Science.gov

Sample records for core fuel reload

  1. Modeling and design of a reload PWR core for a 48-month fuel cycle

    SciTech Connect

    McMahon, M.V.; Driscoll, M.J.; Todreas, N.E.

    1997-05-01

    The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7{sup w}/{sub o} U{sup 235} for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd{sub 2}O{sub 3}) mixed with the UO{sub 2} of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB{sub 2}) integral fuel burnable absorber (IFBA) coating on the Gd{sub 2}O{sub 3}-UO{sub 2} fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted.

  2. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  3. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2011-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  4. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2013-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  5. Hybrid expert system implementation to determine core reload patterns

    SciTech Connect

    Greek, K.J.; Robinson, A.H.

    1989-01-01

    Determining reactor reload fuel patterns is a computationally intensive problem solving process for which automation can be of significant benefit. Often much effort is expended in the search for an optimal loading. While any modern programming language could be used to automate solution, the specialized tools of artificial intelligence (AI) are the most efficient means of introducing the fuel management expert's knowledge into the search for an optimum reload pattern. Prior research in pressurized water reactor refueling strategies developed FORTRAN programs that automated an expert's basic knowledge to direct a search for an acceptable minimum peak power loading. The dissatisfaction with maintenance of compiled knowledge in FORTRAN programs has served as the motivation for the development of the SHUFFLE expert system. SHUFFLE is written in Smalltalk, an object-oriented programming language, and evaluates loadings as it generates them using a two-group, two-dimensional nodal power calculation compiled in a personal computer-based FORTRAN. This paper reviews the object-oriented representation developed to solve the core reload problem with an expert system tool and its operating prototype, SHUFFLE.

  6. BWR Reload Strategy Based on Fixing Once-Burnt Fuel Between Cycles

    SciTech Connect

    Maag, Elizebeth M.; Knott, Dave

    2001-12-15

    The feasibility of a reload strategy based on fixing the locations of once-burnt fuel between cycles has been evaluated for the Perry nuclear power plant (Perry). This strategy can reduce refueling shuffle critical path time by 3 days without penalty in fuel cycle economics. The scheme works well for Perry because of the extreme cycle energy requirements and the large feed batch size needed to meet those requirements. Cores requiring less energy and a smaller feed batch size have not been investigated.

  7. Fuel management and reloading optimization at EdF

    SciTech Connect

    Rosset, F.D.; Barral, J.C. )

    1993-01-01

    Technical and economical pressurized water reactor (PWR) performances are strongly influenced by fuel management, e.g., fuel utilization and core design. Because of the large number of standardized French PWR units, this question is of considerable importance for Electricite de France (EdF). At present, EdF operates two standardized types of PWR: thirty-four 900-MW and twenty 1300-MW PWRS. Economic optimization will lead to global management of the nuclear power plants in the coming years through three main types of fuel management: four-batch 3.7% UO[sub 2] management and plutonium recycling management for the 900-MW PWRs and extended cycle management for the 1300-MW PWRS. The best optimization is made on each reactor by computing a loading pattern that flattens the power map to ensure a certain flexibility of operation (early shutdown, stretch-out).

  8. Evaluation of reactivity shutdown margin for nuclear fuel reload optimization

    SciTech Connect

    Wong, Hing-Ip; Maldonado, G.I.

    1995-12-31

    The FORMOSA-P code is a nuclear fuel management optimization package that combines simulated annealing (SA) and nodal generalized perturbation theory (GPT). Recent studies at Electricite de France (EdF-Clamart) have produced good results for power-peaking minimizations under multiple limiting control rod configurations. However, since the reactivity shutdown margin is not explicitly treated as an objective or constraint function, then any optimal loading patterns (LPs) are not guaranteed to yield an adequate shutdown margin (SDM). This study describes the implementation of the SDM calculation within a FORMOSA-P optimization. Maintaining all additional computational requirements to a minimum was a key consideration.

  9. Fuel and Core Design Experiences in Cofrentes NPP

    SciTech Connect

    Garcia-Delgado, L.; Lopez-Carbonell, M.T.; Gomez-Bernal, I.

    2002-07-01

    The electricity market deregulation in Spain is increasing the need for innovations in nuclear power generation, which can be achieved in the fuel area by improving fuel and core designs and by introducing vendors competition. Iberdrola has developed the GIRALDA methodology for design and licensing of Cofrentes reloads, and has introduced mixed cores with fuel from different vendors. The application of GIRALDA is giving satisfactory results, and is showing its capability to adequately reproduce the core behaviour. The nuclear design team is acquiring an invaluable experience and a deep knowledge of the core, very useful to support cycle operation. Continuous improvements are expected for the future in design strategies as well as in the application of new technologies to redesign the methodology processes. (authors)

  10. Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core

    SciTech Connect

    Sterbentz, James W

    2007-05-01

    A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle can be achieved with the lower enrichment fuel.

  11. Nuclear core and fuel assemblies

    DOEpatents

    Downs, Robert E.

    1981-01-01

    A fast flux nuclear core of a plurality of rodded, open-lattice assemblies having a rod pattern rotated relative to a rod support structure pattern. Elongated fuel rods are oriented on a triangular array and laterally supported by grid structures positioned along the length of the assembly. Initial inter-assembly contact is through strongbacks at the corners of the support pattern and peripheral fuel rods between adjacent assemblies are nested so as to maintain a triangular pitch across a clearance gap between the other portions of adjacent assemblies. The rod pattern is rotated relative to the strongback support pattern by an angle .alpha. equal to sin .sup.-1 (p/2c), where p is the intra-assembly rod pitch and c is the center-to-center spacing among adjacent assemblies.

  12. IN-CORE FUEL MANAGEMENT: PWR Core Calculations Using MCRAC

    NASA Astrophysics Data System (ADS)

    PetroviĆ, B. G.

    1991-01-01

    The following sections are included: * INTRODUCTION * IN-CORE FUEL MANAGEMENT CALCULATIONS * In-Core Fuel Management * Methodological Problems of In-Core Fuel Management * In-Core Fuel Management Analytical Tools * PENN STATE FUEL MANAGEMENT PACKAGE * Penn State Fuel Management Package (PFMP) * Assembly Data Description (ADD) * Linking PSU-LEOPARD and MCRAC: An Example * MULTICYCLE REACTOR ANALYSIS CODE (MCRAC) * Main Features and Options of MCRAC code * Core geometry * Diffusion equations * 1.5-group model * Multicycle neutronic analysis * Multicycle cost analysis * Criticality search * Power-dependent xenon feedback calculations * Control rod and burnable absorber simulation * Search for LP with flat BOC power distribution * Artificial ADD option * Variable dimensioning technique * RBI version of MCRAC code * Programming changes in PC version * Fuel interchange option * MCRAC Input/Output * General input description * Sample input * Sample output * EXPERIENCE WITH MCRAC CODE * CONCLUSIONS * REFERENCES

  13. A reload and startup plan for conversion of the NIST research reactor

    SciTech Connect

    D. J. Diamond

    2016-03-31

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reload portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.

  14. Benchmark Evaluation of the Neutron Radiography (NRAD) Reactor Upgraded LEU-Fueled Core

    SciTech Connect

    John D. Bess

    2001-09-01

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. The final upgraded core configuration with 64 fuel elements has been completed. Evaluated benchmark measurement data include criticality, control-rod worth measurements, shutdown margin, and excess reactivity. Dominant uncertainties in keff include the manganese content and impurities contained within the stainless steel cladding of the fuel and the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 nuclear data are approximately 1.4% greater than the benchmark model eigenvalue, supporting contemporary research regarding errors in the cross section data necessary to simulate TRIGA-type reactors. Uncertainties in reactivity effects measurements are estimated to be ~10% with calculations in agreement with benchmark experiment values within 2s. The completed benchmark evaluation de-tails are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Experiments (IRPhEP Handbook). Evaluation of the NRAD LEU cores containing 56, 60, and 62 fuel elements have also been completed, including analysis of their respective reactivity effects measurements; they are also available in the IRPhEP Handbook but will not be included in this summary paper.

  15. Mox fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-05-15

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

  16. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    1998-01-01

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  17. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-07-17

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  18. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, M.L.; Rosenstein, R.G.

    1998-10-13

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

  19. A multi-objective shuffled frog leaping algorithm for in-core fuel management optimization

    NASA Astrophysics Data System (ADS)

    Arshi, S. Safaei; Zolfaghari, A.; Mirvakili, S. M.

    2014-10-01

    The efficient operation and in-core fuel management of PWRs are of utmost importance. In the present work, a core reload optimization using Shuffled Frog Leaping (SFL) algorithm is addressed and mapped on nuclear fuel loading pattern optimization. SFL is one of the latest meta-heuristic optimization algorithms which is used for solving the discrete optimization problems and inspired from social behavior of frogs. The algorithm initiates the search from an initial population and carries forward to draw out an optimum result. This algorithm employs the use of memetic evolution by exchanging ideas between the members of the population in each local search. The local search of SFL is similar to particle swarm optimization (PSO) and applying shuffling process accomplishes the information exchange between several local searches to obtain an overall optimum result. To evaluate the proposed technique, Shekel's Foxholes and a VVER-1000 reactor are used as test cases to illustrate performance of SFL. Among numerous neutronic and thermal-hydraulic objectives necessary for a fuel management problem to reach an overall optimum, this paper deals with two neutronic objectives, i.e., maximizing effective multiplication factor and flattening power distribution in the core, to evaluate the capability of applying SFL algorithm for a fuel management problem. The results, convergence rate and reliability of the method are quite promising and show the potential and efficiency of the technique for other optimization applications in the nuclear engineering field.

  20. Fuels and materials for LMFBR core components

    SciTech Connect

    Cox, C M; Jackson, R J; Straalsund, J L

    1984-04-01

    This paper reviews development of fuels and materials for Liquid Metal Fast Breeder Reactor. Included are the status of fuels and materials technology for LMFBR core components. The fuel assembly for the Fast Flux Test Facility, or FFTF, in operation near Richland, Washington, is described. The outer part of the 12-ft long assembly is called a flow channel or duct. Inside are 217 fuel pins, each containing mixed uranium-plutonium oxide fuel pellets. The comparable schematic for control rod or absorber assembly is also shown. The FFTF absorber assembly contains 61 control rods containing boron carbide pellets. Because FFTF is a test reactor, it does not contain blanket assemblies; however, the Clinch River Breeder Reactor blanket assemblies look very similar to the FFTF fuel assembly, except that they each contain 61 UO/sub 2/ rods. Sizes of various LMFBR fuel assemblies are compared. The Clinch River Breeder Reactor fuel assembly is nearly identical to that of FFTF, except for an increased length to accommodate UO/sub 2/ axial blankets within the fuel pins. The DP-1 design is for a large breeder reactor and uses larger ducts and more fuel pins per assembly. By comparison, the fuel assemblies for EBR-II are much smaller, as is the EBR-II core.

  1. NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

    DOEpatents

    Moore, W.T.

    1958-09-01

    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  2. Reload design process at Yankee Atomic Electric Company

    SciTech Connect

    Weader, R.J.

    1986-01-01

    Yankee Atomic Electric Company (YAEC) performs reload design and licensing for their nuclear power plants: Yankee Rowe, Maine Yankee, and Vermont Yankee. Significant savings in labor and computer costs have been achieved in the reload design process by the use of the SIMULATE nodal code using the CASMO assembly burnup code or LEOPARD pin cell burnup code inputs to replace the PDQ diffusion theory code in many required calculations for the Yankee Rowe and Maine Yankee pressurized water reactors (PWRs). An efficient process has evolved for the design of reloads for the Vermont Yankee boiling water reactor (BWR). Due to the major differences in the core design of the three plants, different reload design processes have evolved for each plant.

  3. Solid oxide fuel cell with monolithic core

    DOEpatents

    McPheeters, Charles C.; Mrazek, Franklin C.

    1988-01-01

    A solid oxide fuel cell in which fuel and oxidant gases undergo an electrochemical reaction to produce an electrical output includes a monolithic core comprised of a corrugated conductive sheet disposed between upper and lower generally flat sheets. The corrugated sheet includes a plurality of spaced, parallel, elongated slots which form a series of closed, linear, first upper and second lower gas flow channels with the upper and lower sheets within which a fuel gas and an oxidant gas respectively flow. Facing ends of the fuel cell are generally V-shaped and provide for fuel and oxidant gas inlet and outlet flow, respectively, and include inlet and outlet gas flow channels which are continuous with the aforementioned upper fuel gas and lower oxidant gas flow channels. The upper and lower flat sheets and the intermediate corrugated sheet are preferably comprised of ceramic materials and are securely coupled together such as by assembly in the green state and sintering together during firing at high temperatures. A potential difference across the fuel cell, or across a stacked array of similar fuel cells, is generated when an oxidant gas such as air and a fuel such as hydrogen gas is directed through the fuel cell at high temperatures, e.g., between 700.degree. C. and 1100.degree. C.

  4. Solid oxide fuel cell with monolithic core

    DOEpatents

    McPheeters, C.C.; Mrazek, F.C.

    1988-08-02

    A solid oxide fuel cell in which fuel and oxidant gases undergo an electrochemical reaction to produce an electrical output includes a monolithic core comprised of a corrugated conductive sheet disposed between upper and lower generally flat sheets. The corrugated sheet includes a plurality of spaced, parallel, elongated slots which form a series of closed, linear, first upper and second lower gas flow channels with the upper and lower sheets within which a fuel gas and an oxidant gas respectively flow. Facing ends of the fuel cell are generally V-shaped and provide for fuel and oxidant gas inlet and outlet flow, respectively, and include inlet and outlet gas flow channels which are continuous with the aforementioned upper fuel gas and lower oxidant gas flow channels. The upper and lower flat sheets and the intermediate corrugated sheet are preferably comprised of ceramic materials and are securely coupled together such as by assembly in the green state and sintering together during firing at high temperatures. A potential difference across the fuel cell, or across a stacked array of similar fuel cells, is generated when an oxidant gas such as air and a fuel such as hydrogen gas is directed through the fuel cell at high temperatures, e.g., between 700 C and 1,100 C. 8 figs.

  5. Solid oxide fuel cell having monolithic core

    DOEpatents

    Ackerman, J.P.; Young, J.E.

    1983-10-12

    A solid oxide fuel cell is described for electrochemically combining fuel and oxidant for generating galvanic output, wherein the cell core has an array of electrolyte and interconnect walls that are substantially devoid of any composite inert materials for support. Instead, the core is monolithic, where each electrolyte wall consists of thin layers of cathode and anode materials sandwiching a thin layer of electrolyte material therebetween. The electrolyte walls are arranged and backfolded between adjacent interconnect walls operable to define a plurality of core passageways alternately arranged where the inside faces thereof have only the anode material or only the cathode material exposed. Means direct the fuel to the anode-exposed core passageways and means direct the oxidant to the anode-exposed core passageways and means direct the oxidant to the cathode-exposed core passageway; and means also direct the galvanic output to an exterior circuit. Each layer of the electrolyte and interconnect materials is of the order of 0.002 to 0.01 cm thick; and each layer of the cathode and anode materials is of the order of 0.002 to 0.05 cm thick.

  6. Solid oxide fuel cell having monolithic core

    DOEpatents

    Ackerman, John P.; Young, John E.

    1984-01-01

    A solid oxide fuel cell for electrochemically combining fuel and oxidant for generating galvanic output, wherein the cell core has an array of electrolyte and interconnect walls that are substantially devoid of any composite inert materials for support. Instead, the core is monolithic, where each electrolyte wall consists of thin layers of cathode and anode materials sandwiching a thin layer of electrolyte material therebetween, and each interconnect wall consists of thin layers of the cathode and anode materials sandwiching a thin layer of interconnect material therebetween. The electrolyte walls are arranged and backfolded between adjacent interconnect walls operable to define a plurality of core passageways alternately arranged where the inside faces thereof have only the anode material or only the cathode material exposed. Means direct the fuel to the anode-exposed core passageways and means direct the oxidant to the cathode-exposed core passageway; and means also direct the galvanic output to an exterior circuit. Each layer of the electrolyte and interconnect materials is of the order of 0.002-0.01 cm thick; and each layer of the cathode and anode materials is of the order of 0.002-0.05 cm thick.

  7. Solid oxide fuel cell having monolithic core

    NASA Astrophysics Data System (ADS)

    Ackerman, J. P.; Young, J. E.

    1983-10-01

    A solid oxide fuel cell is described for electrochemically combining fuel and oxidant for generating galvanic output, wherein the cell core has an array of electrolyte and interconnect walls that are substantially devoid of any composite inert materials for support. The core is monolithic, where each electrolyte wall consists of thin layers of cathode and anode materials sandwiching a thin layer of electrolyte material. The electrolyte walls are arranged and backfolded between adjacent interconnect walls operable to define a plurality of core passageways alternately arranged where the inside faces have only the anode material or only the cathode material exposed. Each layer of the electrolyte and interconnect materials 0.002 to 0.01 cm thick; and each layer of the cathode and anode materials is 0.002 to 0.05 cm thick.

  8. Westinghouse Fuel Assemblies Performance after Operation in South-Ukraine NPP Mixed Core

    SciTech Connect

    Abdullayev, A. M.; Kulish, G. V.; Slyeptsov, O.; Slyeptsov, S.; Aleshin, Y.; Sparrow, S.; Lashevych, P.; Sokolov, D.; Latorre, Richard

    2013-09-14

    The evaluation of WWER-1000 Westinghouse fuel performance was done using the results of post–irradiation examinations of six LTAs and the WFA reload batches that have operated normally in mixed cores at South-Ukraine NPP, Unit-3 and Unit-2. The data on WFA/LTA elongation, FR growth and bow, WFA bow and twist, RCCA drag force and drag work, RCCA drop time, FR cladding integrity as well as the visual observation of fuel assemblies obtained during the 2006-2012 outages was utilized. The analysis of the measured data showed that assembly growth, FR bow, irradiation growth, and Zr-1%Nb grid and ZIRLO cladding corrosion lies within the design limits. The RCCA drop time measured for the LTA/WFA is about 1.9 s at BOC and practically does not change at EOC. The measured WFA bow and twist, and data of drag work on RCCA insertion showed that the WFA deformation in the mixed core is mostly controlled by the distortion of Russian FAs (TVSA) having the higher lateral stiffness. The visual inspection of WFAs carried out during the 2012 outages revealed some damage to the Zr-1%Nb grid outer strap for some WFAs during the loading sequence. The performed fundamental investigations allowed identifying the root cause of grid outer strap deformation and proposing the WFA design modifications for preventing damage to SG at a 225 kg handling trip limit.

  9. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    SciTech Connect

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail,

    2010-12-23

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this

  10. Development of an LMR core design using metallic fuel

    SciTech Connect

    Orechwa, Y.; Yang, S.T.

    1986-01-01

    The design and performance of a metal fueled LMR core at the 900 MWth power level is described. Performance measures of preliminary driver and blanket fuel pin designs are calculated with the developmental fuel performance code LIFE-metal. The robustness of the designs to changes in design parameters is shown with respect to plenum length, cladding thickness and cladding temperature.

  11. Role of Minor Actinides for Long-Life Reactor Cores

    SciTech Connect

    Saito, M.; Artisyuk, V.; Shmelev, A.; Nikitin, K.; Peryoga, Y

    2002-07-01

    The paper addresses the study on advanced fuel cycles for LWR oriented to high burnup values that exceed 100 GWd/tHM, thus giving the chance to establish the long-life reactor cores without fuel reloading on site. The key element of this approach is a broad involvement of Minor Actinides whose admixture to 20% enriched uranium fuel provides safe release of initial reactivity excess and improved proliferation resistance properties. (authors)

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

    SciTech Connect

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

    1984-01-01

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

  13. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    SciTech Connect

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are

  14. Bassoon Speeds Vesicle Reloading at a Central Excitatory Synapse

    PubMed Central

    Hallermann, Stefan; Fejtova, Anna; Schmidt, Hartmut; Weyhersmüller, Annika; Silver, R. Angus; Gundelfinger, Eckart D.; Eilers, Jens

    2010-01-01

    Summary Sustained rate-coded signals encode many types of sensory modalities. Some sensory synapses possess specialized ribbon structures, which tether vesicles, to enable high-frequency signaling. However, central synapses lack these structures, yet some can maintain signaling over a wide bandwidth. To analyze the underlying molecular mechanisms, we investigated the function of the active zone core component Bassoon in cerebellar mossy fiber to granule cell synapses. We show that short-term synaptic depression is enhanced in Bassoon knockout mice during sustained high-frequency trains but basal synaptic transmission is unaffected. Fluctuation and quantal analysis as well as quantification with constrained short-term plasticity models revealed that the vesicle reloading rate was halved in the absence of Bassoon. Thus, our data show that the cytomatrix protein Bassoon speeds the reloading of vesicles to release sites at a central excitatory synapse. PMID:21092860

  15. Thermal barrier and support for nuclear reactor fuel core

    DOEpatents

    Betts, Jr., William S.; Pickering, J. Larry; Black, William E.

    1987-01-01

    A thermal barrier/core support for the fuel core of a nuclear reactor having a metallic cylinder secured to the reactor vessel liner and surrounded by fibrous insulation material. A top cap is secured to the upper end of the metallic cylinder that locates and orients a cover block and post seat. Under normal operating conditions, the metallic cylinder supports the entire load exerted by its associated fuel core post. Disposed within the metallic cylinder is a column of ceramic material, the height of which is less than that of the metallic cylinder, and thus is not normally load bearing. In the event of a temperature excursion beyond the design limits of the metallic cylinder and resulting in deformation of the cylinder, the ceramic column will abut the top cap to support the fuel core post.

  16. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    SciTech Connect

    Mohammed, Abdul Aziz Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

    2016-01-22

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  17. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    NASA Astrophysics Data System (ADS)

    Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

    2016-01-01

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  18. Core-shell fuel cell electrodes

    DOEpatents

    Adzic, Radoslav; Bliznakov, Stoyan; Vukmirovic, Miomir

    2017-07-25

    Embodiments of the disclosure relate to electrocatalysts. The electrocatalyst may include at least one gas-diffusion layer having a first side and a second side, and particle cores adhered to at least one of the first and second sides of the at least one gas-diffusion layer. The particle cores includes surfaces adhered to the at least one of the first and second sides of the at least one gas-diffusion layer and surfaces not in contact with the at least one gas-diffusion layer. Furthermore, a thin layer of catalytically atoms may be adhered to the surfaces of the particle cores not in contact with the at least one gas-diffusion layer.

