chemical composition were directly correlated with the improvements in performance. We report outstanding reproducibility in both composition and performance among the three different batches.

  2. Impingement heat flux by dispersed molten metal fuel on a horizontal stainless steel structure

    International Nuclear Information System (INIS)

    Gabor, J.D.; Purviance, R.T.; Aeschlimann, R.W.; Spencer, B.W.

    1989-01-01

    Although the Integral Fast Reactor (IFR) possesses inherent safety features, an assessment of the consequences of melting of the metal fuel is necessary for risk analysis. As part of this effort an experimental study was conducted to determine the depths of sodium at 600 C required for pour streams of various molten uranium alloys (U, U-5 wt % Zr, U-10 wt % Zr, and U-10 wt % Fe) to break up and solidify. The quenched particulate material, which was in the shape of filaments and sheets, formed coolable beds because of the high void-age (∼0.9) and large particle size (∼10 mm). In a test with a 0.15-m sodium depth, the fragments from a pure uranium pour stream did not completely solidify but formed an agglomerated mass which did not fuse to the base plate. However, the agglomerated fragments of U-10 wt % Fe eutectic fused to the stainless steel base plate. An analysis of the temperature response of a 25-mm thick base plate was made by volume averaging the properties of the sodium and metal phases and assuming two semi-infinite solids coming into contact. Good agreement was obtained with the data during the initial 5 to 10 s of the contact period. 16 refs., 5 figs., 2 tabs

  3. Fundamentals of Hydrocarbon Upgrading to Liquid Fuels and Commodity Chemicals over Catalytic Metallic Nanoparticles

    Science.gov (United States)

    Chen, Tao

    Promising new technologies for biomass conversion into fuels and chemical feedstocks rely on the production of bio-oils, which need to be upgraded in order to remove oxygen-containing hydrocarbons and water. A high oxygen concentration makes bio-oils acidic and corrosive, unstable during storage, and less energetically valuable per unit weight than petroleum-derived hydrocarbons. Although there are efficient processes for the production of bio-oils, there are no efficient technologies for their upgrading. Current technologies utilize traditional petroleum refining catalysts, which are not optimized for biomass processing. New upgrading technologies are, therefore, urgently needed for development of sustainable energy resources. Development of such new technologies, however, is severely hindered by a lack of fundamental understanding of how oxygen and oxygen-containing hydrocarbons derived from biomass interact with promising noble-metal catalysts. In this study, kinetic reaction measurements, catalyst characterization and quantum chemical calculations using density functional theory were combined for determining adsorption modes and reaction mechanisms of hydrocarbons in the presence of oxygen on surfaces of catalytic noble-metal nanoparticles. The results were used for developing improved catalyst formulations and optimization of reaction conditions. The addition of molybdenum to platinum catalysts was shown to improve catalytic activity, stability, and selectivity in hydrodeoxygenation of acetic acid, which served as a model biomass compound. The fundamental results that describe interactions of oxygen and hydrocarbons with noble-metal catalysts were extended to other reactions and fields of study: evaluation of the reaction mechanism for hydrogen peroxide decomposition, development of improved hydrogenation catalysts and determination of adsorption modes of a spectroscopic probe molecule.

  4. Standard guide for pyrophoricity/combustibility testing in support of pyrophoricity analyses of metallic uranium spent nuclear fuel

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This guide covers testing protocols for testing the pyrophoricity/combustibility characteristics of metallic uranium-based spent nuclear fuel (SNF). The testing will provide basic data for input into more detailed computer codes or analyses of thermal, chemical, and mechanical SNF responses. These analyses would support the engineered barrier system (EBS) design bases and safety assessment of extended interim storage facilities and final disposal in a geologic repository. The testing also could provide data related to licensing requirements for the design and operation of a monitored retrievable storage facility (MRS) or independent spent fuel storage installation (ISFSI). 1.2 This guide describes testing of metallic uranium and metallic uranium-based SNF in support of transportation (in accordance with the requirements of 10CFR71), interim storage (in accordance with the requirements of 10CFR72), and geologic repository disposal (in accordance with the requirements of 10CFR60/63). The testing described ...