  19. Advances in fuel management and on-line core monitoring

    SciTech Connect

    Stout, R.B.; Hansen, L.E.; Patten, T.W.

    1988-01-01

    Advanced Nuclear Fuels Corporation (ANF) has developed and implemented advanced core power distribution monitoring methods for BWRs and PWRs based on the three dimensional nodal simulator codes used for incore fuel management design and analysis. The use of these methods has resulted in a more accurate assessment of the core power distribution and corresponding increased operating margins. These increased margins allow for more economical fuel cycle designs. Since the initial application in 1982, ANF has made enhancements to the incore monitoring system. These enhancements have permitted more rapid analysis of local power changes, power distribution projections during ascent to full power and on-line statistical analysis of the incore detector signal. The on-line analysis implemented in BWRs has also been developed for application PWRs. In the future, reactors are expected to operate with longer fuel cycles, more aggressive low radial leakage loadings, load follow and use higher burnup fuel. These advances will require more burnable neutron absorbers and more sophisticated fuel designs. To accommodate these advances, the fuel management methodologies and measurement system will require improvements. The state-of-the-art methods provided by ANF provide incore monitoring systems compatible with these expected needs.

  20. Nuclear Fuel Management Optimization: A Work in Progress

    SciTech Connect

    Turinsky, Paul J.

    2005-07-15

    The focus of this overview for this issue of Nuclear Technology, which contains papers presented at the American Nuclear Society Advances in Nuclear Fuel Management III (ANFM-III) 2004 topical meeting, is to introduce the subject of nuclear fuel management for light water reactors. A total of 23 papers was presented on this topic at ANFM-III. Nuclear fuel management involves making the so-called out-of-core and in-core decisions. Simply put, the out-of-core decisions address the attributes of the new (fresh) fuel that will be fabricated and the partially burnt (shuffled) fuel to reinsert into the core for additional energy production. The in-core decisions address where the fresh and burnt fuel along with burnable poisons should be located in the core. The above applies to batch refueling strategies, e.g., pressurized water reactors and boiling water reactors (BWRs). For BWRs, additional in-core decisions enter to address control rod pattern paired with core flow rate as a function of burnup. It is obvious that the out-of-core and in-core decisions are coupled.The objective of nuclear fuel management is to minimize the cost of electrical energy generation subject to operational and safety constraints. Since fuel resides in the core for several cycles, a multicycle assessment is required to make nuclear fuel management decisions. For nearly four decades there has been an effort to develop automated computational capability to assist the reload core nuclear design engineer in making nuclear fuel management decisions. This development has ranged from employment of heuristic rules to utilization of mathematical optimization approaches. This overview reviews the development of nuclear fuel management optimization capabilities by first defining the problem, then describing current capabilities, and finally projecting where future capabilities need to be developed to support the needs of reload core nuclear design engineers.

  1. Serially connected solid oxide fuel cells having monolithic cores

    DOEpatents

    Herceg, J.E.

    1985-05-20

    Disclosed is a solid oxide fuel cell for electrochemically combining fuel and oxidant for generating galvanic output. The cell core has an array of cell segments electrically serially connected in the flow direction, each segment consisting of electrolyte walls and interconnect that are substantially devoid of any composite inert materials for support. Instead, the core is monolithic, where each electrolyte wall consists of thin layers of cathode and anode materials sandwiching a thin layer of electrolyte material therebetween. Means direct the fuel to the anode-exposed core passageways and means direct the oxidant to the cathode-exposed core passageways; and means also direct the galvanic output to an exterior circuit. Each layer of the electrolyte composite materials is of the order of 0.002 to 0.01 cm thick; and each layer of the cathode and anode materials is of the order of 0.002 to 0.05 cm thick. Between 2 and 50 cell segments may be connected in series.

  2. An americium-fueled gas core nuclear rocket

    SciTech Connect

    Kammash, T.; Galbraith, D.L.; Jan, T. )

    1993-01-10

    A gas core fission reactor that utilizes americium in place of uranium is examined for potential utilization as a nuclear rocket for space propulsion. The isomer [sup 242m]Am with a half life of 141 years is obtained from an (n, [gamma]) capture reaction with [sup 241]Am, and has the highest known thermal fission cross section. We consider a 7500 MW reactor, whose propulsion characteristics with [sup 235]U have already been established, and re-examine it using americium. We find that the same performance can be achieved at a comparable fuel density, and a radial size reduction (of both core and moderator/reflector) of about 70%.

  3. Laser cutting apparatus for nuclear core fuel subassembly

    DOEpatents

    Walch, Allan P.; Caruolo, Antonio B.

    1982-02-23

    The object of the invention is to provide a system and apparatus which employs laser cutting to disassemble a nuclear core fuel subassembly. The apparatus includes a gantry frame (C) which straddles the core fuel subassembly (14), an x-carriage (22) travelling longitudinally above the frame which carries a focus head assembly (D) having a vertically moving carriage (46) and a laterally moving carriage (52), a system of laser beam transferring and focusing mirrors carried by the x-carriage and focusing head assembly, and a shroud follower (F) and longitudinal follower (G) for following the shape of shroud (14) to maintain a beam focal point (44) fixed upon the shroud surface for accurate cutting.

  4. Core conversion of the Portuguese research reactor to LEU fuel

    SciTech Connect

    Marques, J.G.; Ramos, A.R.; Kocher, A.

    2008-07-15

    Core conversion of the Portuguese Research Reactor (RPI) to LEU fuel is being performed within IAEA's Technical Cooperation project POR/4/016, with financial support from the US and Portugal. CERCA was selected as manufacturer of the LEU assemblies by the IAEA after an international call for bids. CERCA provided a comprehensive package to the RPI which included the mechanical verification of the design of the assemblies, their manufacture and arrangements for a joint inspection of the finished assemblies. The LEU fuel assemblies were manufactured within 8 months upon final approval of the design. The safety analyses for the core conversion to LEU fuel were made with the assistance of the RERTR program and were submitted for review by the IAEA and by Portuguese authorities in January 2007. Revised documents were submitted in June 2007 addressing the issues raised during review. Regulatory approval was received in early August and core conversion was done in early September. All measured safety parameters are within the defined acceptance limits. Operation at full power is expected by the end of October. (author)

  5. Physics implications of oxide and metal fuel on the design of small LMFBR cores

    SciTech Connect

    Orechwa, Y.; Khalil, H.

    1984-09-17

    Slower growth projections in the demand for electricity and advances in metal fuel technology have shifted some of the emphasis in fast reactor development in the US from large oxide cores to small cores and also renewed interest in metal fuel. Cores constrained by diameter and fuel burnup exhibit many similar neutronic performance characteristics. However, some parameters such as reactivity coefficients, for example, are very different. The physics parameters of the four cores studied suggest that metal fueled cores, although less developed than oxide cores, are more flexible in adapting to currently changing deployment scenarios.

  6. Pressurized water reactor core parameter prediction using an artificial neural network

    SciTech Connect

    Kim, Han Gon; Chang, Soon Heung; Lee, Byung Ho )

    1993-01-01

    In pressurized water reactors, the fuel reloading problem has significant meaning in terms of both safety and economics. The local power peaking factor must be kept lower than a predetermined value during a cycle, and the effective multiplication factor must be maximized to extract the maximum energy. If these core parameters could be obtained in a very short time, the optimal fuel reloading patterns would be found more effectively and quickly. A very fast core parameter prediction system is developed using the back propagation neural network. This system predicts the core parameters several hundred times as fast as the reference numerical code, within an error of a few percent. The effects of the variation of the training rate coefficients, the momentum, and the hidden layer units are also discussed.

  7. Fuel and core testing plan for a target fueled isotope production reactor.

    SciTech Connect

    Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

    2010-12-01

    In recent years there has been an unstable supply of the critical diagnostic medical isotope 99Tc. Several concepts and designs have been proposed to produce 99Mo the parent nuclide of 99Tc, at a commercial scale sufficient to stabilize the world supply. This work lays out a testing and experiment plan for a proposed 2 MW open pool reactor fueled by Low Enriched Uranium (LEU) 99Mo targets. The experiments and tests necessary to support licensing of the reactor design are described and how these experiments and tests will help establish the safe operating envelop for a medical isotope production reactor is discussed. The experiments and tests will facilitate a focused and efficient licensing process in order to bring on line a needed production reactor dedicated to supplying medical isotopes. The Target Fuel Isotope Reactor (TFIR) design calls for an active core region that is approximately 40 cm in diameter and 40 cm in fuel height. It contains up to 150 cylindrical, 1-cm diameter, LEU oxide fuel pins clad with Zircaloy (zirconium alloy), in an annular hexagonal array on a {approx}2.0 cm pitch surrounded, radially, by a graphite or a Be reflector. The reactor is similar to U.S. university reactors in power, hardware, and safety/control systems. Fuel/target pin fabrication is based on existing light water reactor fuel fabrication processes. However, as part of licensing process, experiments must be conducted to confirm analytical predictions of steady-state power and accident conditions. The experiment and test plan will be conducted in phases and will utilize existing facilities at the U.S. Department of Energy's Sandia National Laboratories. The first phase is to validate the predicted reactor core neutronics at delayed critical, zero power and very low power. This will be accomplished by using the Sandia Critical Experiment (CX) platform. A full scale TFIR core will be built in the CX and delayed critical measurements will be taken. For low power experiments

  8. Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

    SciTech Connect

    Stauff, N.E.; Klim, T.K.; Taiwo, T.A.; Fiorina, C.; Franceschini, F.

    2013-07-01

    A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueled cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic

  9. The whole-core LEU silicide fuel demonstration in the JMTR

    SciTech Connect

    Aso, Tomokazu; Akashi, Kazutomo; Nagao, Yoshiharu

    1997-08-01

    The JMTR was fully converted to LEU silicide (U{sub 3}Si{sub 2}) fuel with cadmium wires as burnable absorber in January, 1994. The reduced enrichment program for the JMTR was initiated in 1979, and the conversion to MEU (enrichment ; 45%) aluminide fuel was carried out in 1986 as the first step of the program. The final goal of the program was terminated by the present LEU conversion. This paper describes the results of core physics measurement through the conversion phase from MEU fuel core to LEU fuel core. Measured excess reactivities of the LEU fuel cores are mostly in good agreement with predicted values. Reactivity effect and burnup of cadmium wires, therefore, were proved to be well predicted. Control rod worth in the LEU fuel core is mostly less than that in the MEU fuel core. Shutdown margin was verified to be within the safety limit. There is no significant difference in temperature coefficient of reactivity between the MEU and LEU fuel cores. These results verified that the JMTR was successfully and safely converted to LEU fuel. Extension of the operating cycle period was achieved and reduction of spend fuel elements is expected by using the fuel with high uranium density.

  10. On the flexibility of high temperature reactor cores for high-and low-enriched fuel

    SciTech Connect

    Bzandes, S.; Lonhert, G.

    1982-07-01

    The operational flexibility of a high temperature reactor (HTR) is not restricted to either a low- or a high-enriched fuel cycle. Both fuel cycles are possible for the same core design. The fuel cycle cost is, however, penalized for low-enriched fuel; in addition, higher uranium consumption is required. Hence, an HTR is most economical to operate in the high-enriched thorium-uranium fuel cycle.

  11. Heat transfer analysis of fuel assemblies in a heterogeneous gas core nuclear rocket

    NASA Technical Reports Server (NTRS)

    Watanabe, Yoichi; Appelbaum, Jacob; Diaz, Nils; Maya, Isaac

    1991-01-01

    Heat transfer problems of a heterogeneous gaseous core nuclear rocket were studied. The reactor core consists of 1.5-m long hexagonal fuel assemblies filled with pressurized uranium tetrafluoride (UF4) gas. The fuel gas temperature ranges from 3500 to 7000 K at a nominal operating condition of 40 atm. Each fuel assembly has seven coolant tubes, through which hydrogen propellant flows. The propellant temperature is not constrained by the fuel temperature but by the maximum temperature of the graphite coolant tube. For a core achieving a fission power density of 1000 MW/cu m, the propellant core exit temperature can be as high as 3200 K. The physical size of a 1250 MW gaseous core nuclear rocket is comparable with that of a NERVA-type solid core nuclear rocket. The engine can deliver a specific impulse of 1020 seconds and a thrust of 330 kN.

  12. Heat transfer analysis of fuel assemblies in a heterogeneous gas core nuclear rocket

    NASA Technical Reports Server (NTRS)

    Watanabe, Yoichi; Appelbaum, Jacob; Diaz, Nils; Maya, Isaac

    1991-01-01

    Heat transfer problems of a heterogeneous gaseous core nuclear rocket were studied. The reactor core consists of 1.5-m long hexagonal fuel assemblies filled with pressurized uranium tetrafluoride (UF4) gas. The fuel gas temperature ranges from 3500 to 7000 K at a nominal operating condition of 40 atm. Each fuel assembly has seven coolant tubes, through which hydrogen propellant flows. The propellant temperature is not constrained by the fuel temperature but by the maximum temperature of the graphite coolant tube. For a core achieving a fission power density of 1000 MW/cu m, the propellant core exit temperature can be as high as 3200 K. The physical size of a 1250 MW gaseous core nuclear rocket is comparable with that of a NERVA-type solid core nuclear rocket. The engine can deliver a specific impulse of 1020 seconds and a thrust of 330 kN.

  13. One-dimensional kinetics modifications for BWR reload methods

    SciTech Connect

    Chandola, V.; Robichaud, J.D.

    1990-01-01

    Yankee Atomic Electric Company (YAEC) currently uses RETRAN-02 to analyze limiting transients and establish operating minimum critical power ratio (MCPR) limits for Vermont Yankee (VY) boiling water reactor (BWR) reload analysis. The US Nuclear Regulatory Commission-approved analysis methods, used in previous cycles, use the point-kinetics modeling option in RETRAN-02 to represent transient-induced neutronic feedback. RETRAN-02 also contains a one-dimensional (1-D) kinetics neutronic feedback model option that provides a more accurate transient power prediction than the point-kinetics model. In the past few fuel cycles, the thermal or MCPR operating margin at VY has eroded due to increases in fuel cycle length. To offset this decrease, YAEC has developed the capability to use the more accurate 1-D kinetics RETRAN option. This paper reviews the qualification effort for the YAEC BWR methods. This paper also presents a comparison between RETRAN-02 predictions using 1-D and point kinetics for the limiting transient, and demonstrates the typical gain in thermal margin from 1-D kinetics.

  14. Neutronic analysis of the conversion of HEU to LEU fuel for a 5-MW MTR core

    SciTech Connect

    Pazirandeh, A.; Bartsch, G.

    1987-01-01

    In recent years, due to cessation of highly enriched uranium (HEU) fuel supply, practical steps have been taken to substitute HEU fuel in almost all research reactors by medium-enriched uranium or low-enriched uranium (LEU) fuels. In this study, a neutronic calculation of a 5-MW research reactor core fueled with HEU (93% /sup 235/U) is presented. In order to assess the performance of the core with the LEU (< 20%) fuel replacement, while keeping fuel element geometry nearly unchanged, several different /sup 235/U loadings were examined. The core consists of 22 standard fuel elements (SFEs) and 6 control fuel elements (CFEs). Each fuel elements has 18 curved plates of which two end plates are dummies. Initial /sup 235/U content is 195 g /sup 235/U/SFE and 9.7 g /sup 235/U/CFE or /PFE. In all calculations the permitted changes to the fuel elements are (a) 18 active plates per SFE, (b) fuel plates assumed to be flat, and (c) 8 or 9 active plates per CFE.

  15. Use of Solid Hydride Fuel for Improved long-Life LWR Core Designs

    SciTech Connect

    Greenspan, E

    2006-04-30

    The primary objective of this project was to assess the feasibility of improving the performance of PWR and BWR cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the bus-bar cost of electricity (COE). Additional performance measures considered are safety, fuel bundle design simplicity – in particular for BWR’s, and plutonium incineration capability. It was found that hydride fuel can safely operate in PWR’s and BWR’s without restricting the linear heat generation rate of these reactors relative to that attainable with oxide fuel. A couple of promising applications of hydride fuel in PWR’s and BWR’s were identified: (1) Eliminating dedicated water moderator volumes in BWR cores thus enabling to significantly increase the cooled fuel rods surface area as well as the coolant flow cross section area in a given volume fuel bundle while significantly reducing the heterogeneity of BWR fuel bundles thus achieving flatter pin-by-pin power distribution. The net result is a possibility to significantly increase the core power density – on the order of 30% and, possibly, more, while greatly simplifying the fuel bundle design. Implementation of the above modifications is, though, not straightforward; it requires a design of completely different control system that could probably be implemented only in newly designed plants. It also requires increasing the coolant pressure drop across the core. (2) Recycling plutonium in PWR’s more effectively than is possible with oxide fuel by virtue of a couple of unique features of hydride fuel – reduced inventory of U-238 and increased inventory of hydrogen. As a result, the hydride fuelled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it incinerates in one pass is double that of the MOX fuel. The fissile fraction of the Pu in the discharged hydride fuel is only ~2/3 that of the MOX fuel and the

  16. A Modified Nitride-Based Fuel for Long Core Life and Proliferation Resistance

    SciTech Connect

    Ebbinghaus, B; Choi, J; Meier, T

    2003-10-22

    A modified nitride-based uranium fuel to support the small, secured, transportable, and autonomous reactor (SSTAR) concept is initiated at Lawrence Livermore National laboratory (LLNL). This project centers on the evaluation of modified uranium nitride fuels imbedded with other inert (e.g. ZrN), neutron-absorbing (e.g. HfN) , or breeding (e.g. ThN) nitrides to enhance the fuel properties to achieve long core life with a compact reactor design. A long-life fuel could minimize the need for on-site refueling and spent-fuel storage. As a result, it could significantly improve the proliferation resistance of the reactor/fuel systems. This paper discusses the potential benefits and detriments of modified nitride-based fuels using the criteria of compactness, long-life, proliferation resistance, fuel safety, and waste management. Benefits and detriments are then considered in recommending a select set of compositions for further study.

  17. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    NASA Astrophysics Data System (ADS)

    Rouf; Su'ud, Zaki

    2016-08-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  18. Assessment of sensitivity of neutron-physical parameters of fast neutron reactor to purification of reprocessed fuel from minor actinides

    NASA Astrophysics Data System (ADS)

    Cherny, V. A.; Kochetkov, L. A.; Nevinitsa, A. I.

    2013-12-01

    The work is devoted to computational investigation of the dependence of basic physical parameters of fast neutron reactors on the degree of purification of plutonium from minor actinides obtained as a result of pyroelectrochemical reprocessing of spent nuclear fuel and used for manufacturing MOX fuel to be reloaded into the reactors mentioned. The investigations have shown that, in order to preserve such important parameters of a BN-800 type reactor as the criticality, the sodium void reactivity effect, the Doppler effect, and the efficiency of safety rods, it is possible to use the reprocessed fuel without separation of minor actinides for refueling (recharging) the core.

  19. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    SciTech Connect

    Bi, G.; Liu, C.; Si, S.

    2012-07-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no

  20. French PWRs: In-core fuel management evolution and future direction

    SciTech Connect

    Rome, M. ); LeBars, M. )

    1993-01-01

    French pressurized water reactor (PWR) fuel management has been carried out until now with the objective of cycle cost reduction. This paper presents the feasibility study of an 18 month cycle for 1300-MW PWRs, three-batch fuel management with 4.00% enrichment, performed by EdF in 1992; the qualifications of core calculation methodology with gadolinia integrated burnable neutron absorbers through the Gedeon critical experiment carried out in the Melusine reactor in Grenoble; and the progress in core calculational methods, which were carried out to determine core safety with the increase in cycle length to 18 months.

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

    NASA Astrophysics Data System (ADS)

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

    2014-09-01

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

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

    SciTech Connect

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

    2014-09-30

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

  3. Fuel Design and Core Layout for a Gas-Cooled Fast Reactor

    SciTech Connect

    Rooijen, W.F.G. van; Kloosterman, J.L.; Hagen, T.H.J.J. van der; Dam, H. van

    2005-09-15

    The gas-cooled fast reactor (GCFR) is regarded as the primary candidate for a future sustainable nuclear power system. In this paper a general core layout is presented for a 2400-MW(thermal) GCFR. Two fuel elements are discussed: a TRISO-based coated particle and the innovative hollow sphere concept. Sustainability calls for recycling of all minor actinides (MAs) in the core and a breeding gain close to unity. A fuel cycle is designed allowing operation over a long period, requiring refueling with {sup 238}U only. The evolution of nuclides in the GCFR core is calculated using the SCALE system (one-dimensional and three-dimensional). Calculations were done over multiple irradiation cycles including reprocessing. The result is that it is possible to design a fuel and GCFR core with a breeding gain around unity, with recycling of all MAs from cycle to cycle. The burnup reactivity swing is small, improving safety. After several fuel batches an equilibrium core is reached. MA loading in the core remains limited, and the fuel temperature coefficient is always negative.

  4. Variant 22: Spatially-Dependent: Transient Processes in MOX Fueled Core

    SciTech Connect

    Pavlovichev, A.M.

    2001-09-28

    This work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactors and presents the results of spatial kinetics calculational benchmarks. The examinations were carried out with the following purposes: to verify one of spatial neutronic kinetics model elaborated in KI, to understand sensibility of the model to neutronics difference of UOX and MOX cores, and to compare in future point and spatial kinetics models (on the base of a set of selected accidents) in view of eventual creation of RELAP option with 3D kinetics. The document contains input data and results of model operation of three emergency dynamic processes in the VVER-1000 core: (1) Central control rod ejection by pressure drop caused by destroying of the moving mechanism cover. (2) Overcooling of the reactor core caused by steam line rupture and non-closure of steam generator stop valve. (3) The boron dilution of coolant in part of the VVER-1000 core caused by penetration of the distillate slug into the core at start up of non-working loop. These accidents have been applied to: (1) Uranium reference core that is the so-called Advanced VVER-1000 core with Zirconium fuel pins claddings and guide tubes. A number of assemblies contained 18 boron BPRs while first year operating. (2) MOX core with about 30% MOX fuel. At a solving it was supposed that MOX-fuel thermophysical characteristics are identical to uranium fuel ones. The calculations were carried out with the help of the program NOSTRA/1/, simulating VVER dynamics that is briefly described in Chapter 1. Chapter 3 contains the description of reference Uranium and MOX cores that are used in calculations. The neutronics calculations of MOX core with about 30% MOX fuel are named ''Variant 2 1''. Chapters 4-6 contain the calculational results of three above mentioned benchmark accidents that compose in a whole the ''Variant 22''.

  5. Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors

    SciTech Connect

    Todosow M.; Todosow M.; Raitses, G. Galperin, A.

    2009-07-12

    Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the

  6. Fuel-rod temperature transients during LWR degraded-core accidents

    SciTech Connect

    Briscoe, F; Rivard, J B; Young, M F

    1982-01-01

    Heat transfer models of fuel rods and coolant have been developed in support of LWR damaged fuel studies underway at Sandia National Laboratories for the NRC. A one-dimensional, full-length model simulates a PWR fuel rod; a two-dimensional, 0.5 m model simulates 9-rod bundle experiments to be performed in the Annular Core Research Reactor. The models include zircaloy oxidation, heat transfer by convecting steam/hydrogen flow, and radiation between surfaces through an absorbing/emitting gas. Characteristics of the one-dimensional reactor fuel rod model for two types of accident sequence are reported, as well as comparisons with MARCH code results.

  7. Contribution of fuel vibrations to ex-core neutron noise during the first and second fuel cycles of the Sequoyah-1 pressurized water reactor

    SciTech Connect

    Sweeney, F.J.; March-Leuba, J.; Smith, C.M.

    1984-01-01

    Noise measurements were performed during the first and second fuel cycles of the Sequoyah-1 pressurized water reactor (PWR) to observe long-term changes in the ex-core neutron signatures. Increases in the ex-core neutron noise amplitude were observed throughout the 0.1- to 50.0-Hz range. In-core noise measurements indicate that fuel assembly vibrations contribute significantly to the ex-core neutron noise at nearly all frequencies in this range, probably due to mechanical or acoustic coupling with other vibrating internal structures. Space-dependent kinetics calculations show that ex-core neutron noise induced by fixed-amplitude fuel assembly vibrations will increase over a fuel cycle because of soluble boron and fuel concentration changes associated with burnup. These reactivity effects can also lead to 180/sup 0/ phase shifts between cross-core detectors. We concluded that it may be difficult to separate the changes in neutron noise due to attenuation (shielding) effects of structural vibrations from changes due to reactivity effects of fuel assembly motion on the basis of neutron noise amplitude or phase information. Amplitudes of core support barrel vibrations inferred from ex-core neutron noise measurements using calculated scale factors are likely to have a high degree of uncertainty, since these scale factors usually do not account for neutron noise generated by fuel assembly vibrations. Modifications in fuel management or design may also lead to altered neutron noise signature behavior over a fuel cycle.

  8. Tailored Core Shell Cathode Powders for Solid Oxide Fuel Cells

    SciTech Connect

    Swartz, Scott

    2015-03-23

    In this Phase I SBIR project, a “core-shell” composite cathode approach was evaluated for improving SOFC performance and reducing degradation of lanthanum strontium cobalt ferrite (LSCF) cathode materials, following previous successful demonstrations of infiltration approaches for achieving the same goals. The intent was to establish core-shell cathode powders that enabled high performance to be obtained with “drop-in” process capability for SOFC manufacturing (i.e., rather than adding an infiltration step to the SOFC manufacturing process). Milling, precipitation and hetero-coagulation methods were evaluated for making core-shell composite cathode powders comprised of coarse LSCF “core” particles and nanoscale “shell” particles of lanthanum strontium manganite (LSM) or praseodymium strontium manganite (PSM). Precipitation and hetero-coagulation methods were successful for obtaining the targeted core-shell morphology, although perfect coverage of the LSCF core particles by the LSM and PSM particles was not obtained. Electrochemical characterization of core-shell cathode powders and conventional (baseline) cathode powders was performed via electrochemical impedance spectroscopy (EIS) half-cell measurements and single-cell SOFC testing. Reliable EIS testing methods were established, which enabled comparative area-specific resistance measurements to be obtained. A single-cell SOFC testing approach also was established that enabled cathode resistance to be separated from overall cell resistance, and for cathode degradation to be separated from overall cell degradation. The results of these EIS and SOFC tests conclusively determined that the core-shell cathode powders resulted in significant lowering of performance, compared to the baseline cathodes. Based on the results of this project, it was concluded that the core-shell cathode approach did not warrant further investigation.

  9. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B.

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  10. Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B.

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (about 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)

  11. Performance of the MTR core with MOX fuel using the MCNP4C2 code.

    PubMed

    Shaaban, Ismail; Albarhoum, Mohamad

    2016-08-01

    The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively.

  12. Integral manifolding structure for fuel cell core having parallel gas flow

    DOEpatents

    Herceg, J.E.

    1983-10-12

    Disclosed herein are manifolding means for directing the fuel and oxidant gases to parallel flow passageways in a fuel cell core. Each core passageway is defined by electrolyte and interconnect walls. Each electrolyte and interconnect wall consists respectively of anode and cathode materials layered on the opposite sides of electrolyte material, or on the opposite sides of interconnect material. A core wall projects beyond the open ends of the defined core passageways and is disposed approximately midway between and parallel to the adjacent overlaying and underlying interconnect walls to define manifold chambers therebetween on opposite sides of the wall. Each electrolyte wall defining the flow passageways is shaped to blend into and be connected to this wall in order to redirect the corresponding fuel and oxidant passageways to the respective manifold chambers either above or below this intermediate wall. Inlet and outlet connections are made to these separate manifold chambers respectively, for carrying the fuel and oxidant gases to the core, and for carrying their reaction products away from the core.

  13. Integral manifolding structure for fuel cell core having parallel gas flow

    DOEpatents

    Herceg, Joseph E.

    1984-01-01

    Disclosed herein are manifolding means for directing the fuel and oxidant gases to parallel flow passageways in a fuel cell core. Each core passageway is defined by electrolyte and interconnect walls. Each electrolyte and interconnect wall consists respectively of anode and cathode materials layered on the opposite sides of electrolyte material, or on the opposite sides of interconnect material. A core wall projects beyond the open ends of the defined core passageways and is disposed approximately midway between and parallel to the adjacent overlaying and underlying interconnect walls to define manifold chambers therebetween on opposite sides of the wall. Each electrolyte wall defining the flow passageways is shaped to blend into and be connected to this wall in order to redirect the corresponding fuel and oxidant passageways to the respective manifold chambers either above or below this intermediate wall. Inlet and outlet connections are made to these separate manifold chambers respectively, for carrying the fuel and oxidant gases to the core, and for carrying their reaction products away from the core.

  14. Experience in PWR and BWR mixed-oxide fuel management

    SciTech Connect

    Schlosser, G.J.; Krebs, W.; Urban, P. )

    1993-04-01

    Germany has adopted the strategy of a closed fuel cycle using reprocessing and recycling. The central issue today is plutonium recycling by the use of U-Pu mixed oxide (MOX) in pressurized water reactors (PWRs) and boiling water reactors (BWRs). The design of MOX fuel assemblies and fuel management in MOX-containing cores are strongly influenced by the nuclear properties of the plutonium isotopes. Optimized MOX fuel assembly designs for PWRs currently use up to three types of MOX fuel rods having different plutonium contents with natural uranium or uranium tailings as carrier material but without burnable absorbers. The MOX fuel assembly designs for BWRs use four to six rod types with different plutonium contents and Gd[sub 2]O[sub 3]/UO[sub 2] burnable absorber rods. Both the PWR and the BWR designs attain good burnup equivalence and compatibility with uranium fuel assemblies. High flexibility exists in the loading schemes relative to the position and number of MOX fuel assemblies in the reloads and in the core as a whole. The Siemens experience with MOX fuel assemblies is based on the insertion of 318 MOX fuel assemblies in eight PWRs and 168 in BWRs and pressurized heavy water reactors so far. The primary operating results include information on the cycle length, power distribution, reactivity coefficients, and control rod worth of cores containing MOX fuel assemblies.

  15. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    SciTech Connect

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  16. Spatial Kinetics Calculations of MOX Fueled Core: Variant 22

    SciTech Connect

    Pavlovichev, A.M.

    2001-01-11

    This work is part of a Joint US/Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactors and presents the results of spatial kinetics calculational benchmarks. The examinations were carried out with the following purposes: to verify one of spatial neutronic kinetics model elaborated in KI, to understand sensibility of the model to neutronics difference of UOX and MOX cores, to compare in future point and spatial kinetics models (on the base of a set of selected accidents) in view of eventual creation of RELAP option with 3D kinetics. The document contains input data and results of model operation of three emergency dynamic processes in the VVER-1000 core: central control rod ejection by pressure drop caused by destroying of the moving mechanism cover; overcooling of the reactor core caused by steam line rupture and non-closure of steam generator stop valve; and the boron dilution of coolant in part of the VVER-1000 core caused by penetration of the distillate slug into the core at start up of non-working loop.

  17. Solid oxide fuel cell having monolithic cross flow core and manifolding

    DOEpatents

    Poeppel, Roger B.; Dusek, Joseph T.

    1984-01-01

    This invention discloses a monolithic core construction having the flow passageways for the fuel and for the oxidant gases extended transverse to one another, whereby full face core manifolding can be achieved for these gases and their reaction products. The core construction provides that only anode material surround each fuel passageway and only cathode material surround each oxidant passageway, each anode and each cathode further sandwiching at spaced opposing sides electrolyte and interconnect materials to define electrolyte and interconnect walls. Webs of the cathode and anode material hold the electrolyte and interconnect walls spaced apart to define the flow passages. The composite anode and cathode wall structures are further alternately stacked on one another (with the separating electrolyte or interconnect material typically being a single common layer) whereby the fuel passageway and the oxidant passageways are disposed transverse to one another.

  18. Solid oxide fuel cell having monolithic cross flow core and manifolding

    DOEpatents

    Poeppel, R.B.; Dusek, J.T.

    1983-10-12

    This invention discloses a monolithic core construction having the flow passageways for the fuel and for the oxidant gases extended transverse to one another, whereby full face core manifolding can be achieved for these gases and their reaction products. The core construction provides that only anode material surround each fuel passageway and only cathode material surround each oxidant passageway, each anode and each cathode further sandwiching at spaced opposing sides electrolyte and interconnect materials to define electrolyte and interconnect walls. Webs of the cathode and anode material hold the electrolyte and interconnect walls spaced apart to define the flow passages. The composite anode and cathode wall structures are further alternately stacked on one another (with the separating electrolyte or interconnect material typically being a single common layer) whereby the fuel passageways and the oxidant passageways are disposed transverse to one another.

  19. Analysis of in-core coolant temperatures of FFTF instrumented fuels tests at full power

    SciTech Connect

    Hoth, C.W

    1981-01-01

    Two full size highly instrumented fuel assemblies were inserted into the core of the Fast Flux Test Facility in December of 1979. The major objectives of these instrumented tests are to provide verification of the FFTF core conditions and to characterize temperature patterns within FFTF driver fuel assemblies. A review is presented of the results obtained during the power ascents and during irradiation at a constant reactor power of 400 MWt. The results obtained from these instrumented tests verify the conservative nature of the design methods used to establish core conditions in FFTF. The success of these tests also demonstrates the ability to design, fabricate, install and irradiate complex, instrumented fuel tests in FFTF using commercially procured components.

  20. Evaluation of storing Shippingport Core II spent blanket fuel assemblies in the T Plant PWR Core II fuel pool without active cooling

    SciTech Connect

    Gilbert, E.R.; Lanning, D.D.; Dana, C.M.; Hedengren, D.C.

    1994-10-01

    PWR Core II fuel pool chiller-off test was conducted because it appeared possible that acceptable pool-water temperatures could be maintained without operating the chillers, thus saving hundreds of thousands of dollars in maintenance and replacement costs. Test results showed that the water-cooling capability is no longer needed to maintain pool temperature below 38{degrees}C (100{degrees}F).

  1. Core-shell Au-Pd nanoparticles as cathode catalysts for microbial fuel cell applications

    NASA Astrophysics Data System (ADS)

    Yang, Gaixiu; Chen, Dong; Lv, Pengmei; Kong, Xiaoying; Sun, Yongming; Wang, Zhongming; Yuan, Zhenhong; Liu, Hui; Yang, Jun

    2016-10-01

    Bimetallic nanoparticles with core-shell structures usually display enhanced catalytic properties due to the lattice strain created between the core and shell regions. In this study, we demonstrate the application of bimetallic Au-Pd nanoparticles with an Au core and a thin Pd shell as cathode catalysts in microbial fuel cells, which represent a promising technology for wastewater treatment, while directly generating electrical energy. In specific, in comparison with the hollow structured Pt nanoparticles, a benchmark for the electrocatalysis, the bimetallic core-shell Au-Pd nanoparticles are found to have superior activity and stability for oxygen reduction reaction in a neutral condition due to the strong electronic interaction and lattice strain effect between the Au core and the Pd shell domains. The maximum power density generated in a membraneless single-chamber microbial fuel cell running on wastewater with core-shell Au-Pd as cathode catalysts is ca. 16.0 W m-3 and remains stable over 150 days, clearly illustrating the potential of core-shell nanostructures in the applications of microbial fuel cells.

  2. Core-shell Au-Pd nanoparticles as cathode catalysts for microbial fuel cell applications

    PubMed Central

    Yang, Gaixiu; Chen, Dong; Lv, Pengmei; Kong, Xiaoying; Sun, Yongming; Wang, Zhongming; Yuan, Zhenhong; Liu, Hui; Yang, Jun

    2016-01-01

    Bimetallic nanoparticles with core-shell structures usually display enhanced catalytic properties due to the lattice strain created between the core and shell regions. In this study, we demonstrate the application of bimetallic Au-Pd nanoparticles with an Au core and a thin Pd shell as cathode catalysts in microbial fuel cells, which represent a promising technology for wastewater treatment, while directly generating electrical energy. In specific, in comparison with the hollow structured Pt nanoparticles, a benchmark for the electrocatalysis, the bimetallic core-shell Au-Pd nanoparticles are found to have superior activity and stability for oxygen reduction reaction in a neutral condition due to the strong electronic interaction and lattice strain effect between the Au core and the Pd shell domains. The maximum power density generated in a membraneless single-chamber microbial fuel cell running on wastewater with core-shell Au-Pd as cathode catalysts is ca. 16.0 W m−3 and remains stable over 150 days, clearly illustrating the potential of core-shell nanostructures in the applications of microbial fuel cells. PMID:27734945

  3. Core-shell Au-Pd nanoparticles as cathode catalysts for microbial fuel cell applications.

    PubMed

    Yang, Gaixiu; Chen, Dong; Lv, Pengmei; Kong, Xiaoying; Sun, Yongming; Wang, Zhongming; Yuan, Zhenhong; Liu, Hui; Yang, Jun

    2016-10-13

    Bimetallic nanoparticles with core-shell structures usually display enhanced catalytic properties due to the lattice strain created between the core and shell regions. In this study, we demonstrate the application of bimetallic Au-Pd nanoparticles with an Au core and a thin Pd shell as cathode catalysts in microbial fuel cells, which represent a promising technology for wastewater treatment, while directly generating electrical energy. In specific, in comparison with the hollow structured Pt nanoparticles, a benchmark for the electrocatalysis, the bimetallic core-shell Au-Pd nanoparticles are found to have superior activity and stability for oxygen reduction reaction in a neutral condition due to the strong electronic interaction and lattice strain effect between the Au core and the Pd shell domains. The maximum power density generated in a membraneless single-chamber microbial fuel cell running on wastewater with core-shell Au-Pd as cathode catalysts is ca. 16.0 W m(-3) and remains stable over 150 days, clearly illustrating the potential of core-shell nanostructures in the applications of microbial fuel cells.

  4. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    DOEpatents

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  5. Method of fabricating a monolithic core for a solid oxide fuel cell

    NASA Astrophysics Data System (ADS)

    Zwick, S. A.; Ackerman, J. P.

    1983-10-01

    A method is disclosed for forming a core for use in a solid oxide fuel cell that electrochemically combines fuel and oxidant for generating galvanic output. The core has an array of electrolyte and interconnect walls that are substantially devoid of any composite inert materials for support consisting instead only of the active anode, cathode, electrolyte and interconnect materials. Each electrolyte wall consists of cathode and anode materials sandwiching electrolyte material therebetween, and each interconnect wall consists of the cathode and anode materials sandwiching interconnect material therebetween. The method consists of building up the electrolyte and interconnect walls by depositing each material on individually and endwise of the wall itself, where each material deposit is sequentially applied for one cycle; and where the depositing cycle is repeated many times until the material buildup is sufficient to formulate the core. The core is heat cured to become dimensionally and structurally stable.

  6. Fuel performance models for high-temperature gas-cooled reactor core design

    SciTech Connect

    Stansfield, O.M.; Simon, W.A.; Baxter, A.M.

    1983-09-01

    Mechanistic fuel performance models are used in high-temperature gas-cooled reactor core design and licensing to predict failure and fission product release. Fuel particles manufactured with defective or missing SiC, IPyC, or fuel dispersion in the buffer fail at a level of less than 5 x 10/sup -4/ fraction. These failed particles primarily release metallic fission products because the OPyC remains intact on 90% of the particles and retains gaseous isotopes. The predicted failure of particles using performance models appears to be conservative relative to operating reactor experience.

  7. Optimal reload strategies for identify-and-destroy missions

    NASA Astrophysics Data System (ADS)

    Hyland, John C.; Smith, Cheryl M.

    2004-09-01

    In this problem an identification vehicle must re-acquire a fixed set of suspected targets and determine whether each suspected target is a mine or a false alarm. If a target is determined to be a mine, the identification vehicle must neutralize it by either delivering one of a limited number of on-board bombs or by assigning the neutralization task to one of a limited number of single-shot suicide vehicles. The identification vehicle has the option to reload. The singleshot suicide vehicles, however, cannot be replenished. We have developed an optimal path planning and reload strategy for this identify and destroy mission that takes into account the probabilities that suspected targets are mines, the costs to move between targets, the costs to return to and from the reload point, and the cost to reload. The mission is modeled as a discrete multi-dimensional Markov process. At each target position the vehicle decides based on the known costs, probabilities, the number of bombs on board (r), and the number of remaining one-shot vehicles (s) whether to move directly on to the next target or to reload before continuing and whether to destroy any mine with an on-board bomb or a one-shot suicide vehicle. The approach recursively calculates the minimum expected overall cost conditioned on all possible values r and s. The recursion is similar to dynamic programming in that it starts at the last suspected target location and works its way backwards to the starting point. The approach also uses a suboptimal traveling salesman strategy to search over candidate deployment locations to calculate the best initial deployment point where the reloads will take place.

  8. Whole-core neutron transport calculations without fuel-coolant homogenization

    SciTech Connect

    Smith, M. A.; Tsoulfanidis, N.; Lewis, E. E.; Palmiotti, G.; Taiwo, T. A.

    2000-02-10

    The variational nodal method implemented in the VARIANT code is generalized to perform full core transport calculations without spatial homogenization of cross sections at either the fuel-pin cell or fuel assembly level. The node size is chosen to correspond to one fuel-pin cell in the radial plane. Each node is divided into triangular finite subelements, with the interior spatial flux distribution represented by piecewise linear trial functions. The step change in the cross sections at the fuel-coolant interface can thus be represented explicitly in global calculations while retaining the fill spherical harmonics capability of VARIANT. The resulting method is applied to a two-dimensional seven-group representation of a LWR containing MOX fuel assemblies. Comparisons are made of the accuracy of various space-angle approximations and of the corresponding CPU times.

  9. Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Fourth semiannual report, July-December 1980. [BWR

    SciTech Connect

    Rosenbaum, H.S.

    1981-03-01

    This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts have been developed for possible demonstration: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the scope of this program one of these concepts had to be selected for a large-scale demonstration in a commercial power reactor. The selection was made to demonstrate Zr-liner fuel and to include bundles which have liners prepared from either low oxygen sponge zirconium or of crystal bar zirconium. The demonstration is intended to include a total of 132 barrier bundles in the reload for Quad Cities Unit 2, Cycle 6. In the current report period changes in the nuclear design were made to respond to changes in the Energy Utilization Plan for Quad Cities Unit 2. Bundle designs were completed, and were licensed for use in a BWR/3. The core specific licensing will be done as part of the reload license for Quad Cities Unit 2, Cycle 6.

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

    NASA Astrophysics Data System (ADS)

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

    2017-01-01

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

  11. Core-Protected Platinum Monolayer Shell High-Stability Electrocatalysts for Fuel-Cell Cathodes

    SciTech Connect

    K Sasaki; H Naohara; Y Cai; Y Choi; P Liu; M Vukmirovic; J Wang; R Adzic

    2011-12-31

    Platinum monolayers can act as shells for palladium nanoparticles to lead to electrocatalysts with high activities and an ultralow platinum content, but high platinum utilization. The stability derives from the core protecting the shell from dissolution. In fuel-cell tests, no loss of platinum was observed in 200,000 potential cycles, whereas loss of palladium was significant.

  12. Core-Protected Platinum Monolayer Shell High-Stability Electrocatalysts for Fuel-Cell Cathodes

    SciTech Connect

    Adzic, R.R.; Sasaki, K.; Naohara, H.; Cai, Y.; Choi, Y.M.; Liu, P.; Vukmirovic, M.B.; Wang, J.X.

    2010-11-08

    More than skin deep: Platinum monolayers can act as shells for palladium nanoparticles to lead to electrocatalysts with high activities and an ultralow platinum content, but high platinum utilization. The stability derives from the core protecting the shell from dissolution. In fuel-cell tests, no loss of platinum was observed in 200?000 potential cycles, whereas loss of palladium was significant.