  5. Fast breeder fuel cycle

    International Nuclear Information System (INIS)

    1978-07-01

    This contribution is prepared for the answer to the questionnaire of working group 5, subgroup B. B.1. is the short review of the fast breeder fuel cycles based on the reference large commercial Japanese LMFBR. The LMFBRs are devided into two types. FBR-A is the reactor to be used before 2000, and its burnup and breeding ratio are relatively low. The reference fuel cycle requirement is calculated based on the FBR-A. FBR-B is the one to be used after 2000, and its burnup and breeding ratio are relatively high. B.2. is basic FBR fuel reprocessing scheme emphasizing the differences with LWR reprocessing. This scheme is based on the conceptual design and research and development work on the small scale LMFBR reprocessing facility of Japan. The facility adopts a conventional PUREX process except head end portions. The report also describes the effects of technical modifications of conventional reprocessing flow sheets, and the problems to be solved before the adoption of these alternatives

  6. Behavior of low-burnup metallic fuels for the integral fast reactor at elevated temperatures in ex-reactor tests

    International Nuclear Information System (INIS)

    Tsai, Hanchung; Liu, Yung Y.; Wang, Da-Yung; Kramer, J.M.

    1991-07-01

    A series of ex-reactor heating tests on low burnup U-26wt.%Pu-10wt.%Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel/cladding metallurgical interaction and its effect on cladding integrity at elevated temperatures. The reaction between the fuel and cladding caused liquid-phase formation and dissolution of the inner surface of the cladding. The rate of cladding penetration was below the existing design correlation, which provides a conservative margin to cladding failure. In a test which enveloped a wide range of postulated reactor transient events, a substantial temporal cladding integrity margin was demonstrated for an intact, whole fuel pin. The cause of the eventual pin breach was reaction-induced cladding thinning combined with fission-gas pressure loading. The behavior of the breached pin was benign. 7 refs., 7 figs., 1 tab

  7. A separate effect study of the influence of metallic fission products on CsI radioactive release from nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Di Lemma, F.G., E-mail: fidelma.dilemma@gmail.com [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Department of Radiation Science and Technology, Faculty of Applied Sciences, Delft University of Technology, Delft, 2629 JB (Netherlands); Colle, J.Y., E-mail: jean-yves.colle@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Beneš, O. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Konings, R.J.M. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Department of Radiation Science and Technology, Faculty of Applied Sciences, Delft University of Technology, Delft, 2629 JB (Netherlands)

    2015-10-15

    The chemistry of cesium and iodine is of main importance to quantify the radioactive release in case of a nuclear reactor accident, or sabotage involving irradiated nuclear materials. We studied the interaction of CsI with different metallic fission products such as Mo and Ru. These elements can be released from nuclear fuel when exposed to oxidising conditions, as in the case of contact of overheated nuclear fuel with air (e.g. in a spent fuel cask sabotage, uncovering of a spent fuel pond, or air ingress accidents). Experiments were performed by vaporizing mixtures of the compounds in air, and analysing the produced aerosols in view of a possible gas–gas and gas–aerosol reactions between the compounds. These results were compared with the gaseous species predicted by thermochemical equilibrium calculations and experimental equilibrium vaporization tests using Knudsen Effusion Mass Spectrometry.