  13. Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

    NASA Astrophysics Data System (ADS)

    Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

    2017-09-01

    The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

  14. Oyster Creek fuel thermal margin during core thermal-hydraulic oscillations

    SciTech Connect

    Dougher, J.D.

    1990-01-01

    The Oyster Creek nuclear facility, a boiling water reactor (BWR)-2 plant type, has never experienced core thermal-hydraulic instability. Power oscillations, however, have been observed in other BWR cores both domestically and internationally. Two modes of oscillations have been observed, core wide and regional half-core. During core wide oscillations, the neutron flux in the core oscillates in the radial fundamental mode. During regional half-core oscillations, higher order harmonics in the radial plane result in out-of-phase oscillations with the neutron flux in one half of the core oscillating 180 deg out-of-phase with the neutron flux in the other half of the core. General Design Criteria 12 requires either prevention or detection and suppression of power oscillations which could result in violations of fuel design limits. Analyses performed by General Electric have demonstrated that for large-magnitude oscillations the potential exists for violation of the safety limit minimum critical power ratio (MCPR). However, for plants with a flow-biased neutron flux scram automatic mitigation of oscillations may be provided at an oscillation magnitude below that at which the safety limit is challenged. Plant-specific analysis for Oyster Creek demonstrates that the existing average power range monitor (APRM) system will sense and suppress power oscillations prior to violation of any safety limits.

  15. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    SciTech Connect

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  16. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    SciTech Connect

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  17. In-Core Fuel Management with Biased Multiobjective Function Optimization

    SciTech Connect

    Shatilla, Youssef A.; Little, David C.; Penkrot, Jack A.; Holland, Richard Andrew

    2000-06-15

    The capability of biased multiobjective function optimization has been added to the Westinghouse Electric Company's (Westinghouse's) Advanced Loading Pattern Search code (ALPS). The search process, given a user-defined set of design constraints, proceeds to minimize a global parameter called the total value associated with constraints compliance (VACC), an importance-weighted measure of the deviation from limit and/or margin target. The search process takes into consideration two equally important user-defined factors while minimizing the VACC, namely, the relative importance of each constraint with respect to the others and the optimization of each constraint according to its own objective function. Hence, trading off margin-to-design limits from where it is abundantly available to where it is badly needed can now be accomplished. Two practical methods are provided to the user for input of constraints and associated objective functions. One consists of establishing design limits based on traditional core design parameters such as assembly/pin burnup, power, or reactivity. The second method allows the user to write a program, or script, to define a logic not possible through ordinary means. This method of script writing was made possible through the application resident compiler feature of the technical user language integration processor (tulip), developed at Westinghouse. For the optimization problems studied, ALPS not only produced candidate loading patterns (LPs) that met all of the conflicting design constraints, but in cases where the design appeared to be over constrained gave a wide range of LPs that came very close to meeting all the constraints based on the associated objective functions.

  18. Spent Fuel Test-Climax: core logging for site investigation and instrumentation

    SciTech Connect

    Wilder, D.G.; Yow, J.L. Jr.; Thorpe, R.K.

    1982-05-28

    As an integral part of the Spent Fuel Test-Climax 5150 ft (1570 m) of granite core was obtained. This core was diamond drilled in various sizes, mainly 38-mm and 76-mm diameters. The core was teken with single tube core barrels and was unoriented. Techniques used to drill and log this core are discussed, as well as techniques to orient the core. Of the 5150 ft (1570 m) of core more than 3645 ft (1111 m) was retained and logged in some detail. As a result of the core logging, geologic discontinuities were identified, joint frequency and spacing characterized. Discontinuities identified included several joint sets, shear zones and faults. Correlations based on coring along were generally found to be impossible, even for the more prominent features. The only feature properly correlated from the exploratory drilling was the fault system at the end of the facility, but it was not identified from the exploratory core as a fault. Identification of discontinuities was later helped by underground mapping that identified several different joint sets with different characteristics. It was found that joint frequency varied from 0.3 to 1.1 joint per foot of core for open fractures and from 0.3 to 3.3/ft for closed or healed fractures. Histograms of fracture spacing indicate that there is likely a random distribution of spacing superimposed upon uniformly spaced fractures. It was found that a low angle joint set had a persistent mean orientation. These joints were healed and had pervasive wall rock alteration which made identification of joints in this set possible. The recognition of a joint set with known attitude allowed orientation of much of the core. This orientation technique was found to be effective. 10 references, 25 figures, 4 tables.

  19. Status of core conversion with LEU silicide fuel in JRR-4

    SciTech Connect

    Nakajima, Teruo; Ohnishi, Nobuaki; Shirai, Eiji

    1997-08-01

    Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel used and swimming pool type reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965, JRR-4 has been used for shielding experiments, radioisotope production, neutron activation analyses, training for reactor engineers and so on for about 30 years. Within the framework of the RERTR Program, the works for conversion to LEU fuel are now under way, and neutronic and thermal-hydraulic calculations emphasizing on safety and performance aspects are being carried out. The design and evaluation for the core conversion are based on the Guides for Safety Design and Evaluation of research and testing reactor facilities in Japan. These results show that the JRR-4 will be able to convert to use LEU fuel without any major design change of core and size of fuel element. LEU silicide fuel (19.75%) will be used and maximum neutron flux in irradiation hole would be slightly decreased from present neutron flux value of 7x10{sup 13}(n/cm{sup 2}/s). The conversion works are scheduled to complete in 1998, including with upgrade of the reactor building and utilization facilities.

  20. Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core

    SciTech Connect

    Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

    1985-01-01

    The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO/sub 2/ and UO/sub 2//metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO/sub 2/ crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process.

  1. Coupling the core analysis program DeCART to the fuel performance application BISON

    SciTech Connect

    Gleicher, F. N.; Spencer, B.; Novascone, S.; Williamson, R.; Martineau, R. C.; Rose, M.; Downar, T. J.; Collins, B.

    2013-07-01

    The 3D neutron transport and core analysis program DeCART was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the method of characteristics) to a high fidelity fuel performance program, both of which can simulate 3D problems. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during burnup or a fast transient. BISON implicitly solves coupled thermomechanical equations for the fuel on a sub-millimeter level finite element mesh. A method was developed for mapping the fission rate density and fast neutron flux from DeCART to BISON. Multiple depletion cases were run with one-way data transfer from DeCART to BISON. The one-way data transfer of fission rate density is shown to agree with the fission rate density obtained from an internal Lassman-style model in BISON. One-way data transfer was also demonstrated in a 3D case in which azimuthal asymmetry was induced in the fission rate density profile of a fuel rod modeled in DeCART. Two-way data transfer was established by mapping the temperature distribution from BISON to DeCART. A Picard iterative algorithm was developed for the loose coupling with two-way data transfer. (authors)

  2. Start-up fuel and power flattening of sodium-cooled candle core

    SciTech Connect

    Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake; Sekimoto, Hiroshi

    2013-07-01

    The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing.

  3. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  4. Fission product release phenomena during core melt accidents in metal fueled heavy water reactors

    SciTech Connect

    Ellison, P G; Hyder, M L; Monson, P R; Randolph, H W; Hagrman, D L; McClure, P R; Leonard, M T

    1990-01-01

    The phenomena that determine fission product release rates from a core melting accident in a metal-fueled, heavy water reactor are described in this paper. This information is obtained from the analysis of the current metal fuel experimental data base and from the results of analytical calculations. Experimental programs in place at the Savannah River Site are described that will provide information to resolve uncertainties in the data base. The results of the experiments will be incorporated into new severe accident computer codes recently developed for this reactor design. 47 refs., 4 figs.

  5. Method of fabricating a monolithic core for a solid oxide fuel cell

    DOEpatents

    Zwick, Stanley A.; Ackerman, John P.

    1985-01-01

    A method is disclosed for forming a core for use in a solid oxide fuel cell that electrochemically combines fuel and oxidant for generating galvanic output. The core has an array of electrolyte and interconnect walls that are substantially devoid of any composite inert materials for support consisting instead only of the active anode, cathode, electrolyte and interconnect materials. Each electrolyte wall consists of cathode and anode materials sandwiching electrolyte material therebetween, and each interconnect wall consists of the cathode and anode materials sandwiching interconnect material therebetween. The electrolyte and interconnect walls define a plurality of substantially parallel core passageways alternately having respectively the inside faces thereof with only the anode material or with only the cathode material exposed. In the wall structure, the electrolyte and interconnect materials are only 0.002-0.01 cm thick; and the cathode and anode materials are only 0.002-0.05 cm thick. The method consists of building up the electrolyte and interconnect walls by depositing each material on individually and endwise of the wall itself, where each material deposit is sequentially applied for one cycle; and where the depositing cycle is repeated many times until the material buildup is sufficient to formulate the core. The core is heat cured to become dimensionally and structurally stable.

  6. Fuel and Core Design for Long Operating Cycle Simplified BWR (LSBWR)

    SciTech Connect

    Noriyuki Yoshida; Kouji Hiraiwa; Mikihide Nakamaru; Hideaki Heki

    2002-07-01

    This paper describes an innovative core concept currently being developed for long operating cycle simplified BWR (LSBWR). The LSBWR adopts the long cycle operation (15 years) for the elimination of the fuel pool and the refueling machines and for the capacity usage ratio improvement. To achieve long cycle operation, a combination of enriched gadolinium and 0.7- times sized small bundle with peripheral-positioned gadolinium rod is adopted as a key design concept. A nuclear design for fuel bundle has been determined based on three dimensional nuclear and thermal hydraulic calculation. A core performance has been evaluated based on this bundle design and thermal performance and reactivity characteristics indicated preferable value. (authors)

  7. Accelerator-driven subcritical fission in molten salt core: Closing the nuclear fuel cycle for green nuclear energy

    SciTech Connect

    McIntyre, Peter; Assadi, Saeed; Badgley, Karie; Baker, William; Comeaux, Justin; Gerity, James; Kellams, Joshua; McInturff, Al; Pogue, Nathaniel; Sattarov, Akhdiyor; Sooby, Elizabeth; Tsvetkov, Pavel; Phongikaroon, Supathorn; Simpson, Michael

    2013-04-19

    A technology for accelerator-driven subcritical fission in a molten salt core (ADSMS) is being developed as a basis for the destruction of the transuranics in used nuclear fuel. The molten salt fuel is a eutectic mixture of NaCl and the chlorides of the transuranics and fission products. The core is driven by proton beams from a strong-focusing cyclotron stack. This approach uniquely provides an intrinsically safe means to drive a core fueled only with transuranics, thereby eliminating competing breeding terms.

  8. Exhaust nozzle control and core engine fuel control for turbofan engine

    SciTech Connect

    Camp, V.T.

    1981-10-13

    This control serves to optimize thrust during steady state and transient operation modes of a turbofan engine of the mixed flow type by adjusting or trimming the exhaust nozzle area as a function of fan pressure ratio and fan rotor speed and by adjusting or trimming the core engine fuel flow as a function of fan rotor speed and/or turbine inlet temperature. The control serves to enhance stability by assuring airflow in the engine and its inlet is within a given value avoiding inlet buzz and high distortion to the engine and avoiding even transient operation in conditions that might cause compressor flow instability or stall. Fuel flow is adjusted or trimmed as a function of fan rotor speed or turbine inlet temperature limits depending on which is calling for the least amount of fuel.

  9. Performance of metal and oxide fuel cores during accidents in large liquid-metal-cooled reactors

    SciTech Connect

    Royl, P.H.; Kussmaul, G. ); Cahalan, J.E.; Wigeland, R.A. ); Friedel, G. ); Moreau, J. ); Perks, M. )

    1992-02-01

    This paper reports on a cooperative effort among European and U.S. analysts, which is an assessment of the comparative safety performance of metal and oxide fuels during accidents in a 3500-MW (thermal), pool-type, liquid-metal-cooled reactor (LMR) is performed. The study focuses on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower, and the unprotected loss-of-heat-sink (ULOHS). Core designs with a similar power output that have been previously analyzed in Europe under ULOF accident conditions are also included in this comparison. Emphasis is placed on identification of design features that provide passive, self-limiting responses to postulated accident 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 do oxide-fueled reactors of the same design.

  10. Core Fueling and Edge Particle Flux Analysis in Ohmically and Auxiliary Heated NSTX Plasmas

    SciTech Connect

    V.A. Soukhanovskii; R. Maingi; R. Raman; H.W. Kugel; B.P. LeBlanc; L. Roquemore; C.H. Skinner; NSTX Research Team

    2002-06-12

    The Boundary Physics program of the National Spherical Torus Experiment (NSTX) is focusing on optimization of the edge power and particle flows in b * 25% L- and H-mode plasmas of t {approx} 0.8 s duration heated by up to 6 MW of high harmonic fast wave and up to 5 MW of neutral beam injection. Particle balance and core fueling efficiencies of low and high field side gas fueling of L-mode homic and NBI heated plasmas have been compared using an analytical zero dimensional particle balance model and measured ion and neutral fluxes. Gas fueling efficiencies are in the range of 0.05-0.20 and do not depend on discharge magnetic configuration, density or poloidal location of the injector. The particle balance modeling indicates that the addition of HFS fueling results in a reversal of the wall loading rate and higher wall inventories. Initial particle source estimates obtained from neutral pressure and spectroscopic measurements indicate that ion flux into the divertor greatly exceeds midplane ion flux from the main plasma, suggesting that the scrape-off cross-field transport plays a minor role in diverted plasmas. Present analysis provides the basis for detailed fluid modeling of core and edge particle flows and particle confinement properties of NSTX plasmas. This research was supported by the U.S. Department of Energy under contracts No. DE-AC02-76CH03073, DE-AC05-00OR22725, and W-7405-ENG-36.

  11. A new technique for effective core fueling and density control in RFX-mod

    NASA Astrophysics Data System (ADS)

    de Masi, Gianluca; Auriemma, Fulvio; Cavazzana, Roberto; Martines, Emilio; Spizzo, Gianluca

    2014-10-01

    High current plasmas in the RFX-mod Reversed Field Pinch device can be presently sustained either operating at low density (ne/nG < 0.3, being nG the Greenwald density) or transiently at high density by pellet injection. Discharges at ne/nG > 0.3 are difficult to sustain due to the high ohmic power required and a confinement properties downgrading. In these regimes, the transport mechanism results in a hollow density profile preventing an effective core fueling. A different behavior is observed in Ultra-low q configuration (q[r = a] > 0), in which the increased particle diffusivity produces flat density profiles and makes easier neutral particle penetration. In this contribution we show the main results of a new method to produce a more effective core fueling based on the previous empirical observations. The idea was to produce during the discharges narrow time windows with q[r = a] >= 0 values and, during this phase, to apply a strong gas puffing. This experimental condition is found to allow an increased particles core penetration. From the operational point of view, a lower input power was needed to sustain the discharges with similar core density. A deeper analysis through the ASTRA code will highlight the relation between transport properties and magnetic topology.

  12. Multicycle PWR in-core fuel management through one-and-a-half-dimensional core modeling

    SciTech Connect

    Petrovic, B.G.; Levine, S.H. )

    1992-01-01

    The one-and-a-half-dimensional (1 and 1/2-D) model of a pressurized water reactor (PWR) core employs one-dimensional (1-D) diffusion calculation followed by a fast, few-step procedure to unfold the 1-D results into the two-dimensional (2-D) results. A computer code was developed based on that model. The initial benchmarking has shown the code to be almost as fast as a plain 1-D code and significantly faster than the analogous 2-D code (10 to 100 times for a typical problem). Yet, it provides results in 2-D form and better accuracy than the 1-D code. The model itself and the initial benchmarking were described in more detail elsewhere. The code has since been enhanced, and a multicycle analysis option has been implemented. This paper presents results of benchmarking the model using the actual data for three successive cycles of the Krsko nuclear power plant and complex low-leakage loading patterns.

  13. Enhancement of the inherent self-protection of the fast sodium reactor cores with oxide fuel

    SciTech Connect

    Eliseev, V.A.; Malisheva, I.V.; Matveev, V.I.; Egorov, A.V.; Maslov, P.A.

    2013-07-01

    With the development and research into the generation IV fast sodium reactors, great attention is paid to the enhancement of the core inherent self-protection characteristics. One of the problems dealt here is connected with the reduction of the reactivity margin so that the control rods running should not result in the core overheating and melting. In this paper we consider the possibilities of improving the core of BN-1200 with oxide fuel by a known method of introducing an axial fertile layer into the core. But unlike earlier studies this paper looks at the possibility of using such a layer not only for improving breeding, but also for reducing sodium void reactivity effect (SVRE). This proposed improvement of the BN-1200 core does not solve the problem of strong interference in control and protection system (CPS) rods of BN-1200, but they reduce significantly the reactivity margin for burn-up compensation. This helps compensate all the reactivity balances in the improved core configurations without violating constraints on SVRE value.

  14. An extended conventional fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

    SciTech Connect

    Scarangella, M. J.

    2012-07-01

    The B and W mPower{sup TM} reactor is a small pressurized water reactor (PWR) with an integral once-through steam generator and a thermal output of about 500 MW; it is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height PWR assemblies with the familiar 17 x 17 fuel rod array. The Babcock and Wilcox Company (B and W) is offering a core loading and cycle management plan for a four-year cycle based on its presumed attractiveness to potential customers. This option is a once-through fuel cycle in which the entire core is discharged and replaced after four years. In addition, a conventional fuel utilization strategy, employing a periodic partial reload and shuffle, was developed as an alternative to the four-year once-through fuel cycle. This study, which was performed using the Studsvik core design code suite, is a typical multi-cycle projection analysis of the type performed by most fuel management organizations such as fuel vendors and utilities. In the industry, the results of such projections are used by the financial arms of these organizations to assist in making long-term decisions. In the case of the B and W mPower reactor, this analysis demonstrates flexibility for customers who consider the once-through fuel cycle unacceptable from a fuel utilization standpoint. As expected, when compared to the once-through concept, reloads of the B and W mPower reactor will achieve higher batch average discharge exposure, will have adequate shut-down margin, and will have a relatively flat hot excess reactivity trend at the expense of slightly increased peaking. (authors)

  15. PWR core and spent fuel pool analysis using scale and nestle

    SciTech Connect

    Murphy, J. E.; Maldonado, G. I.; St Clair, R.; Orr, D.

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  16. A Methodology for Selecting High Thermal-Hydraulic Performance Fuel Configurations for Tightly Packed Epithermal Core Designs

    SciTech Connect

    Romano, Antonino; Todreas, Neil E.

    2002-07-15

    Cylindrical fuel pins with wires are the design of choice for tightly packed fuel arrays. However, it is important to investigate novel fuel configurations in order to increase the thermal margins. Hence, new fuel designs have been studied for the epithermal option of the light water-cooled IRIS core. These designs are also of potential use in other tightly packed, epithermal advanced core designs.First, design equations have been used to determine number, height, and size of the principal features (clad, gap, fuel cross-sectional area) of the novel fuel configurations under investigation. Then, performance indices have been introduced to relate fuel geometrical characteristics to selected thermal-hydraulic parameters, such as pressure drop, critical heat flux (CHF), fuel centerline temperature, and clad surface temperature and stress distribution. Finally, variously shaped fuel configurations, including cylindrical, triangular, square, and hexagonal, have been ranked according to the performance indicators.The hexagonal fuel pins, both twisted and straight, proved to be good solutions for the epithermal tight core of the light water-cooled IRIS reactor, with performances comparable to those of the cylindrical fuel with wires. In particular, for water-to-fuel ratios {approx}0.33, the twisted hexagonal shape is the preferable design with a reduction of the total pressure drop by 16% and an increase of the CHF margin by 200%, compared to the traditional cylindrical pins with grids. Furthermore, the straight hexagonal shape allows flatter subchannel velocity profiles, wall shear stress, and wall temperature distributions. However, geometric constraints unfortunately do not allow application of the twisted hexagonal shape for smaller water-to-fuel ratios, which is a design regime of more favorable epithermal neutronics performance. In this regime, the cylindrical pins with wires are the solution of choice.

  17. Core materials development for the fuel cycle R&D program

    NASA Astrophysics Data System (ADS)

    Maloy, S. A.; Toloczko, M.; Cole, J.; Byun, T. S.

    2011-08-01

    The Fuel Cycle Research and Development program is investigating methods of burning minor actinides in a transmutation fuel. One of the challenges of achieving this goal is to develop fuels capable of reaching extreme burnup levels (e.g. 40%). To achieve such high burnup levels' fast reactor core materials (cladding and duct) must be able to withstand very high doses (>300 dpa design goal) while in contact with the coolant and the fuel. Thus, these materials must withstand radiation effects that promote low temperature embrittlement, radiation induced segregation, high temperature helium embrittlement, swelling, accelerated creep, corrosion with the coolant, and chemical interaction with the fuel (FCCI). To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Test specimens of ferritic/martensitic alloys (T91/HT-9) previously irradiated in the FFTF reactor up to 210 dpa at a temperature range of 350-750 °C are presently being tested. This includes analysis of a duct made of HT-9 after irradiation to a total dose of 155 dpa at temperatures from 370 to 510 °C. Compact tension, charpy and tensile specimens have been machined from this duct and mechanical testing as well as SANS and Mossbauer spectroscopy are currently being performed. Initial results from compression testing and Charpy testing reveal a strong increase in yield stress (˜400 MPa) and a large increase in DBTT (up to 230 °C) for specimens irradiated at 383 °C to a dose of 28 dpa. Less hardening and a smaller increase in DBTT was observed for specimens irradiated at higher temperatures up to 500 °C. Advanced radiation tolerant materials are also being developed to enable the desired extreme fuel burnup levels. Specifically, coatings are being developed to minimize FCCI, and research is underway to fabricate large heats of radiation tolerant oxide dispersion steels with homogeneous oxide

  18. Nanomagnetism of Core-Shell Magnetic Nanoparticles and Application in Spent Nuclear Fuel Separation

    NASA Astrophysics Data System (ADS)

    Tarsem Singh, Maninder Kaur

    This dissertation presents the study on novel core-shell magnetic nanoparticles (NPs) with unique magnetic properties. Understanding the fundamental physics of antiferromagnetic - ferromagnetic interactions is essential to apply in different applications. Chromium (Cr) doped and undoped core-shell iron/iron-oxide NPs have been synthesized using cluster deposition system and studied with respect to their nanostructures, morphologies, sizes, chemical composition and magnetic properties. The room-temperature magnetic properties of Fe based NPs shows the strong dependence of intra/inter-particle interaction on NP size. The Cr-doped Fe NP shows the origin of sigma-FeCr phase at very low Cr concentration (2 at.%) unlike others reported at high Cr content and interaction reversal from dipolar to exchange interaction. A theoretical model of watermelon is constructed based on the experimental results and core-shell NP system in order to explain the physics of exchange interaction in Cr-doped Fe particles. The magnetic nanoparticle---chelator separation nanotechnology is investigated for spent nuclear fuel recycling and is reported 97% and 80% of extraction for Am(III) and Pu(IV) actinides respectively. If the long-term heat generating actinides such as Am(III) can be efficiently removed from the used fuel raffinates, the volume of material that can be placed in a given amount of repository space can be significantly increased. As it is a simple, versatile, compact, and cost efficient process that minimizes secondary waste and improves storage performance.