  8. An evaluation of retention and disposal options for tritium in fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin, R.W. [E.I. du Pont de Nemours and Co., Aiken, SC (United States). Savannah River Lab.; Hampson, D.C. [Oak Ridge National Lab., TN (United States)

    1987-12-31

    This report assesses the possible options for retention of tritium and its ultimate disposal during future reprocessing of irradiated oxide fuels discharged from light water reactors (LWRs) and liquid metal fast breeder reactors (LMFBRs). The assessment includes an appraisal of the state of the retention and disposal options, an estimate of the dose commitments to the general public, an estimation of the incremental costs of the several retention and disposal options, and the potential reduction of the dose commitments resulting from retention and disposal of the tritium. The assessment is based upon an extensive study of tritium retention in reprocessing completed in 1982 by Grimes et al. Two plants were assumed, one to process LWR oxide fuel and the other to process LMFBR fuel. In each base case plant the tritium was vaporized to the atmosphere. Each of the hypothetical plants was assumed to be constructed during the 1990`s and to operate for a 20-year lifetime beginning in the year 2000 at a rate of 1,500 metric tons of heavy metal (MTHM) per 300-d year. In addition to the base case (Case 1), six other cases which included tritium retention options were examined. Although many of the features of the base-case plants remain unchanged in the tritium retention options, each case requires some additions, deletions, and modifications of portions of the plants. The retained tritium must also be managed and disposed of in a manner that is environmentally acceptable.

  9. Operational-safety advantages of LMFBR's: the EBR-II experience and testing program

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lindsay, R.W.; Golden, G.H.

    1982-01-01

    LMFBR's contain many inherent characteristics that simplify control and improve operating safety and reliability. The EBR-II design is such that good advantage was taken of these characteristics, resulting in a vary favorable operating history and allowing for a program of off-normal testing to further demonstrate the safe response of LMFBR's to upsets. The experience already gained, and that expected from the future testing program, will contribute to further development of design and safety criteria for LMFBR's. Inherently safe characteristics are emphasized and include natural convective flow for decay heat removal, minimal need for emergency power and a large negative reactivity feedback coefficient. These characteristics at EBR-II allow for ready application of computer diagnosis and control to demonstrate their effectiveness in response to simulated plant accidents. This latter testing objective is an important part in improvements in the man-machine interface

  10. Investigation of Redox Metal Oxides for Carbonaceous Fuel Conversion and CO2 Capture

    Science.gov (United States)

    Galinsky, Nathan Lee

    The chemical looping combustion (CLC) process uses metal oxides, also referred to as oxygen carriers, in a redox scheme for conversion of carbonaceous fuels into a concentrated stream of CO2 and steam while also producing heat and electricity. The unique redox scheme of CLC allows CO2 capture with minimal energy penalty. The CLC process performance greatly depends on the oxygen carrier that is chosen. To date, more than 1000 oxygen carriers have been developed for chemical-looping processes using metal oxides containing first-row transition metals. Oxygen carriers are typically mixed with an inert ceramic support to improve their overall mechanical stability and recyclability. This study focuses on design of (i) iron oxide oxygen carriers for conversion of gaseous carbonaceous fuels and (ii) development of perovskite CaMnO 3-d with improved stability and redox properties for conversion of solid fuels. Iron oxide is cheap and environmentally benign. However, it suffers from low activity with carbonaceous fuels due partially to the low ionic conductivity of iron oxides. In order to address the low activity of iron-oxide-based oxygen carriers, support addition has been shown to lower the energy barrier of oxygen anion transport within the oxygen carrier. This work adds a mixed-ionic-and-electronic-conductor (MIEC) support to iron oxide to help facilitate O2- transport inside the lattice of iron oxide. The MIEC-supported iron oxide is compared to commonly used supports including TiO2 and Al2O 3 and the pure ionic conductor support yttria-stabilized zirconia (YSZ) for conversion of different carbonaceous fuels and hydrogen. Results show that the MIEC-supported iron oxide exhibits up to 70 times higher activity than non-MIEC-supported iron oxides for methane conversion. The MIEC supported iron oxide also shows good recyclability with only minor agglomeration and carbon formation observed. The effect of support-iron oxide synergies is further investigated to understand

  11. Performance analysis of a membrane humidifier containing porous metal foam as flow distributor in a PEM fuel cell system

    International Nuclear Information System (INIS)