  19. Core loading pattern optimization of thorium fueled heavy water breeder reactor using genetic algorithm

    SciTech Connect

    Soewono, C. N.; Takaki, N.

    2012-07-01

    In this work genetic algorithm was proposed to solve fuel loading pattern optimization problem in thorium fueled heavy water reactor. The objective function of optimization was to maximize the conversion ratio and minimize power peaking factor. Those objectives were simultaneously optimized using non-dominated Pareto-based population ranking optimal method. Members of non-dominated population were assigned selection probabilities based on their rankings in a manner similar to Baker's single criterion ranking selection procedure. A selected non-dominated member was bred through simple mutation or one-point crossover process to produce a new member. The genetic algorithm program was developed in FORTRAN 90 while neutronic calculation and analysis was done by COREBN code, a module of core burn-up calculation for SRAC. (authors)

  20. Lattice-strain control of the activity in dealloyed core-shell fuel cell catalysts.

    PubMed

    Strasser, Peter; Koh, Shirlaine; Anniyev, Toyli; Greeley, Jeff; More, Karren; Yu, Chengfei; Liu, Zengcai; Kaya, Sarp; Nordlund, Dennis; Ogasawara, Hirohito; Toney, Michael F; Nilsson, Anders

    2010-06-01

    Electrocatalysis will play a key role in future energy conversion and storage technologies, such as water electrolysers, fuel cells and metal-air batteries. Molecular interactions between chemical reactants and the catalytic surface control the activity and efficiency, and hence need to be optimized; however, generalized experimental strategies to do so are scarce. Here we show how lattice strain can be used experimentally to tune the catalytic activity of dealloyed bimetallic nanoparticles for the oxygen-reduction reaction, a key barrier to the application of fuel cells and metal-air batteries. We demonstrate the core-shell structure of the catalyst and clarify the mechanistic origin of its activity. The platinum-rich shell exhibits compressive strain, which results in a shift of the electronic band structure of platinum and weakening chemisorption of oxygenated species. We combine synthesis, measurements and an understanding of strain from theory to generate a reactivity-strain relationship that provides guidelines for tuning electrocatalytic activity.

  1. Scoping studies of the alternative options for defueling, packaging, shipping, and disposing of the TMI-2 spent fuel core

    SciTech Connect

    Anderson, Robert T.

    1980-09-01

    A portion of this fuel will be shipped to nuclear facilities to perform detailed physical examinations. Removal of this fuel from the TMI-2 core is also a significant step in the eventual cleanup of this facility. The report presents a scoping study of the technical operations required for defueling and canning. The TMI fuel when canned could be stored in the spent fuel storage pool. After a period of on-site storage, it is expected that the bulk of the fuel will be shipped off-site for either storage or reprocessing. Evaluation is made of the technical, economic, and institutional factors associated with alternate approaches to disposition of this fuel. Recommendations are presented concerning future generic development tasks needed for the defueling, packaging, on-site shipping of this fuel.

  2. A classification scheme for LWR fuel assemblies

    SciTech Connect

    Moore, R.S.; Williamson, D.A.; Notz, K.J.

    1988-11-01

    With over 100 light water nuclear reactors operating nationwide, representing designs by four primary vendors, and with reload fuel manufactured by these vendors and additional suppliers, a wide variety of fuel assembly types are in existence. At Oak Ridge National Laboratory, both the Systems Integration Program and the Characteristics Data Base project required a classification scheme for these fuels. This scheme can be applied to other areas and is expected to be of value to many Office of Civilian Radioactive Waste Management programs. To develop the classification scheme, extensive information on the fuel assemblies that have been and are being manufactured by the various nuclear fuel vendors was compiled, reviewed, and evaluated. It was determined that it is possible to characterize assemblies in a systematic manner, using a combination of physical factors. A two-stage scheme was developed consisting of 79 assembly types, which are grouped into 22 assembly classes. The assembly classes are determined by the general design of the reactor cores in which the assemblies are, or were, used. The general BWR and PWR classes are divided differently but both are based on reactor core configuration. 2 refs., 15 tabs.

  3. Serrated Au/Pd Core/Shell Nanowires with Jagged Edges for Boosting Liquid Fuel Electrooxidation.

    PubMed

    Zhang, Yu-Ling; Shen, Wen-Jin; Kuang, Wen-Tao; Guo, Shaojun; Li, Yong-Jun; Wang, Ze-Hong

    2017-06-09

    Integration of 1D, core/shell, and jagged features into one entity may provide a promising avenue for further enhancing catalyst performance. However, designing such unique nanostructures is extremely challenging. Herein, 1D serrated Au/Pd core/shell nanowires (CSNWs) with jagged edges were produced simply by a one-pot, dual-capping-agent-assisted method involving co-reduction, galvanic replacement, directional coalescence of preformed nanoparticles, and site-selective epitaxial growth of Pd. Au/PdCSNWs, compared with the commercially available Pd/C, exhibited enhanced electrocatalytic performance towards liquid fuel oxidation because of the synergistic effect of the electronic structure and low-coordinated jagged edges. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  4. Design assumptions and bases for small D-T-fueled Sperical Tokamak (ST) fusion core

    SciTech Connect

    Peng, Y.K.M.; Galambos, J.D.; Fogarty, P.J.

    1996-12-31

    Recent progress in defining the assumptions and clarifying the bases for a small D-T-fueled ST fusion core are presented. The paper covers several issues in the physics of ST plasmas, the technology of neutral beam injection, the engineering design configuration, and the center leg material under intense neutron irradiation. This progress was driven by the exciting data from pioneering ST experiments, a heightened interest in proof-of-principle experiments at the MA level in plasma current, and the initiation of the first conceptual design study of the small ST fusion core. The needs recently identified for a restructured fusion energy sciences program have provided a timely impetus for examining the subject of this paper. Our results, though preliminary in nature, strengthen the case for the potential realism and attractiveness of the ST approach.

  5. Heuristic rules embedded genetic algorithm for in-core fuel management optimization

    NASA Astrophysics Data System (ADS)

    Alim, Fatih

    The objective of this study was to develop a unique methodology and a practical tool for designing loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and BP placement in order to achieve maximum cycle length while satisfying the safety constraints. Genetic Algorithms (GA) have been already used to solve this problem for LP optimization for both PWR and Boiling Water Reactor (BWR). The GA, which is a stochastic method works with a group of solutions and uses random variables to make decisions. Based on the theories of evaluation, the GA involves natural selection and reproduction of the individuals in the population for the next generation. The GA works by creating an initial population, evaluating it, and then improving the population by using the evaluation operators. To solve this optimization problem, a LP optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed in the framework of this thesis. This code is applicable for all types of PWR cores having different geometries and structures with an unlimited number of FA types in the inventory. To reach this goal, an innovative GA is developed by modifying the classical representation of the genotype. To obtain the best result in a shorter time, not only the representation is changed but also the algorithm is changed to use in-core fuel management heuristics rules. The improved GA code was tested to demonstrate and verify the advantages of the new enhancements. The developed methodology is explained in this thesis and preliminary results are shown for the VVER-1000 reactor hexagonal geometry core and the TMI-1 PWR. The improved GA code was tested to verify the advantages of new enhancements. The core physics code

  6. NEPHTIS: Core depletion validation relying on 2D transport core calculations with the APOLLO2 code

    SciTech Connect

    Damian, F.; Raepsaet, X.; Groizard, M.; Poinot, C.

    2006-07-01

    The CEA, in collaboration with EDF and AREVA-NP, is developing a core modelling tool called NEPHTIS, for Neutronic Process for HTGR Innovating Systems and dedicated at present day to the prismatic block-type HTGR (High Temperature Gas-Cooled Reactors). Due to the lack of usable HTGR experimental results, the confidence in this neutronic computational tool relies essentially on comparisons to reference or best-estimate calculations. In the present analysis, the Aleppo deterministic transport code has been selected as reference for validating core depletion simulations carried out within NEPHTIS. These reference calculations were performed on fully detailed 2D core configurations using the Method of Characteristics. The latter has been validated versus Monte Carlo method for different static core configurations [1], [2] and [3]. All the presented results come from an annular HTGR core loaded with uranium-based fuel (15% enrichment). During the core depletion validation, reactivity, reaction rates distributions and nuclei concentrations have been compared. In addition, the impact of various physical and geometrical parameters such as the core loading (one-through or batch-wise reloading) and the amount of burnable poison has been investigated during the validation phases. The results confirm that NEPHTIS is able to predict the core reactivity with uncertainties of {+-}350 pcm. At the end of the core irradiation, the U-235 consumption is calculated within {+-} 0, 7 % while the plutonium mass discharged from the core is calculated within {+-}1 %. As far as the core power distributions are concerned, small discrepancies ( and < 2.3 %) can be observed on the fuel block-averaged power distribution in the core. (authors)

  7. Myocardial Reloading after Extracorporeal Membrane Oxygenation Alters Substrate Metabolism While Promoting Protein Synthesis

    SciTech Connect

    Kajimoto, Masaki; Priddy, Colleen M.; Ledee, Dolena; Xu, Chun; Isern, Nancy G.; Olson, Aaron; Des Rosiers, Christine; Portman, Michael A.

    2013-08-19

    Extracorporeal membrane oxygenation (ECMO) unloads the heart providing a bridge to recovery in children after myocardial stunning. Mortality after ECMO remains high.Cardiac substrate and amino acid requirements upon weaning are unknown and may impact recovery. We assessed the hypothesis that ventricular reloading modulates both substrate entry into the citric acid cycle (CAC) and myocardial protein synthesis. Fourteen immature piglets (7.8-15.6 kg) were separated into 2 groups based on ventricular loading status: 8 hour-ECMO (UNLOAD) and post-wean from ECMO (RELOAD). We infused [2-13C]-pyruvate as an oxidative substrate and [13C6]-L-leucine, as a tracer of amino acid oxidation and protein synthesis into the coronary artery. RELOAD showed marked elevations in myocardial oxygen consumption above baseline and UNLOAD. Pyruvate uptake was markedly increased though RELOAD decreased pyruvate contribution to oxidative CAC metabolism.RELOAD also increased absolute concentrations of all CAC intermediates, while maintaining or increasing 13C-molar percent enrichment. RELOAD also significantly increased cardiac fractional protein synthesis rates by >70% over UNLOAD. Conclusions: RELOAD produced high energy metabolic requirement and rebound protein synthesis. Relative pyruvate decarboxylation decreased with RELOAD while promoting anaplerotic pyruvate carboxylation and amino acid incorporation into protein rather than to the CAC for oxidation. These perturbations may serve as therapeutic targets to improve contractile function after ECMO.

  8. Creation of a Full-Core HTR Benchmark with the Fort St. Vrain Initial Core and Assessment of Uncertainties in the FSV Fuel Composition and Geometry

    SciTech Connect

    Martin, William R.; Lee, John C.; baxter, Alan; Wemple, Chuck

    2012-03-31

    Information and measured data from the intial Fort St. Vrain (FSV) high temperature gas reactor core is used to develop a benchmark configuration to validate computational methods for analysis of a full-core, commercial HTR configuration. Large uncertainties in the geometry and composition data for the FSV fuel and core are identified, including: (1) the relative numbers of fuel particles for the four particle types, (2) the distribution of fuel kernel diameters for the four particle types, (3) the Th:U ratio in the initial FSV core, (4) and the buffer thickness for the fissile and fertile particles. Sensitivity studies were performed to assess each of these uncertainties. A number of methods were developed to assist in these studies, including: (1) the automation of MCNP5 input files for FSV using Python scripts, (2) a simple method to verify isotopic loadings in MCNP5 input files, (3) an automated procedure to conduct a coupled MCNP5-RELAP5 analysis for a full-core FSV configuration with thermal-hydraulic feedback, and (4) a methodology for sampling kernel diameters from arbitrary power law and Gaussian PDFs that preserved fuel loading and packing factor constraints. A reference FSV fuel configuration was developed based on having a single diameter kernel for each of the four particle types, preserving known uranium and thorium loadings and packing factor (58%). Three fuel models were developed, based on representing the fuel as a mixture of kernels with two diameters, four diameters, or a continuous range of diameters. The fuel particles were put into a fuel compact using either a lattice-bsed approach or a stochastic packing methodology from RPI, and simulated with MCNP5. The results of the sensitivity studies indicated that the uncertainties in the relative numbers and sizes of fissile and fertile kernels were not important nor were the distributions of kernel diameters within their diameter ranges. The uncertainty in the Th:U ratio in the intial FSV core was

  9. Supported Core@Shell Electrocatalysts for Fuel Cells: Close Encounter with Reality

    PubMed Central

    Hwang, Seung Jun; Yoo, Sung Jong; Shin, Jungho; Cho, Yong-Hun; Jang, Jong Hyun; Cho, Eunae; Sung, Yung-Eun; Nam, Suk Woo; Lim, Tae-Hoon; Lee, Seung-Cheol; Kim, Soo-Kil

    2013-01-01

    Core@shell electrocatalysts for fuel cells have the advantages of a high utilization of Pt and the modification of its electronic structures toward enhancement of the activities. In this study, we suggest both a theoretical background for the design of highly active and stable core@shell/C and a novel facile synthetic strategy for their preparation. Using density functional theory calculations guided by the oxygen adsorption energy and vacancy formation energy, Pd3Cu1@Pt/C was selected as the most suitable candidate for the oxygen reduction reaction in terms of its activity and stability. These predictions were experimentally verified by the surfactant-free synthesis of Pd3Cu1/C cores and the selective Pt shell formation using a Hantzsch ester as a reducing agent. In a similar fashion, Pd@Pd4Ir6/C catalyst was also designed and synthesized for the hydrogen oxidation reaction. The developed catalysts exhibited high activity, high selectivity, and 4,000 h of long-term durability at the single-cell level. PMID:23419683

  10. Modeling gadolinium-bearing fuel in Ringhals PWRs using CASMO/SIMULATE

    SciTech Connect

    Kurcyusz, E. )

    1993-01-01

    Ringhals units 2, 3, and 4 are Westinghouse three-loop, 157-assembly pressurized water reactors (PWRs) operated by Vattenfall. Originally, all three reactors were loaded in an out-in scheme using reload fuel without burnable poisons. In recent cycles, gadolinium-bearing fuel was introduced to enable a low-leakage loading pattern and minimize fuel cycle costs. This paper focuses on the Fragema 17 x 17 AFA design with peripheral gadolinium rods loaded in units 3 and 4. The Ringhals units are modeled using the Studsvik core management system, consisting of the CASMO-3 transport theory lattice physics code,and the SIMULATE-3 advanced nodal reactor analysis code. The results of the studies verifying the accuracy of CASMO-3/SIMULATE-3 on the assemblies with peripheral gadolinium rods are presented in this paper. The verification was carried out against CASMO-3 color-set calculations and measured reactor data.

  11. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    SciTech Connect

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  12. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    SciTech Connect

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  13. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  14. ALARA Analysis for Shippingport Pressurized Water Reactor Core 2 Fuel Storage in the Canister Storage Building (CSB)

    SciTech Connect

    LEWIS, M.E.

    2000-04-06

    The addition of Shippingport Pressurized Water Reactor (PWR) Core 2 Blanket Fuel Assembly storage in the Canister Storage Building (CSB) will increase the total cumulative CSB personnel exposure from receipt and handling activities. The loaded Shippingport Spent Fuel Canisters (SSFCs) used for the Shippingport fuel have a higher external dose rate. Assuming an MCO handling rate of 170 per year (K East and K West concurrent operation), 24-hr CSB operation, and nominal SSFC loading, all work crew personnel will have a cumulative annual exposure of less than the 1,000 mrem limit.

  15. Effect of buoyancy on fuel containment in an open-cycle gas-core nuclear rocket engine.

    NASA Technical Reports Server (NTRS)

    Putre, H. A.

    1971-01-01

    Analysis aimed at determining the scaling laws for the buoyancy effect on fuel containment in an open-cycle gas-core nuclear rocket engine, so conducted that experimental conditions can be related to engine conditions. The fuel volume fraction in a short coaxial flow cavity is calculated with a programmed numerical solution of the steady Navier-Stokes equations for isothermal, variable density fluid mixing. A dimensionless parameter B, called the Buoyancy number, was found to correlate the fuel volume fraction for large accelerations and various density ratios. This parameter has the value B = 0 for zero acceleration, and B = 350 for typical engine conditions.

  16. Contribution of neutron-capture reactions in energy release in the fuel core of BN-600

    NASA Astrophysics Data System (ADS)

    Bahdanovich, R. B.; Romanenko, V. I.; Tikhomirov, G. V.

    2017-01-01

    The use of modern computing powers and calculation methods allows to get closer to reality results of modelling, as well as to explore areas inaccessible to the experiment. Until now, the calculation of the energy released from the capture of neutrons in the reactor core has been given little attention. The method for calculation of the effective energy release components in a nuclear reactor allows to specify the values used by engineering programs for capture energy release in fast reactors. The paper presents improved method and the results of calculation of three models of the reactor BN-600. It is shown that the contribution of capture energy release in effective energy release for fresh fuel is equal to 4%, which is more than for VVER reactors. During the calculation we created a simple calculation model of the fast reactor, considering its features.

  17. Fuel/propellant mixing in an open-cycle gas core nuclear rocket engine

    NASA Astrophysics Data System (ADS)

    Guo, Xu; Wehrmeyer, Joseph A.

    1997-01-01

    A numerical investigation of the mixing of gaseous uranium and hydrogen inside an open-cycle gas core nuclear rocket engine (spherical geometry) is presented. The gaseous uranium fuel is injected near the centerline of the spherical engine cavity at a constant mass flow rate, and the hydrogen propellant is injected around the periphery of the engine at a five degree angle to the wall, at a constant mass flow rate. The main objective is to seek ways to minimize the mixing of uranium and hydrogen by choosing a suitable injector geometry for the mixing of light and heavy gas streams. Three different uranium inlet areas are presented, and also three different turbulent models (k-ɛ model, RNG k-V model, and RSM model) are investigated. The commercial CFD code, FLUENT, is used to model the flow field. Uranium mole fraction, axial mass flux, and radial mass flux contours are obtained.

  18. Measurement of gamma field parameters in core with LEU fuel IRT-4M using TL detectors

    SciTech Connect

    Bily, T.

    2008-07-15

    Thermoluminescent dosimeters represent very useful tool for gamma fields parameters measurements at nuclear research reactors, especially at zero power ones. {sup 7}LiF:Mg,Ti and {sup 7}LiF:Mg,Cu,P type TL dosimeters enable determination of only gamma component in mixed neutron - gamma field. At VR-1 reactor operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague the integral characteristics of gamma rays field were investigated, especially its spatial distribution and time behaviour, i.e. the non-saturated delayed gamma ray emission influence. Measured spatial distributions were compared with monte carlo code MCNP5 calculations. Although MCNP cannot generate delayed gamma rays from fission, the relative gamma dose rate distribution is within {+-} 15% with measured values. The experiments were carried out with core configuration C1 consisting of LEU fuel IRT-4M (19.7 %). (author)

  19. The effect of fuel thermal conductivity on the behavior of LWR cores during loss-of-coolant accidents

    SciTech Connect

    Terrani, Kurt A.; Wang, Dean; Ott, Larry J.; Montgomery, Robert O.

    2014-05-01

    The effect of variation in thermal conductivity of light water reactor fuel elements on core response during loss-of-coolant accident scenarios is examined. Initially, a simplified numerical analysis is utilized to determine the time scales associated with dissipation of stored energy from the fuel into the coolant once the fission reaction is stopped. The analysis is then followed by full reactor system thermal-hydraulics analysis of a typical boiling and pressurized water reactor subjected to a large break loss-of-coolant accident scenario using the TRACE code. Accordingly, sensitivity analyses to examine the effect of an increase in fuel thermal conductivity, up to 500%, on fuel temperature evolution during these transients are performed. Given the major differences in thermal-hydraulics design aspects of boiling and pressurized water reactors, different fuel and temperature responses during the simulated loss-of-coolant transients are observed.

  20. On the core deuterium-tritium fuel ratio and temperature measurements in DEMO

    NASA Astrophysics Data System (ADS)

    Kiptily, V. G.

    2015-02-01

    Comparing with ITER, the experimental fusion machine under construction, the next-step test fusion power plant, DEMO will be characterized by a very long pulse/steady-state operation and much higher plasma volume and fusion power. The substantially increased level of neutron and gamma fluxes will require reducing the physical access to the plant. It means some conventional diagnostics for the fusion plasma control will not be suitable in DEMO. Development of diagnostics along with the machine design is a primary task for the test plant. The deuterium-tritium fuel ratio and temperature are among important parameters, which should be under control. In this paper, a novel technique for the core fuel ratio and temperature diagnostics is proposed. It is based on measurements and comparison of the rates T(p, γ)4He and D(T, γ)5He nuclear reactions that take place in the hot deuterium-tritium plasma. Based on detection of high-energy gamma-rays, this diagnostic is robust, efficient and does not require direct access to the plasma. It could be included in the loop of the burning plasma control system. A feasibility of the diagnostic in experiments on JET and ITER is also discussed.