    Afshari, Ebrahim; Baharlou Houreh, Nasser

    2014-01-01

    Highlights: • Three metal foam configurations for the membrane humidifier are introduced. • The performances of the humidifiers containing metal foam are investigated. • A 3D CFD model is developed to compare the introduced humidifiers with one another. • Using metal foam at dry side has no positive effect on the humidifier performance. - Abstract: Using metal foam as flow distributor in membrane humidifier for proton exchange membrane (PEM) fuel cell system has some unique characteristics like more water transfer, low manufacturing complexity and low cost compared to the conventional flow channel plate. Metal foam can be applied at wet side or dry side or both sides of a humidifier. The three-dimensional CFD models are developed to investigate the performance of the above mentioned meanwhile compare them with the conventional humidifier. This model consists of a set of coupled equations including conservations of mass, momentum, species and energy for all regions of the humidifier. The results indicate that with the metal foam installed at wet side and both sides, water recovery ratio and dew point at dry side outlet are more than that of the conventional humidifier, indicating a better humidifier performance; while using metal foam at dry side has no positive effect on humidifier performance. At dry side mass flow rates higher than 10 mgr/s pressure drop in humidifier containing metal foam at wet side is lower than that of the conventional humidifier. As the mass flow rate increases from 9 to 15 mgr/s humidifier containing metal foam at wet side has better performance, while at mass flow rates lower than 9 mgr/s, the humidifier containing metal foam at both sides has better performance. At dry side inlet temperatures lower than 303 K, humidifier containing metal foam at wet side has better performance and at temperatures higher than 303 K, humidifier containing metal foam at both sides has better performance

  12. Flow-induced vibration in LMFBR steam generators: a state-of-the-art review

    International Nuclear Information System (INIS)

    Shin, Y.S.; Wambsganss, M.W.

    1975-05-01

    This state-of-the-art review identifies and discusses existing methods of flow-induced vibration analysis applicable to steam generators, their limitations, and base-technology needs. Also included are discussions of five different LMFBR steam-generator configurations and important design considerations, failure experiences, possible flow-induced excitation mechanisms, vibration testing, and available methods of vibration analysis. The objectives are to aid LMFBR steam-generator designers in making the best possible evaluation of potential vibration in steam-generator internals, and to provide the basis for development of design guidelines to avoid detrimental flow-induced vibration

  13. Comparative analysis of LMFBR licensing in the United States and other countries - notably France. Executive summary

    International Nuclear Information System (INIS)

    Golay, M.W.; Castillo, M.

    1981-01-01

    The safety-related design aspects and licensing experiences of LMFBR projects in other democratic countries have been studied and contrasted to those in the United States in order to understand the importance of different approaches to safety, and also to understand better the system of the United States. The regulatory systems and LMFBR programs of France and the United States are contrasted in detail, and that of West Germany is also studied. The programs of Japan and the United Kingdom receive considerably less attention, and that of the Soviet Union is ignored

  14. LMFBR safety criteria: cost-benefit considerations under the constraint of an a priori risk criterion

    International Nuclear Information System (INIS)

    Hartung, J.

    1979-01-01

    The role of cost-benefit considerations and a priori risk criteria as determinants of Core Disruptive Accident (CDA)-related safety criteria for large LMFBR's is explored with the aid of quantitative risk and probabilistic analysis methods. A methodology is described which allows a large number of design and siting alternatives to be traded off against each other with the goal of minimizing energy generation costs subject to the constraint of both an a priori risk criterion and a cost-benefit criterion. Application of this methodology to a specific LMFBR design project is described and the results are discussed. 5 refs

  15. Multicell slug flow heat transfer analysis of finite LMFBR bundles

    Energy Technology Data Exchange (ETDEWEB)

    Yeung, M.K.; Wolf, L.