  1. EVALUATION OF CORE PHYSICS ANALYSIS METHODS FOR CONVERSION OF THE INL ADVANCED TEST REACTOR TO LOW-ENRICHMENT FUEL

    SciTech Connect

    Mark DeHart; Gray S. Chang

    2012-04-01

    Computational neutronics studies to support the possible conversion of the ATR to LEU are underway. Simultaneously, INL is engaged in a physics methods upgrade project to put into place modern computational neutronics tools for future support of ATR fuel cycle and experiment analysis. A number of experimental measurements have been performed in the ATRC in support of the methods upgrade project, and are being used to validate the new core physics methods. The current computational neutronics work is focused on performance of scoping calculations for the ATR core loaded with a candidate LEU fuel design. This will serve as independent confirmation of analyses that have been performed previously, and will evaluate some of the new computational methods for analysis of a candidate LEU fuel for ATR.

  2. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    SciTech Connect

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations.

  3. Fuel cell performance of palladium-platinum core-shell electrocatalysts synthesized in gram-scale batches

    SciTech Connect

    Khateeb, Siddique; Su, Dong; Guerreo, Sandra; Darling, Robert M.; Protsailo, Lesia V.; Shao, Minhua

    2016-05-03

    This article presents the performance of palladium-platinum core-shell catalysts (Pt/Pd/C) for oxygen reduction synthesized in gram-scale batches in both liquid cells and polymer-electrolyte membrane fuel cells. Core-shell catalyst synthesis and characterization, ink fabrication, and cell assembly details are discussed. The Pt mass activity of the Pt/Pd core-shell catalyst was 0.95 A mg–1 at 0.9 V measured in liquid cells (0.1 M HClO4), which was 4.8 times higher than a commercial Pt/C catalyst. The performances of Pt/Pd/C and Pt/C in large single cells (315 cm2) were assessed under various operating conditions. The core-shell catalyst showed consistently higher performance than commercial Pt/C in fuel cell testing. A 20–60 mV improvement across the whole current density range was observed on air. Sensitivities to temperature, humidity, and gas composition were also investigated and the core-shell catalyst showed a consistent benefit over Pt under all conditions. However, the 4.8 times activity enhancement predicated by liquid cell measurements was not fully realized in fuel cells.

  4. Fuel cell performance of palladium-platinum core-shell electrocatalysts synthesized in gram-scale batches

    DOE PAGES

    Khateeb, Siddique; Su, Dong; Guerreo, Sandra; ...

    2016-05-03

    This article presents the performance of palladium-platinum core-shell catalysts (Pt/Pd/C) for oxygen reduction synthesized in gram-scale batches in both liquid cells and polymer-electrolyte membrane fuel cells. Core-shell catalyst synthesis and characterization, ink fabrication, and cell assembly details are discussed. The Pt mass activity of the Pt/Pd core-shell catalyst was 0.95 A mg–1 at 0.9 V measured in liquid cells (0.1 M HClO4), which was 4.8 times higher than a commercial Pt/C catalyst. The performances of Pt/Pd/C and Pt/C in large single cells (315 cm2) were assessed under various operating conditions. The core-shell catalyst showed consistently higher performance than commercial Pt/Cmore » in fuel cell testing. A 20–60 mV improvement across the whole current density range was observed on air. Sensitivities to temperature, humidity, and gas composition were also investigated and the core-shell catalyst showed a consistent benefit over Pt under all conditions. However, the 4.8 times activity enhancement predicated by liquid cell measurements was not fully realized in fuel cells.« less

  5. Transition Core Properties during Conversion of the NBSR from HEU to LEU Fuel

    SciTech Connect

    Hanson A. L.; Diamond D.

    2013-10-31

    The transition of the NBSR from HEU to LEU fuel is challenging due to reactivity constraints and the need to maintain an uninterrupted science program, the mission of the NBSR. The transition cannot occur with a full change of HEU to LEU fuel elements since the excess reactivity would be large enough that the NBSR would violate the technical specification for shutdown margin. Manufacturing LEU fuel elements to represent irradiated fuel elements would be cost prohibitive since 26 one-of-a-kind fuel elements would need to be manufactured. For this report a gradual transition from the present HEU fuel to the proposed LEU fuel was studied. The gradual change approach would follow the present fuel management scheme and replace four HEU fuel elements with four LEU fuel elements each cycle. This manuscript reports the results of a series of calculations to predict the neutronic characteristics and how the neutronics will change during the transition from HEU to LEU in the NBSR.

  6. An In-Core Power Deposition and Fuel Thermal Environmental Monitor for Long-Lived Reactor Cores

    SciTech Connect

    Don W. Miller

    2004-09-28

    The primary objective of this program is to develop the Constant Temperature Power Sensor (CTPS) as in-core instrumentation that will provide a detailed map of local nuclear power deposition and coolant thermal-hydraulic conditions during the entire life of the core.

  7. Impact of Americium-241 (n,γ) Branching Ratio on SFR Core Reactivity and Spent Fuel Characteristics

    SciTech Connect

    Hiruta, Hikaru; Youinou, Gilles J.; Dixon, Brent W.

    2016-06-01

    An accurate prediction of core physics and fuel cycle parameters largely depends on the order of details and accuracy in nuclear data taken into account for actual calculations. 241Am is a major gateway nuclide for most of minor actinides and thus important nuclide for core physics and fuel-cycle calculations. The 241Am(n,?) branching ratio (BR) is in fact the energy dependent (see Fig. 1), therefore, it is necessary to taken into account the spectrum effect on the calculation of the average BR for the full-core depletion calculations. Moreover, the accuracy of the BR used in the depletion calculations could significantly influence the core physics performance and post irradiated fuel compositions. The BR of 241Am(n,?) in ENDF/B-VII.0 library is relatively small and flat in thermal energy range, gradually increases within the intermediate energy range, and even becomes larger at the fast energy range. This indicates that the properly collapsed BR for fast reactors could be significantly different from that of thermal reactors. The evaluated BRs are also differ from one evaluation to another. As seen in Table I, average BRs for several evaluated libraries calculated by means of a fast spectrum are similar but have some differences. Most of currently available depletion codes use a pre-determined single value BR for each library. However, ideally it should be determined on-the-fly basis like that of one-group cross sections. These issues provide a strong incentive to investigate the effect of different 241Am(n,?) BRs on core and spent fuel parameters. This paper investigates the impact of the 241Am(n,?) BR on the results of SFR full-core based fuel-cycle calculations. The analysis is performed by gradually increasing the value of BR from 0.15 to 0.25 and studying its impact on the core reactivity and characteristics of SFR spent fuels over extended storage times (~10,000 years).

  8. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    SciTech Connect

    Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; Marshall, Margaret A.

    2014-11-04

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  9. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    DOE PAGES

    Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; ...

    2014-11-04

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding asmore » well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

  10. High performance of SDC and GDC core shell type composite electrolytes using methane as a fuel for low temperature SOFC

    NASA Astrophysics Data System (ADS)

    Irshad, Muneeb; Siraj, Khurram; Raza, Rizwan; Javed, Fayyaz; Ahsan, Muhammad; Shakir, Imran; Rafique, Muhammad Shahid

    2016-02-01

    Nanocomposites Samarium doped Ceria (SDC), Gadolinium doped Ceria (GDC), core shell SDC amorphous Na2CO3 (SDCC) and GDC amorphous Na2CO3 (GDCC) were synthesized using co-precipitation method and then compared to obtain better solid oxide electrolytes materials for low temperature Solid Oxide Fuel Cell (SOFCs). The comparison is done in terms of structure, crystallanity, thermal stability, conductivity and cell performance. In present work, XRD analysis confirmed proper doping of Sm and Gd in both single phase (SDC, GDC) and dual phase core shell (SDCC, GDCC) electrolyte materials. EDX analysis validated the presence of Sm and Gd in both single and dual phase electrolyte materials; also confirming the presence of amorphous Na2CO3 in SDCC and GDCC. From TGA analysis a steep weight loss is observed in case of SDCC and GDCC when temperature rises above 725 °C while SDC and GDC do not show any loss. The ionic conductivity and cell performance of single phase SDC and GDC nanocomposite were compared with core shell GDC/amorphous Na2CO3 and SDC/ amorphous Na2CO3 nanocomposites using methane fuel. It is observed that dual phase core shell electrolytes materials (SDCC, GDCC) show better performance in low temperature range than their corresponding single phase electrolyte materials (SDC, GDC) with methane fuel.

  11. High performance of SDC and GDC core shell type composite electrolytes using methane as a fuel for low temperature SOFC

    SciTech Connect

    Irshad, Muneeb; Siraj, Khurram E-mail: khurram.uet@gmail.com; Javed, Fayyaz; Ahsan, Muhammad; Rafique, Muhammad Shahid; Raza, Rizwan E-mail: khurram.uet@gmail.com; Shakir, Imran

    2016-02-15

    Nanocomposites Samarium doped Ceria (SDC), Gadolinium doped Ceria (GDC), core shell SDC amorphous Na{sub 2}CO{sub 3} (SDCC) and GDC amorphous Na{sub 2}CO{sub 3} (GDCC) were synthesized using co-precipitation method and then compared to obtain better solid oxide electrolytes materials for low temperature Solid Oxide Fuel Cell (SOFCs). The comparison is done in terms of structure, crystallanity, thermal stability, conductivity and cell performance. In present work, XRD analysis confirmed proper doping of Sm and Gd in both single phase (SDC, GDC) and dual phase core shell (SDCC, GDCC) electrolyte materials. EDX analysis validated the presence of Sm and Gd in both single and dual phase electrolyte materials; also confirming the presence of amorphous Na{sub 2}CO{sub 3} in SDCC and GDCC. From TGA analysis a steep weight loss is observed in case of SDCC and GDCC when temperature rises above 725 °C while SDC and GDC do not show any loss. The ionic conductivity and cell performance of single phase SDC and GDC nanocomposite were compared with core shell GDC/amorphous Na{sub 2}CO{sub 3} and SDC/ amorphous Na{sub 2}CO{sub 3} nanocomposites using methane fuel. It is observed that dual phase core shell electrolytes materials (SDCC, GDCC) show better performance in low temperature range than their corresponding single phase electrolyte materials (SDC, GDC) with methane fuel.

  12. Lattice-Strain Control of Exceptional Activity in Dealloyed Core-Shell Fuel Cell Catalysts

    SciTech Connect

    Strasser, Peter

    2011-08-19

    We present a combined experimental and theoretical approach to demonstrate how lattice strain can be used to continuously tune the catalytic activity of the oxygen reduction reaction (ORR) on bimetallic nanoparticles that have been dealloyed. The sluggish kinetics of the ORR is a key barrier to the adaptation of fuel cells and currently limits their widespread use. Dealloyed Pt-Cu bimetallic nanoparticles, however, have been shown to exhibit uniquely high reactivity for this reaction. We first present evidence for the formation of a core-shell structure during dealloying, which involves removal of Cu from the surface and subsurface of the precursor nanoparticles. We then show that the resulting Pt-rich surface shell exhibits compressive strain that depends on the composition of the precursor alloy. We next demonstrate the existence of a downward shift of the Pt d-band, resulting in weakening of the bond strength of intermediate oxygenated species due to strain. Finally, we combine synthesis, strain, and catalytic reactivity in an experimental/theoretical reactivity-strain relationship which provides guidelines for the rational design of strained oxygen reduction electrocatalysts. The stoichiometry of the precursor, together with the dealloying conditions, provides experimental control over the resulting surface strain and thereby allows continuous tuning of the surface electrocatalytic reactivity - a concept that can be generalized to other catalytic reactions.

  13. Detailed OEDGE Modeling of Core-Pedestal Fueling in DIII-D

    NASA Astrophysics Data System (ADS)

    Elder, J. D.; Stangeby, P. C.; Leonard, A. W.; Fenstermacher, M. E.; Boedo, J. A.; Rudakov, D. L.; Bray, B. D.; Brooks, N. H.; Watkins, J. G.; Unterberg, E. A.

    2011-10-01

    The OEDGE code is used to model the deuterium neutral density and ionization distribution inside the separatrix for an attached L-mode SAPP discharge and an attached ELMy H-mode discharge. The background plasma solution is determined by empirical plasma reconstruction matching as many diagnostic measurements as possible. Recycling fluxes are obtained from measurements by Langmuir probes and spectroscopic measurements of Dα. The relative importance of wall, divertor and recombination sources to core and pedestal fueling are assessed. In addition, the sensitivity of the ionization source location to the details of the plasma solution in the divertor is examined. Several models for plasma-wall contact are used to estimate the strength of the wall recycling source. In the L-mode case, ionization profiles peak at the flux surface ~1.3 cm inboard of the separatrix (mapped to the outer midplane). Work supported by US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-FG02-07ER54917, DE-AC04-94AL85000, and DE-AC05-06OR21300.

  14. Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

    NASA Astrophysics Data System (ADS)

    Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

    2017-01-01

    The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

  15. Iridium-decorated palladium-platinum core-shell catalysts for oxygen reduction reaction in proton exchange membrane fuel cell.

    PubMed

    Wang, Chen-Hao; Hsu, Hsin-Cheng; Wang, Kai-Ching

    2014-08-01

    Carbon-supported Pt, Pd, Pd-Pt core-shell (Pt(shell)-Pd(core)/C) and Ir-decorated Pd-Pt core-shell (Ir-decorated Pt(shell)-Pd(core)/C) catalysts were synthesized, and their physical properties, electrochemical behaviors, oxygen reduction reaction (ORR) characteristics and proton exchange membrane fuel cell (PEMFC) performances were investigated herein. From the XRD patterns and TEM images, Ir-decorated Pt(shell)-Pd(core)/C has been confirmed that Pt was deposited on the Pd nanoparticle which had the core-shell structure. Ir-decorated Pt(shell)-Pd(core)/C has more positive OH reduction peak than Pt/C, which is beneficial to weaken the binding energy of Pt-OH during the ORR. Thus, Ir-decorated Pt(shell)-Pd(core)/C has higher ORR activity than Pt/C. The maximum power density of H2-O2 PEMFC using Ir-decorated Pt(shell)-Pd(core)/C is 792.2 mW cm(-2) at 70°C, which is 24% higher than that using Pt/C. The single-cell accelerated degradation test of PEMFC using Ir-decorated Pt(shell)-Pd(core)/C shows good durability by the potential cycling of 40,000 cycles. This study concludes that Ir-decorated Pt(shell)-Pd(core)/C has the low Pt content, but it can facilitate the low-cost and high-efficient PEMFC. Copyright © 2013 Elsevier Inc. All rights reserved.

  16. Ordered bilayer ruthenium-platinum core-shell nanoparticles as carbon monoxide-tolerant fuel cell catalysts.

    PubMed

    Hsieh, Yu-Chi; Zhang, Yu; Su, Dong; Volkov, Vyacheslav; Si, Rui; Wu, Lijun; Zhu, Yimei; An, Wei; Liu, Ping; He, Ping; Ye, Siyu; Adzic, Radoslav R; Wang, Jia X

    2013-01-01

    Fabricating subnanometre-thick core-shell nanocatalysts is effective for obtaining high surface area of an active metal with tunable properties. The key to fully realize the potential of this approach is a reliable synthesis method to produce atomically ordered core-shell nanoparticles. Here we report new insights on eliminating lattice defects in core-shell syntheses and opportunities opened for achieving superior catalytic performance. Ordered structural transition from ruthenium hcp to platinum fcc stacking sequence at the core-shell interface is achieved via a green synthesis method, and is verified by X-ray diffraction and electron microscopic techniques coupled with density functional theory calculations. The single crystalline Ru cores with well-defined Pt bilayer shells resolve the dilemma in using a dissolution-prone metal, such as ruthenium, for alleviating the deactivating effect of carbon monoxide, opening the door for commercialization of low-temperature fuel cells that can use inexpensive reformates (H2 with CO impurity) as the fuel.

  17. Muscle regeneration during hindlimb unloading results in a reduction in muscle size after reloading

    NASA Technical Reports Server (NTRS)

    Mozdziak, P. E.; Pulvermacher, P. M.; Schultz, E.

    2001-01-01

    The hindlimb-unloading model was used to study the ability of muscle injured in a weightless environment to recover after reloading. Satellite cell mitotic activity and DNA unit size were determined in injured and intact soleus muscles from hindlimb-unloaded and age-matched weight-bearing rats at the conclusion of 28 days of hindlimb unloading, 2 wk after reloading, and 9 wk after reloading. The body weights of hindlimb-unloaded rats were significantly (P < 0.05) less than those of weight-bearing rats at the conclusion of hindlimb unloading, but they were the same (P > 0.05) as those of weight-bearing rats 2 and 9 wk after reloading. The soleus muscle weight, soleus muscle weight-to-body weight ratio, myofiber diameter, number of nuclei per millimeter, and DNA unit size were significantly (P < 0.05) smaller for the injured soleus muscles from hindlimb-unloaded rats than for the soleus muscles from weight-bearing rats at each recovery time. Satellite cell mitotic activity was significantly (P < 0.05) higher in the injured soleus muscles from hindlimb-unloaded rats than from weight-bearing rats 2 wk after reloading, but it was the same (P > 0.05) as in the injured soleus muscles from weight-bearing rats 9 wk after reloading. The injured soleus muscles from hindlimb-unloaded rats failed to achieve weight-bearing muscle size 9 wk after reloading, because incomplete compensation for the decrease in myonuclear accretion and DNA unit size expansion occurred during the unloading period.

  18. Muscle regeneration during hindlimb unloading results in a reduction in muscle size after reloading

    NASA Technical Reports Server (NTRS)

    Mozdziak, P. E.; Pulvermacher, P. M.; Schultz, E.

    2001-01-01

    The hindlimb-unloading model was used to study the ability of muscle injured in a weightless environment to recover after reloading. Satellite cell mitotic activity and DNA unit size were determined in injured and intact soleus muscles from hindlimb-unloaded and age-matched weight-bearing rats at the conclusion of 28 days of hindlimb unloading, 2 wk after reloading, and 9 wk after reloading. The body weights of hindlimb-unloaded rats were significantly (P < 0.05) less than those of weight-bearing rats at the conclusion of hindlimb unloading, but they were the same (P > 0.05) as those of weight-bearing rats 2 and 9 wk after reloading. The soleus muscle weight, soleus muscle weight-to-body weight ratio, myofiber diameter, number of nuclei per millimeter, and DNA unit size were significantly (P < 0.05) smaller for the injured soleus muscles from hindlimb-unloaded rats than for the soleus muscles from weight-bearing rats at each recovery time. Satellite cell mitotic activity was significantly (P < 0.05) higher in the injured soleus muscles from hindlimb-unloaded rats than from weight-bearing rats 2 wk after reloading, but it was the same (P > 0.05) as in the injured soleus muscles from weight-bearing rats 9 wk after reloading. The injured soleus muscles from hindlimb-unloaded rats failed to achieve weight-bearing muscle size 9 wk after reloading, because incomplete compensation for the decrease in myonuclear accretion and DNA unit size expansion occurred during the unloading period.

  19. Reloading partly recovers bone mineral density and mechanical properties in hind limb unloaded rats

    NASA Astrophysics Data System (ADS)

    Zhao, Fan; Li, Dijie; Arfat, Yasir; Chen, Zhihao; Liu, Zonglin; Lin, Yu; Ding, Chong; Sun, Yulong; Hu, Lifang; Shang, Peng; Qian, Airong

    2014-12-01

    Skeletal unloading results in decreased bone formation and bone mass. During long-term space flight, the decreased bone mass is impossible to fully recover. Therefore, it is necessary to develop the effective countermeasures to prevent spaceflight-induced bone loss. Hindlimb Unloading (HLU) simulates effects of weightlessness and is utilized extensively to examine the response of musculoskeletal systems to certain aspects of space flight. The purpose of this study is to investigate the effects of a 4-week HLU in rats and subsequent reloading on the bone mineral density (BMD) and mechanical properties of load-bearing bones. After HLU for 4 weeks, the rats were then subjected to reloading for 1 week, 2 weeks and 3 weeks, and then the BMD of the femur, tibia and lumbar spine in rats were assessed by dual energy X-ray absorptiometry (DXA) every week. The mechanical properties of the femur were determined by three-point bending test. Dry bone and bone ash of femur were obtained through Oven-Drying method and were weighed respectively. Serum alkaline phosphatase (ALP) and serum calcium were examined through ELISA and Atomic Absorption Spectrometry. The results showed that 4 weeks of HLU significantly decreased body weight of rats and reloading for 1 week, 2 weeks or 3 weeks did not recover the weight loss induced by HLU. However, after 2 weeks of reloading, BMD of femur and tibia of HLU rats partly recovered (+10.4%, +2.3%). After 3 weeks of reloading, the reduction of BMD, energy absorption, bone mass and mechanical properties of bone induced by HLU recovered to some extent. The changes in serum ALP and serum calcium induced by HLU were also recovered after reloading. Our results indicate that a short period of reloading could not completely recover bone after a period of unloading, thus some interventions such as mechanical vibration or pharmaceuticals are necessary to help bone recovery.

  20. Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

    SciTech Connect

    Katoh, Yutai; Terrani, Kurt A.

    2015-08-01

    Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

  1. Generalized Thermohydraulics Module GENFLO for Combining With the PWR Core Melting Model, BWR Recriticality Neutronics Model and Fuel Performance Model

    SciTech Connect

    Miettinen, Jaakko; Hamalainen, Anitta; Pekkarinen, Esko

    2002-07-01

    Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined codes. The basic field equations in GENFLO are for the phase mass, the mixture momentum and phase energy conservation equations. The phase separation is solved with the drift flux model. The basic variables to be solved are the pressure, void fraction, mixture velocity, gas enthalpy, liquid enthalpy, and concentration of non-condensable gas fractions. The validation of the thermohydraulic solution alone includes large break LOCA reflooding experiments and in specific for the severe accident conditions QUENCH tests. In the recriticality analysis the core neutronics is simulated with a two-dimensional transient neutronics code TWODIM. The recriticality with one rapid prompt peak is expected during a severe accident scenario, where the control rods have been melted and ECCS reflooding is started after the depressurization. The GENFLO module simulates the BWR thermohydraulics in this application. The core melting module has been developed for the real time operator training by using the APROS engineering simulators. The core heatup, oxidation, metal and fuel pellet relocation and corium pool formation into the lower plenum are calculated. In this application the GENFLO model simulates the PWR vessel thermohydraulics. In the fuel performance analysis the fuel rod transient behavior is simulated with the FRAPTRAN code. GENFLO simulates the subchannel around a single fuel rod and delivers the heat transfer on the cladding surface for the FRAPTRAN. The transient boundary conditions for the subchannel are transmitted from the system code for operational transient, loss of coolant accidents and

  2. Aluminum/uranium fuel foaming/recriticality considerations for production reactor core-melt accidents

    SciTech Connect

    Hyder, M.L.; Ellison, P.G. ); Cronenberg, A.W. )

    1990-01-01

    Severe accident studies for the Savannah River production reactors indicate that if coherent fuel melting and relocation occur in the absence of target melting, in-vessel recriticality may be achieved. In this paper, fuel-melt/target interaction potential is assessed where fission gas-induced fuel foaming and melt attack on target material are evaluated and compared with available data. Models are developed to characterize foams for irradiated aluminum-based fuel. Predictions indicate transient foaming, the extent of which is governed by fission gas inventory, heating transient conditions, and bubble coalescence behavior. The model also indicates that metallic foams are basically unstable and will collapse, which largely depends on film tenacity and melt viscosity considerations. For high-burnup fuel, extensive foaming lasting tens of seconds is predicted, allowing molten fuel to contact and cause melt ablation of concentric targets. For low-burnup fuel, contact can not be assured. 9 refs., 4 figs., 4 tabs.

  3. GPU Based General-Purpose Parallel computing to Solve Nuclear Reactor In-Core fuel Management Design and Operation Problem

    NASA Astrophysics Data System (ADS)

    Prayudhatama, D.; Waris, A.; Kurniasih, N.; Kurniadi, R.

    2010-06-01

    In-core fuel management study is a crucial activity in nuclear power plant design and operation. Its common problem is to find an optimum arrangement of fuel assemblies inside the reactor core. Main objective for this activity is to reduce the cost of generating electricity, which can be done by altering several physical properties of the nuclear reactor without violating any of the constraints imposed by operational and safety considerations. This research try to address the problem of nuclear fuel arrangement problem, which is, leads to the multi-objective optimization problem. However, the calculation of the reactor core physical properties itself is a heavy computation, which became obstacle in solving the optimization problem by using genetic algorithm optimization. This research tends to address that problem by using the emerging General Purpose Computation on Graphics Processing Units (GPGPU) techniques implemented by C language for CUDA (Compute Unified Device Architecture) parallel programming. By using this parallel programming technique, we develop parallelized nuclear reactor fitness calculation, which is involving numerical finite difference computation. This paper describes current prototype of the parallel algorithm code we have developed on CUDA, that performs one hundreds finite difference calculation for nuclear reactor fitness evaluation in parallel by using GPU G9 Hardware Series developed by NVIDIA.

  4. GPU Based General-Purpose Parallel computing to Solve Nuclear Reactor In-Core fuel Management Design and Operation Problem

    SciTech Connect

    Prayudhatama, D.; Waris, A.; Kurniasih, N.; Kurniadi, R.

    2010-06-22

    In-core fuel management study is a crucial activity in nuclear power plant design and operation. Its common problem is to find an optimum arrangement of fuel assemblies inside the reactor core. Main objective for this activity is to reduce the cost of generating electricity, which can be done by altering several physical properties of the nuclear reactor without violating any of the constraints imposed by operational and safety considerations. This research try to address the problem of nuclear fuel arrangement problem, which is, leads to the multi-objective optimization problem. However, the calculation of the reactor core physical properties itself is a heavy computation, which became obstacle in solving the optimization problem by using genetic algorithm optimization.This research tends to address that problem by using the emerging General Purpose Computation on Graphics Processing Units (GPGPU) techniques implemented by C language for CUDA (Compute Unified Device Architecture) parallel programming. By using this parallel programming technique, we develop parallelized nuclear reactor fitness calculation, which is involving numerical finite difference computation. This paper describes current prototype of the parallel algorithm code we have developed on CUDA, that performs one hundreds finite difference calculation for nuclear reactor fitness evaluation in parallel by using GPU G9 Hardware Series developed by NVIDIA.

  5. Validation of the REBUS-3/RCT methodologies for EBR-II core-follow analysis

    SciTech Connect

    McKnight, R.D.

    1992-01-01

    One of the many tasks to be completed at EBR-2/FCF (Fuel Cycle Facility) regarding fuel cycle closure for the Integral Fast Reactor (IFR) is to develop and install the systems to be used for fissile material accountancy and control. The IFR fuel cycle and pyrometallurgical process scheme determine the degree of actinide of actinide buildup in the reload fuel assemblies. Inventories of curium, americium and neptunium in the fuel will affect the radiation and thermal environmental conditions at the fuel fabrication stations, the chemistry of reprocessing, and the neutronic performance of the core. Thus, it is important that validated calculational tools be put in place for accurately determining isotopic mass and neutronic inputs to FCF for both operational and material control and accountancy purposes. The primary goal of this work is to validate the REBUS-2/RCT codes as tools which can adequately compute the burnup and isotopic distribution in binary- and ternary-fueled Mark-3, Mark-4, and Mark-5 subassemblies. 6 refs.

  6. A plutonium-fueled high-moderated pressurized water reactor for the next century

    SciTech Connect

    Barbrault, P.

    1996-02-01

    Within the framework of French reprocessing policy, for several years, Electricite de France has been studying a high-moderating-ratio (HMR) pressurized water reactor that could accept 100% mixed-oxide (MOX) reloads. Total plutonium content is 9% to ensure a discharge burnup of 60,000 MWd/tonne. A high-moderating ratio (2.5 instead of 2.0) is obtained by replacing 36 fuel rods by water holes. This solution combines the advantages of high moderation (better efficiency of soluble boron, control rods, etc.) and technological continuity. The core should contain 241 fuel assemblies for a total thermal output of 4,250 MW(thermal). The fuel management is easy, but core control requires the use of {sup 10}B-enriched boron carbide for the control rods and {sup 10}B-enriched soluble boric acid for the primary system, thereby ensuring satisfactory core behavior under accident conditions such as control rod ejection and unexpected valve opening on the secondary side. The advantages of this 100% MOX core compared with a 50% MOX core are discussed. This concept is fully compatible with the future European pressurized reactor (EPR). This 100% MOX HMR reactor could be the plutonium version of the EPR.

  7. Analytical support for the ORR (Oak Ridge Research Reactor) whole-core LEU U/sub 3/Si/sub 2/-Al fuel demonstration

    SciTech Connect

    Bretscher, M.M.

    1986-01-01

    Analytical methods used to analyze neutronic data from the whole-core LEU fuel demonstration in the Oak Ridge Research Reactor are briefly discussed. Calculated eigenvalues corresponding to measured critical control rod positions are presented for each core used in the gradual transition from an all HEU to an all LEU configuration. Some calculated and measured results, including ..beta../sub eff//l/sub p/, are compared for HEU and LEU fresh fuel criticals. Finally, the perturbing influences of the six voided beam tubes on certain core parameters are examined. For reasons yet to be determined, differential shim rod worths are not well-calculated in partially burned cores.

  8. A Technique to Determine Billet Core Charge Weight for P/M Fuel Tubes

    SciTech Connect

    Peacock, H.B.

    2001-07-02

    The core length in an extruded tube depends on the weight of powder in the billet core. In the past, the amount of aluminum powder needed to give a specified core length was determined empirically. This report gives a technique for calculating the weight of aluminum powder for the P/M core. An equation has been derived which can be used to determine the amount of aluminum needed for P/M billet core charge weights. Good agreement was obtained when compared to Mark 22 tube extrusion data. From the calculated charge weight, the elastomeric bag can be designed and made to compact the U3O8-Al core.

  9. The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

    NASA Astrophysics Data System (ADS)

    Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

    2017-01-01

    The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

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

    SciTech Connect

    Chodak, III, Paul

    1996-05-01

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

  11. Global optimization and oxygen dissociation on polyicosahedral Ag32Cu6 core-shell cluster for alkaline fuel cells

    NASA Astrophysics Data System (ADS)

    Zhang, N.; Chen, F. Y.; Wu, X. Q.

    2015-07-01

    The structure of 38 atoms Ag-Cu cluster is studied by using a combination of a genetic algorithm global optimization technique and density functional theory (DFT) calculations. It is demonstrated that the truncated octahedral (TO) Ag32Cu6 core-shell cluster is less stable than the polyicosahedral (pIh) Ag32Cu6 core-shell cluster from the atomistic models and the DFT calculation shows an agreeable result, so the newfound pIh Ag32Cu6 core-shell cluster is further investigated for potential application for O2 dissociation in oxygen reduction reaction (ORR). The activation energy barrier for the O2 dissociation on pIh Ag32Cu6 core-shell cluster is 0.715 eV, where the d-band center is -3.395 eV and the density of states at the Fermi energy level is maximal for the favorable absorption site, indicating that the catalytic activity is attributed to a maximal charge transfer between an oxygen molecule and the pIh Ag32Cu6 core-shell cluster. This work revises the earlier idea that Ag32Cu6 core-shell nanoparticles are not suitable as ORR catalysts and confirms that Ag-Cu nanoalloy is a potential candidate to substitute noble Pt-based catalyst in alkaline fuel cells.

  12. Global optimization and oxygen dissociation on polyicosahedral Ag32Cu6 core-shell cluster for alkaline fuel cells.

    PubMed

    Zhang, N; Chen, F Y; Wu, X Q

    2015-07-07

    The structure of 38 atoms Ag-Cu cluster is studied by using a combination of a genetic algorithm global optimization technique and density functional theory (DFT) calculations. It is demonstrated that the truncated octahedral (TO) Ag32Cu6 core-shell cluster is less stable than the polyicosahedral (pIh) Ag32Cu6 core-shell cluster from the atomistic models and the DFT calculation shows an agreeable result, so the newfound pIh Ag32Cu6 core-shell cluster is further investigated for potential application for O2 dissociation in oxygen reduction reaction (ORR). The activation energy barrier for the O2 dissociation on pIh Ag32Cu6 core-shell cluster is 0.715 eV, where the d-band center is -3.395 eV and the density of states at the Fermi energy level is maximal for the favorable absorption site, indicating that the catalytic activity is attributed to a maximal charge transfer between an oxygen molecule and the pIh Ag32Cu6 core-shell cluster. This work revises the earlier idea that Ag32Cu6 core-shell nanoparticles are not suitable as ORR catalysts and confirms that Ag-Cu nanoalloy is a potential candidate to substitute noble Pt-based catalyst in alkaline fuel cells.

  13. Global optimization and oxygen dissociation on polyicosahedral Ag32Cu6 core-shell cluster for alkaline fuel cells

    PubMed Central

    Zhang, N.; Chen, F. Y.; Wu, X.Q.

    2015-01-01

    The structure of 38 atoms Ag-Cu cluster is studied by using a combination of a genetic algorithm global optimization technique and density functional theory (DFT) calculations. It is demonstrated that the truncated octahedral (TO) Ag32Cu6 core-shell cluster is less stable than the polyicosahedral (pIh) Ag32Cu6 core-shell cluster from the atomistic models and the DFT calculation shows an agreeable result, so the newfound pIh Ag32Cu6 core-shell cluster is further investigated for potential application for O2 dissociation in oxygen reduction reaction (ORR). The activation energy barrier for the O2 dissociation on pIh Ag32Cu6 core-shell cluster is 0.715 eV, where the d-band center is −3.395 eV and the density of states at the Fermi energy level is maximal for the favorable absorption site, indicating that the catalytic activity is attributed to a maximal charge transfer between an oxygen molecule and the pIh Ag32Cu6 core-shell cluster. This work revises the earlier idea that Ag32Cu6 core-shell nanoparticles are not suitable as ORR catalysts and confirms that Ag-Cu nanoalloy is a potential candidate to substitute noble Pt-based catalyst in alkaline fuel cells. PMID:26148904

  14. User's guide for the REBUS-3 fuel cycle analysis capability

    SciTech Connect

    Toppel, B.J.

    1983-03-01

    REBUS-3 is a system of programs designed for the fuel-cycle analysis of fast reactors. This new capability is an extension and refinement of the REBUS-3 code system and complies with the standard code practices and interface dataset specifications of the Committee on Computer Code Coordination (CCCC). The new code is hence divorced from the earlier ARC System. In addition, the coding has been designed to enhance code exportability. Major new capabilities not available in the REBUS-2 code system include a search on burn cycle time to achieve a specified value for the multiplication constant at the end of the burn step; a general non-repetitive fuel-management capability including temporary out-of-core fuel storage, loading of fresh fuel, and subsequent retrieval and reloading of fuel; significantly expanded user input checking; expanded output edits; provision of prestored burnup chains to simplify user input; option of fixed-or free-field BCD input formats; and, choice of finite difference, nodal or spatial flux-synthesis neutronics in one-, two-, or three-dimensions.

  15. 9 CFR 325.18 - Diverting of shipments, breaking of seals, and reloading by carrier in emergency; reporting to...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... seals, and reloading by carrier in emergency; reporting to Regional Director. 325.18 Section 325.18... CERTIFICATION TRANSPORTATION § 325.18 Diverting of shipments, breaking of seals, and reloading by carrier in...) In case of wreck or similar extraordinary emergency, the Department seals on a railroad car or...

  16. 9 CFR 325.18 - Diverting of shipments, breaking of seals, and reloading by carrier in emergency; reporting to...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... seals, and reloading by carrier in emergency; reporting to Regional Director. 325.18 Section 325.18... CERTIFICATION TRANSPORTATION § 325.18 Diverting of shipments, breaking of seals, and reloading by carrier in...) In case of wreck or similar extraordinary emergency, the Department seals on a railroad car or...

  17. The experience of using Endo GIA™ Radial Reload with Tri-Staple™ Technology for various lung surgery.

    PubMed

    Ema, Toshinari

    2014-10-01

    Endo GIA™ Radial Reload with Tri-Staple™ Technology (RR) is a device for colorectal surgery. However, with its rounded staple line, Radial Reload is suitable for various lung surgeries. We use the device for lung wedge resection, and cutting bronchus in lung lobectomy. The total number of use counts up to 56 fires, and all fires came out well.

  18. Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

    SciTech Connect

    Savander, V. I.; Shumskiy, B. E.; Pinegin, A. A.

    2016-12-15

    The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

  19. Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

    NASA Astrophysics Data System (ADS)

    Savander, V. I.; Shumskiy, B. E.; Pinegin, A. A.

    2016-12-01

    The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

  20. Intracellular Ca2+ transients in mouse soleus muscle after hindlimb unloading and reloading

    NASA Technical Reports Server (NTRS)

    Ingalls, C. P.; Warren, G. L.; Armstrong, R. B.; Hamilton, S. L. (Principal Investigator)

    1999-01-01

    The objective of this study was to determine whether altered intracellular Ca(2+) handling contributes to the specific force loss in the soleus muscle after unloading and/or subsequent reloading of mouse hindlimbs. Three groups of female ICR mice were studied: 1) unloaded mice (n = 11) that were hindlimb suspended for 14 days, 2) reloaded mice (n = 10) that were returned to their cages for 1 day after 14 days of hindlimb suspension, and 3) control mice (n = 10) that had normal cage activity. Maximum isometric tetanic force (P(o)) was determined in the soleus muscle from the left hindlimb, and resting free cytosolic Ca(2+) concentration ([Ca(2+)](i)), tetanic [Ca(2+)](i), and 4-chloro-m-cresol-induced [Ca(2+)](i) were measured in the contralateral soleus muscle by confocal laser scanning microscopy. Unloading and reloading increased resting [Ca(2+)](i) above control by 36% and 24%, respectively. Although unloading reduced P(o) and specific force by 58% and 24%, respectively, compared with control mice, there was no difference in tetanic [Ca(2+)](i). P(o), specific force, and tetanic [Ca(2+)](i) were reduced by 58%, 23%, and 23%, respectively, in the reloaded animals compared with control mice; however, tetanic [Ca(2+)](i) was not different between unloaded and reloaded mice. These data indicate that although hindlimb suspension results in disturbed intracellular Ca(2+) homeostasis, changes in tetanic [Ca(2+)](i) do not contribute to force deficits. Compared with unloading, 24 h of physiological reloading in the mouse do not result in further changes in maximal strength or tetanic [Ca(2+)](i).

  1. Myocardial reloading after extracorporeal membrane oxygenation alters substrate metabolism while promoting protein synthesis.

    PubMed

    Kajimoto, Masaki; O'Kelly Priddy, Colleen M; Ledee, Dolena R; Xu, Chun; Isern, Nancy; Olson, Aaron K; Des Rosiers, Christine; Portman, Michael A

    2013-08-19

    Extracorporeal membrane oxygenation (ECMO) unloads the heart, providing a bridge to recovery in children after myocardial stunning. ECMO also induces stress which can adversely affect the ability to reload or wean the heart from the circuit. Metabolic impairments induced by altered loading and/or stress conditions may impact weaning. However, cardiac substrate and amino acid requirements upon weaning are unknown. We assessed the hypothesis that ventricular reloading with ECMO modulates both substrate entry into the citric acid cycle (CAC) and myocardial protein synthesis. Sixteen immature piglets (7.8 to 15.6 kg) were separated into 2 groups based on ventricular loading status: 8-hour ECMO (UNLOAD) and postwean from ECMO (RELOAD). We infused into the coronary artery [2-(13)C]-pyruvate as an oxidative substrate and [(13)C6]-L-leucine as an indicator for amino acid oxidation and protein synthesis. Upon RELOAD, each functional parameter, which were decreased substantially by ECMO, recovered to near-baseline level with the exclusion of minimum dP/dt. Accordingly, myocardial oxygen consumption was also increased, indicating that overall mitochondrial metabolism was reestablished. At the metabolic level, when compared to UNLOAD, RELOAD altered the contribution of various substrates/pathways to tissue pyruvate formation, favoring exogenous pyruvate versus glycolysis, and acetyl-CoA formation, shifting away from pyruvate decarboxylation to endogenous substrate, presumably fatty acids. Furthermore, there was also a significant increase of tissue concentrations for all CAC intermediates (≈80%), suggesting enhanced anaplerosis, and of fractional protein synthesis rates (>70%). RELOAD alters both cytosolic and mitochondrial energy substrate metabolism, while favoring leucine incorporation into protein synthesis rather than oxidation in the CAC. Improved understanding of factors governing these metabolic perturbations may serve as a basis for interventions and thereby improve

  2. Myocardial Reloading After Extracorporeal Membrane Oxygenation Alters Substrate Metabolism While Promoting Protein Synthesis

    PubMed Central

    Kajimoto, Masaki; O'Kelly Priddy, Colleen M.; Ledee, Dolena R.; Xu, Chun; Isern, Nancy; Olson, Aaron K.; Rosiers, Christine Des; Portman, Michael A.

    2013-01-01

    Background Extracorporeal membrane oxygenation (ECMO) unloads the heart, providing a bridge to recovery in children after myocardial stunning. ECMO also induces stress which can adversely affect the ability to reload or wean the heart from the circuit. Metabolic impairments induced by altered loading and/or stress conditions may impact weaning. However, cardiac substrate and amino acid requirements upon weaning are unknown. We assessed the hypothesis that ventricular reloading with ECMO modulates both substrate entry into the citric acid cycle (CAC) and myocardial protein synthesis. Methods and Results Sixteen immature piglets (7.8 to 15.6 kg) were separated into 2 groups based on ventricular loading status: 8‐hour ECMO (UNLOAD) and postwean from ECMO (RELOAD). We infused into the coronary artery [2‐13C]‐pyruvate as an oxidative substrate and [13C6]‐L‐leucine as an indicator for amino acid oxidation and protein synthesis. Upon RELOAD, each functional parameter, which were decreased substantially by ECMO, recovered to near‐baseline level with the exclusion of minimum dP/dt. Accordingly, myocardial oxygen consumption was also increased, indicating that overall mitochondrial metabolism was reestablished. At the metabolic level, when compared to UNLOAD, RELOAD altered the contribution of various substrates/pathways to tissue pyruvate formation, favoring exogenous pyruvate versus glycolysis, and acetyl‐CoA formation, shifting away from pyruvate decarboxylation to endogenous substrate, presumably fatty acids. Furthermore, there was also a significant increase of tissue concentrations for all CAC intermediates (≈80%), suggesting enhanced anaplerosis, and of fractional protein synthesis rates (>70%). Conclusions RELOAD alters both cytosolic and mitochondrial energy substrate metabolism, while favoring leucine incorporation into protein synthesis rather than oxidation in the CAC. Improved understanding of factors governing these metabolic perturbations may

  3. Temporal changes in sarcomere lesions of rat adductor longus muscles during hindlimb reloading

    NASA Technical Reports Server (NTRS)

    Krippendorf, B. B.; Riley, D. A.

    1994-01-01

    Focal sarcomere disruptions were previously observed in adductor longus muscles of rats flown approximately two weeks aboard the Cosmos 1887 and 2044 biosatellite flights. These lesions, characterized by breakage and loss of myofilaments and Z-line streaming, resembled damage induced by unaccustomed exercise that includes eccentric contractions in which muscles lengthen as they develop tension. We hypothesized that sarcomere lesions in atrophied muscles of space flow rats were not produced in microgravity by muscle unloading but resulted from muscle reloading upon re-exposure to terrestrial gravity. To test this hypothesis, we examined temporal changes in sarcomere integrity of adductor longus muscles from rats subjected to 12.5 days of hindlimb suspension unloading and subsequent reloading by return to vivarium cages for 0, 6, 12, or 48 hours of normal weightbearing. Our ultrastructural observations suggested that muscle unloading (0 h reloading) induced myofibril misalignment associated with myofiber atrophy. Muscle reloading for 6 hours induced focal sarcomere lesions in which cross striations were abnormally widened. Such lesions were electron lucent due to extensive myofilament loss. Lesions in reloaded muscles showed rapid restructuring. By 12 hours of reloading, lesions were moderately stained foci and by 48 hours darkly stained foci in which the pattern of cross striations was indistinct at the light and electron microscopic levels. These lesions were spanned by Z-line-like electron dense filamentous material. Our findings suggest a new role for Z-line streaming in lesion restructuring: rather than an antecedent to damage, this type of Z-line streaming may be indicative of rapid, early sarcomere repair.