    1978-12-01

    An analytical two-dimensional, multi-region, multi-cell technique has been developed for the thermal analysis of LMFBR rod bundles. Local temperature fields of various unit cells were obtained for 7, 19, and 37-rod bundles of different geometries and power distributions. The validity of the technique has been verified by its excellent agreement with the THTB calculational result. By comparing the calculated fully-developed circumferential clad temperature distribution with those of the experimental measurements, an axial correction factor has been derived to account for the entrance effect for practical considerations. Moreover, the knowledge of the local temperature field of the rod bundle leads to the determination of the effective mixing lengths L/sub ij/ for adjacent subchannels of various geometries. It was shown that the implementation of the accurately determined L/sub ij/ into COBRA-IIIC calculations has fairly significant effects on intersubchannel mixing. In addition, a scheme has been proposed to couple the 2-D distributed and lumped parameter calculation by COBRA-IIIC such that the entrance effect can be implanted into the distributed parameter analysis. The technique has demonstrated its applicability for a 7-rod bundle and the results of calculation were compared to those of three-dimensional analyses and experimental measurements.

  16. LMFBR plant design features for sodium spill and fire protection

    International Nuclear Information System (INIS)

    Palm, R.E.

    1982-01-01

    Design features have been developed for an LMFBR plant to protect the concrete structures from potential liquid spills and fires and prevent sodium-concrete reactions. The inclusion of these features in the plant design reduces the severity of design basis accident conditions imposed on containment and other critical plant structures. Steel liners are provided in cells containing radioactive sodium systems, and catch pans are located in non-radioactive sodium system cells. The design requirements and descriptions of each of these protective features are presented. The loading conditions, analytical approach and numerical results are also included. Design of concrete cell structures that are subject to high temperature effects from sodium spills is discussed. The structural design considers the influence of high temperature on design properties of concrete and carbon steel materials based on results of a comprehensive test program. The development of these design features and high temperature design considerations for the Clinch River Breeder Reactor Plant (CRBRP) are presented in this paper

  17. Development status of metallic, dispersion and non-oxide advanced and alternative fuels for power and research reactors

    International Nuclear Information System (INIS)

    2003-09-01

    The current thermal power reactors use less than 1% of the energy contained in uranium. Long term perspectives aiming at a better economical extraction of the potential supplied by uranium motivated the development of new reactor types and, of course, new fuel concepts. Most of them dated from the sixties including liquid metal cooled fast (FR) and high temperature gas cooled (HTGR) reactors. Unfortunately, these impulses slowed down during the last twenty years; nuclear energy had to face political and consensus problems, in particular in the United States of America and in Europe, resulting from the consequences of the TMI and Chernobyl accidents. Good economical results obtained by the thermal power reactors also contributed to this process. During the last twenty years mainly France, India, Japan and the Russian Federation have maintained a relatively high level of technological development with appropriate financial items, in particular, in fuel research for the above mentioned reactor types. China and South Africa are now progressing in development of FR/HTGR and HTGR technologies, respectively. The purpose of this report is not only to summarise knowledge accumulated in the fuel research since the beginning of the sixties. This subject has been well covered in literature up to the end of the eighties. This report rather concentrates on the 'advanced fuels 'for the current different types of reactors including metallic, carbide and nitride fuels for fast reactors, so-called 'cold' fuels and fuels to burn excessive ex-weapons plutonium in thermal power reactors, alternative fuels for small size and research reactors. Emphasis has been put on the aspects of fabrication and irradiation behaviour of these fuels; available basic data concerning essential properties that help to understand the phenomena have been mentioned as well. This report brings complementary information to the earlier published monographs and concerns developments carried out after the early

  18. Radioactivity concentration and heavy metal content in fuel oil and oil-ashes in Venezuela

    International Nuclear Information System (INIS)

    Barros, H.; Sajo-Bohus, L.; Abril, J.M.; Greaves, E.D.

    2004-01-01

    as well as Fuel Oil No6. Gamma spectrometry was used for measuring 226 Ra, 214 Pb, 214 Bi, 228 Ac, 212 Pb, 212 Bi, 208 Tl and 40 K, and heavy metallic cations were determined by ICP-MS which allow also the direct determination of 238 U. In this material (oil ashes) the total activity concentration is above 300 Bq.kg-1. In fact the combustion concentrate also heavy metal cations leading to an increasing in its concentrations, as is the case of Pb, Ni, Fe, Mn, V and Zn. Finally some recommendations are included for use of oil ashes as an additive for building materials. (author)

  19. Recycling Of Uranium- And Plutonium-Contaminated Metals From Decommissioning Of The Hanau Fuel Fabrication Plant

    International Nuclear Information System (INIS)

    Kluth, T.; Quade, U.; Lederbrink, F. W.

    2003-01-01

    Decommissioning of a nuclear facility comprises not only actual dismantling but also, above all, management of the resulting residual materials and waste. Siemens Decommissioning Projects (DP) in Hanau has been involved in this task since 1995 when the decision was taken to decommission and dismantle the Hanau Fuel Fabrication Plant. Due to the decommissioning, large amounts of contaminated steel scrap have to be managed. The contamination of this metal scrap can be found almost exclusively in the form of surface contamination. Various decontamination technologies are involved, as there are blasting and wiping. Often these methods are not sufficient to meet the free release limits. In these cases, SIEMENS has decided to melt the scrap at Siempelkamp's melting plant. The plant is licensed according to the German Radiation Protection Ordinance Section 7 (issue of 20.07.2001). The furnace is a medium frequency induction type with a load capacity of 3.2 t and a throughput of 2 t/h for steel melting. For safety reasons, the furnace is widely operated by remote handling. A highly efficient filter system of cyclone, bag filter and HEPA-filter in two lines retains the dust and aerosol activity from the off-gas system. The slag is solidified at the surface of the melt and gripped before pouring the liquid iron into a chill. Since 1989, in total 15,000 t have been molten in the plant, 2,000 t of them having been contaminated steel scrap from the decommissioning of fuel fabrication plants. Decontamination factors could be achieved between 80 and 100 by the high affinity of the uranium to the slag former. The activity is transferred to the slag up to nearly 100 %. Samples taken from metal, slag and dust are analyzed by gamma measurements of the 186 keV line of U235 and the 1001 keV line of Pa234m for U238. All produced ingots showed a remaining activity less than 1 Bq/g and could be released for industrial reuse

  20. Upper limits to americium concentration in large sized sodium-cooled fast reactors loaded with metallic fuel

    International Nuclear Information System (INIS)

    Zhang, Youpeng; Wallenius, Janne

    2014-01-01

    Highlights: • The americium transmutation capability of Integral Fast Reactor was investigated. • The impact from americium introduction was parameterized by applying SERPENT Monte Carlo calculations. • Higher americium content in metallic fuel leads to a power penalty, preserving consistent safety margins. - Abstract: Transient analysis of a large sized sodium-cooled reactor loaded with metallic fuel modified by different fractions of americium have been performed. Unprotected loss-of-offsite power, unprotected loss-of-flow and unprotected transient-over-power accidents were simulated with the SAS4A/SASSYS code based on the geometrical model of an IFR with power rating of 2500 MW th , using safety parameters obtained with the SERPENT Monte Carlo code. The Ti-modified austenitic D9 steel, having higher creep rupture strength, was considered as the cladding and structural material apart from the ferritic/martensitic HT9 steel. For the reference case of U–12Pu–1Am–10Zr fuel at EOEC, the margin to fuel melt during a design basis condition UTOP is about 50 K for a maximum linear rating of 30 kW/m. In order to maintain a margin of 50 K to fuel failure, the linear power rating has to be reduced by ∼3% and 6% for 2 wt.% and 3 wt.% Am introduction into the fuel respectively. Hence, an Am concentration of 2–3 wt.% in the fuel would lead to a power penalty of 3–6%, permitting a consumption rate of 3.0–5.1 kg Am/TW h th . This consumption rate is significantly higher than the one previously obtained for oxide fuelled SFRs