  4. Intracellular Ca2+ transients in mouse soleus muscle after hindlimb unloading and reloading

    NASA Technical Reports Server (NTRS)

    Ingalls, C. P.; Warren, G. L.; Armstrong, R. B.; Hamilton, S. L. (Principal Investigator)

    1999-01-01

    The objective of this study was to determine whether altered intracellular Ca(2+) handling contributes to the specific force loss in the soleus muscle after unloading and/or subsequent reloading of mouse hindlimbs. Three groups of female ICR mice were studied: 1) unloaded mice (n = 11) that were hindlimb suspended for 14 days, 2) reloaded mice (n = 10) that were returned to their cages for 1 day after 14 days of hindlimb suspension, and 3) control mice (n = 10) that had normal cage activity. Maximum isometric tetanic force (P(o)) was determined in the soleus muscle from the left hindlimb, and resting free cytosolic Ca(2+) concentration ([Ca(2+)](i)), tetanic [Ca(2+)](i), and 4-chloro-m-cresol-induced [Ca(2+)](i) were measured in the contralateral soleus muscle by confocal laser scanning microscopy. Unloading and reloading increased resting [Ca(2+)](i) above control by 36% and 24%, respectively. Although unloading reduced P(o) and specific force by 58% and 24%, respectively, compared with control mice, there was no difference in tetanic [Ca(2+)](i). P(o), specific force, and tetanic [Ca(2+)](i) were reduced by 58%, 23%, and 23%, respectively, in the reloaded animals compared with control mice; however, tetanic [Ca(2+)](i) was not different between unloaded and reloaded mice. These data indicate that although hindlimb suspension results in disturbed intracellular Ca(2+) homeostasis, changes in tetanic [Ca(2+)](i) do not contribute to force deficits. Compared with unloading, 24 h of physiological reloading in the mouse do not result in further changes in maximal strength or tetanic [Ca(2+)](i).

  5. Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

    SciTech Connect

    Karel I. Kingrey

    2003-04-01

    This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.

  6. Design and synthesis of degradation-resistant core-shell catalysts for proton exchange membrane fuel cells

    NASA Astrophysics Data System (ADS)

    Koh, Joon-Ho; Abbaraju, Ravikanth; Parthasarathy, Preethy; Virkar, Anil V.

    2014-09-01

    Core@shell catalysts supported on carbon with Ag core and Pt shell were synthesized by a chemical process. EDX spectra confirmed the formation of Ag@Pt core-shell catalysts. High resolution transmission electron microscopy (HRTEM) revealed that the Pt-shell was epitaxially matched to the Ag core. Electrochemical tests on proton exchange membrane (PEM) fuel cells made with Ag@Pt core-shell catalysts as cathode exhibited comparable performance to cells made using commercial Pt-Co-based catalysts as cathodes. Theoretical work suggested that Ag@Pt catalysts should be more stable in the PEMFC applications compared to monolithic Pt catalysts because Pt shell in Ag@Pt catalysts exhibits lower chemical potential of Pt than in monolithic Pt catalysts, thus reducing tendency for dissolution and Ostwald ripening. Lower chemical potential of Pt in the shell is attributed to larger lattice parameter of Ag compared to Pt, which puts the Pt shell in biaxial tension or reduced biaxial compression as compared to monolithic Pt catalysts. Preliminary out-of-cell tests show Ag@Pt catalysts to be stable in an environment containing ionic platinum.

  7. Verification study of thorium cross section in MVP calculation of thorium based fuel core using experimental data

    SciTech Connect

    Mai, V. T.; Fujii, T.; Wada, K.; Kitada, T.; Takaki, N.; Yamaguchi, A.; Watanabe, H.; Unesaki, H.

    2012-07-01

    Considering the importance of thorium data and concerning about the accuracy of Th-232 cross section library, a series of experiments of thorium critical core carried out at KUCA facility of Kyoto Univ. Research Reactor Inst. have been analyzed. The core was composed of pure thorium plates and 93% enriched uranium plates, solid polyethylene moderator with hydro to U-235 ratio of 140 and Th-232 to U-235 ratio of 15.2. Calculations of the effective multiplication factor, control rod worth, reactivity worth of Th plates have been conducted by MVP code using JENDL-4.0 library [1]. At the experiment site, after achieving the critical state with 51 fuel rods inserted inside the reactor, the measurements of the reactivity worth of control rod and thorium sample are carried out. By comparing with the experimental data, the calculation overestimates the effective multiplication factor about 0.90%. Reactivity worth of the control rods evaluation using MVP is acceptable with the maximum discrepancy about the statistical error of the measured data. The calculated results agree to the measurement ones within the difference range of 3.1% for the reactivity worth of one Th plate. From this investigation, further experiments and research on Th-232 cross section library need to be conducted to provide more reliable data for thorium based fuel core design and safety calculation. (authors)

  8. Fuel containment and stability in the gas core nuclear rocket. Final report, April 15, 1993--April 14, 1994

    SciTech Connect

    Kammash, T.

    1996-02-01

    One of the most promising approaches to advanced propulsion that could meet the objectives of the Space Exploration Initiative (SEI) is the open cycle gas core nuclear rocket (GCR). The energy in this device is generated by a fissioning uranium plasma which heats, through radiation, a propellant that flows around the core and exits through a nozzle, thereby converting thermal energy into thrust. Although such a scheme can produce very attractive propulsion parameters in the form of high specific impulse and high thrust, it does suffer from serious physics and engineering problems that must be addressed if it is to become a viable propulsion system. Among the major problems that must be solved are the confinement of the uranium plasma, potential instabilities and control problems associated with the dynamics of the uranium core, and the question of startup and fueling of such a reactor. In this paper, the authors focus their attention on the problems of equilibria and stability of the uranium care, and examine the potential use of an externally applied magnetic field for these purposes. They find that steady state operation of the reactor is possible only for certain care profiles that may not be compatible with the radiative aspect of the system. The authors also find that the system is susceptible to hydrodynamic and acoustic instabilities that could deplete the uranium fuel in a short time if not properly suppressed.

  9. A1-U fuel foaming/recriticality considerations for production reactor core-melt accidents

    SciTech Connect

    Cronenberg, A.W. ); Hyder, M.L.; Ellison, P.G. )

    1990-01-01

    Severe accident studies for the Savannah River production reactors indicate that if coherent fuel melting and relocation occur in the absence of target melting, in-vessel recriticality may be achieved. In this paper, fuel-melt/target interaction potential is assessed, where fission gas-induced fuel foaming and melt attach on target material are evaluated and compared with available data. Models are developed to characterize foams for irradiated Al-based fuel. Predictions indicate transient foaming (the extent of which is governed by fission gas inventory), heating transient, and bubble coalescence behavior. The model also indicates that metallic foams are basically unstable and will collapse, which largely depends on film tenacity and melt viscosity. For high-burnup fuel, foams lasting tens of seconds are predicted, allowing molten fuel to contact and cause melt ablation of concentric targets. For low-burnup fuel, contact can not be assured, thus recriticality may be of concern at reactor startup. 8 refs., 4 figs., 4 tabs.

  10. Fission product retention in TRISCO coated UO sub 2 particle fuels subjected to HTR simulated core heating tests

    SciTech Connect

    Baldwin, C.A.; Kania, M.J.

    1990-11-01

    Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600{degree}C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800{degree}C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800{degree}C and above may exist. 6 refs., 6 figs., 4 tabs.

  11. A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

    SciTech Connect

    Senor, David J.; Painter, Chad L.; Geelhood, Ken J.; Wootan, David W.; Meriwether, George H.; Cuta, Judith M.; Adkins, Harold E.; Matson, Dean W.; Abrego, Celestino P.

    2007-12-01

    Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling, core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.

  12. Comprehensive Measurements and Modeling of SOL, and Core Plasma Fueling and Carbon Sources in DIII-D

    SciTech Connect

    Groth, M; Porter, G; Bray, B; Brooks, N; Fenstermacher, M; Groebner, R; Lasnier, C; Meyer, W; Rensink, M; Rudakov, D; Watkins, J; Zeng, L

    2005-06-24

    Plasma boundary modeling of low density, low confinement plasmas in DIII-D has been benchmarked against a comprehensive set of measurements and indicates that recycling of deuterium ions at the divertor targets, and chemical sputtering at the divertor target plates and walls, can explain the poloidal core fueling profile and core carbon density. Key measurements included the 2-D intensity distribution of deuterium neutral and low-charge state carbon emission in the divertor and around the midplane of the high-field scrape-off layer (SOL). Chemical sputtering plays an important role in producing carbon at the divertor targets and walls, and was found to be a prerequisite to reproduce the measured emission distribution.

  13. The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

    SciTech Connect

    Dunn, Floyd E.; Hu, Lin-wen; Wilson, Erik

    2016-12-01

    The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.

  14. 2D Mesoscale Simulations of Quasielastic Reloading and Unloading in Shock Compressed Aluminum

    NASA Astrophysics Data System (ADS)

    Dwivedi, S. K.

    2007-06-01

    2D mesoscale simulations of planar shock compression, followed by either reloading or unloading, are presented that predict quasi-elastic (QE) response observed experimentally in shocked polycrystalline aluminum. The representative volume element (RVE) of the plate impact experiments included a realistic representation of a grain ensemble with apparent heterogeneities in the polycrystalline sample. Simulations were carried out using a 2D updated Lagrangian finite element code ISP-TROTP incorporating elastic-plastic deformation in grain interior and contact/cohesive methodology to analyze finite strength grain boundaries. Local heterogeneous response was quantified by calculating appropriate material variables along in-situ Lagrangian tracer lines and comparing the temporal variation of their mean values with results from 2D continuum simulations. Simulations were carried out by varying a large number of individual heterogeneities to predict QE response on reloading and unloading from shock state. The heterogeneities important for simulating the QE response identified from these simulations were: hardened grain boundaries, hard inclusions, and micro-porosity. It is shown that the shock-deformed state of polycrystalline aluminum in the presence of these effects is strongly heterogeneous with considerable variations in lateral stresses. This distributed stress state unloads the shear stress from flow stress causing QE response on reloading as well as unloading. The simulated velocity profiles and calculated shear strength and shear stresses for a representative reloading and unloading experimental configuration were found to agree well with the reported experimental data. Work supported by DOE.

  15. Extended Burnup Demonstration Reactor Fuels Program. Annual progress report, April 1983-March 1984. [BWR

    SciTech Connect

    Exarhos, C.A.

    1985-06-20

    The US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities Nuclear Corporation have participated since 1979 in a cooperative Extended Burnup Demonstration Program. Under the program, standard ENC-fabricated reload fuel in the Big Rock Point and Oyster Creek reactor cores has been irradiated to discharge burnups at or beyond 35,000 MWD/MTU, one to two cycles beyond its originally projected exposure life. The program provides for examination of the fuel at poolside before and after each extended burnup cycle as well as for limited destructive hot cell examination. The 1984 progress report covers work performed under the EBD program between April 1983 and March 1984. Major milestones reached during the period include completion of a hot cell examination on four high burnup rods from Big Rock Point and of a poolside on the Oyster Creek EBD fuel at discharge. The hot cell examination of four rods at burnups to 37.2 GWD/MTU confirmed poolside measurements on the same fuel, showing the urania and gadolinia-bearing fuel rods to be in excellent condition. No major cladding degradation, pellet restructuring, or pellet-clad interaction was found in any of the samples examined. The Oyster Creek fuel, examined at an assembly average exposure of 34.5 GWD/MTU, showed good performance with regard to both diametral creepdown and clad oxide accumulation.

  16. Fuel efficient hydrodynamic containment for gas core fission reactor rocket propulsion. Final report, September 30, 1992--May 31, 1995

    SciTech Connect

    Sforza, P.M.; Cresci, R.J.

    1997-05-31

    Gas core reactors can form the basis for advanced nuclear thermal propulsion (NTP) systems capable of providing specific impulse levels of more than 2,000 sec., but containment of the hot uranium plasma is a major problem. The initial phase of an experimental study of hydrodynamic confinement of the fuel cloud in a gas core fission reactor by means of an innovative application of a base injection stabilized recirculation bubble is presented. The development of the experimental facility, a simulated thrust chamber approximately 0.4 m in diameter and 1 m long, is described. The flow rate of propellant simulant (air) can be varied up to about 2 kg/sec and that of fuel simulant (air, air-sulfur hexafluoride) up to about 0.2 kg/sec. This scale leads to chamber Reynolds numbers on the same order of magnitude as those anticipated in a full-scale nuclear rocket engine. The experimental program introduced here is focused on determining the size, geometry, and stability of the recirculation region as a function of the bleed ratio, i.e. the ratio of the injected mass flux to the free stream mass flux. A concurrent CFD study is being carried out to aid in demonstrating that the proposed technique is practical.

  17. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    NASA Astrophysics Data System (ADS)

    Borodkin, Pavel; Borodkin, Gennady; Khrennikov, Nikolay

    2016-02-01

    Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  18. X-ray continuum as a measure of pressure and fuel-shell mix in compressed isobaric hydrogen implosion cores

    NASA Astrophysics Data System (ADS)

    Epstein, R.; Goncharov, V. N.; Marshall, F. J.; Betti, R.; Nora, R.; Christopherson, A. R.; Golovkin, I. E.; MacFarlane, J. J.

    2015-02-01

    Pressure, by definition, characterizes the conditions within an isobaric implosion core at peak compression [Gus'kov et al., Nucl. Fusion 16, 957 (1976); Betti et al., Phys. Plasmas 8, 5257 (2001)] and is a key parameter in quantifying its near-ignition performance [Lawson, Proc. Phys. Soc. London, B 70, 6 (1957); Betti et al., Phys. Plasmas 17, 058102 (2010); Goncharov et al., Phys. Plasmas 21, 056315 (2014); and Glenzer et al., Phys. Plasmas 19, 056318 (2012)]. At high spectral energy, where the x-ray emission from an imploded hydrogen core is optically thin, the emissivity profile can be inferred from the spatially resolved core emission. This emissivity, which can be modeled accurately under hot-core conditions, is dependent almost entirely on the pressure when measured within a restricted spectral range matched to the temperature range anticipated for the emitting volume. In this way, the hot core pressure at the time of peak emission can be inferred from the measured free-free emissivity profile. The pressure and temperature dependences of the x-ray emissivity and the neutron-production rate explain a simple scaling of the total filtered x-ray emission as a constant power of the total neutron yield for implosions of targets of similar design over a broad range of shell implosion isentropes. This scaling behavior has been seen in implosion simulations and is confirmed by measurements of high-isentrope implosions [Sangster et al., Phys. Plasmas 20, 056317 (2013)] on the OMEGA laser system [Boehly et al., Opt. Commun. 133, 495 (1997)]. Attributing the excess emission from less-stable, low-isentrope implosions, above the level expected from this neutron-yield scaling, to the higher emissivity of shell carbon mixed into the implosion's central hot spot, the hot-spot "fuel-shell" mix mass can be inferred.

  19. Pt@Pd(x)Cu(y)/C core-shell electrocatalysts for oxygen reduction reaction in fuel cells.

    PubMed

    Cochell, T; Manthiram, A

    2012-01-17

    A series of carbon-supported core-shell nanoparticles with Pd(x)Cu(y)-rich cores and Pt-rich shells (Pt@Pd(x)Cu(y)/C) has been synthesized by a polyol reduction of the precursors followed by heat treatment to obtain the Pd(x)Cu(y)/C (1 ≤ x ≤ 3 and 0 ≤ y ≤ 5) cores and the galvanic displacement of Pd(x)Cu(y) with [PtCl(4)](2-) to form the Pt shell. The nanoparticles have also been investigated with respect to the oxygen reduction reaction (ORR) in proton-exchange-membrane fuel cells (PEMFCs). X-ray diffraction (XRD) analysis suggests that the cores are highly alloyed and that the galvanic displacement results in a certain amount of alloying between Pt and the underlying Pd(x)Cu(y) alloy core. Transmission electron microscopy (TEM) images show that the Pt@Pd(x)Cu(y)/C catalysts (where y > 0) have mean particle sizes of <8 nm. Compositional analysis by energy-dispersive X-ray spectroscopy (EDS) and X-ray photoelectron spectroscopy (XPS) clearly shows Pt enrichment in the near-surface region of the nanoparticles. Cyclic voltammograms show a positive shift of as much as 40 mV for the onset of Pt-OH formation in the Pt@Pd(x)Cu(y)/C electrocatalysts compared to that in Pt/C. Rotating disk electrode (RDE) measurements of Pt@PdCu(5)/C show an increase in the Pt mass activity by 3.5-fold and noble metal activity by 2.5-fold compared to that of Pt/C. The activity enhancements in RDE and PEMFC measurements are believed to be a result of the delay in the onset of Pt-OH formation. © 2011 American Chemical Society

  20. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    SciTech Connect

    John D. Bess; Barbara H. Dolphin; James W. Sterbentz; Luka Snoj; Igor Lengar; Oliver Köberl

    2013-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.

  1. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    SciTech Connect

    John D. Bess; Barbara H. Dolphin; James W. Sterbentz; Luka Snoj; Igor Lengar; Oliver Köberl

    2012-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.

  2. Shipment of the Light Water Breeder Reactor fuel assemblies from the Shippingport Atomic Power Station to the extended core facility (Idaho) (LWBR Development Program)

    SciTech Connect

    Selsley, I.A.

    1987-10-01

    After successfully operating for 29,047 effective full power hours, the Light Water Breeder Reactor (LWBR) core was defueled prior to total decommissioning of the Shippingport Atomic Power Station. All nuclear fuel and much of the reactor internal hardware was removed from the reactor vessel and prepared for shipment to disposal sites or to the Naval Reactors Expended Core Facility in Idaho for testing or further disassembly. Three M-130 shipping containers were modified to accept LWBR seed, blanket, and reflector fuel modules for rail shipment to the Expended Core Facility. Thirty-nine LWBR fuel modules were transferred in 10 shipments. All shipments were completed successfully, without significant problems. Radiation and personnel exposure levels were carefully controlled.

  3. Mineralogic and petrologic investigation of post-test core samples from the Spent Fuel Test - Climax

    SciTech Connect

    Ryerson, F.J.; Beiriger, J.

    1985-02-01

    We have characterized a suite of samples taken subsequent to the end of the Spent Fuel Test - Climax by petrographic and microanalytical techniques and determined their mineral assemblage, modal properties, and mineral chemistry. The samples were obtained immediately adjacent to the canister borehole at a variety of depths and positions within the canister drift, as well as radially outward from each canister hole. This method of sampling allows variations in post-test mineralogic properties to be evaluated on the basis of (1) depth along a particular canister hole and (2) position within the canister drift, with respect to the heat and radiation sources, and with respect to the pre - test samples. In no case did we find any significant correlation between the mineralogical properties and variables listed above. In short, the Spent Fuel Test - Climax has produced no identifiable mineralogical response in the Climax quartz monzonite. 12 refs., 11 figs., 5 tabs.

  4. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

    SciTech Connect

    Michael A. Pope

    2011-10-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  5. Insulin-like growth factor-1 receptor in mature osteoblasts is required for periosteal bone formation induced by reloading.

    PubMed

    Kubota, Takuo; Elalieh, Hashem Z; Saless, Neema; Fong, Chak; Wang, Yongmei; Babey, Muriel; Cheng, Zhiqiang; Bikle, Daniel D

    2013-11-01

    Skeletal loading and unloading has a pronounced impact on bone remodeling, a process also regulated by insulin-like growth factor 1 (IGF-1) signaling. Skeletal unloading leads to resistance to the anabolic effect of IGF-1, while reloading after unloading restores responsiveness to IGF-1. However, a direct study of the importance of IGF-1 signaling in the skeletal response to mechanical loading remains to be tested. In this study, we assessed the skeletal response of osteoblast-specific Igf-1 receptor deficient (Igf-1r(-/-) ) mice to unloading and reloading. The mice were hindlimb unloaded for 14 days and then reloaded for 16 days. Igf-1r(-/-) mice displayed smaller cortical bone and diminished periosteal and endosteal bone formation at baseline. Periosteal and endosteal bone formation decreased with unloading in Igf-1r(+/+) mice. However, the recovery of periosteal bone formation with reloading was completely inhibited in Igf-1r(-/-) mice, although reloading-induced endosteal bone formation was not hampered. These changes in bone formation resulted in the abolishment of the expected increase in total cross-sectional area with reloading in Igf-1r(-/-) mice compared to the control mice. These results suggest that the Igf-1r in mature osteoblasts has a critical role in periosteal bone formation in the skeletal response to mechanical loading.

  6. Insulin-like growth factor-1 receptor in mature osteoblasts is required for periosteal bone formation induced by reloading

    NASA Astrophysics Data System (ADS)

    Kubota, Takuo; Elalieh, Hashem Z.; Saless, Neema; Fong, Chak; Wang, Yongmei; Babey, Muriel; Cheng, Zhiqiang; Bikle, Daniel D.

    2013-11-01

    Skeletal loading and unloading has a pronounced impact on bone remodeling, a process also regulated by insulin-like growth factor-1 (IGF-1) signaling. Skeletal unloading leads to resistance to the anabolic effect of IGF-1, while reloading after unloading restores responsiveness to IGF-1. However, a direct study of the importance of IGF-1 signaling in the skeletal response to mechanical loading remains to be tested. In this study, we assessed the skeletal response of osteoblast-specific Igf-1 receptor deficient (Igf-1r-/-) mice to unloading and reloading. The mice were hindlimb unloaded for 14 days and then reloaded for 16 days. Igf-1r-/- mice displayed smaller cortical bone and diminished periosteal and endosteal bone formation at baseline. Periosteal and endosteal bone formation decreased with unloading in Igf-1r+/+ mice. However, the recovery of periosteal bone formation with reloading was completely inhibited in Igf-1r-/- mice, although reloading-induced endosteal bone formation was not hampered. These changes in bone formation resulted in the abolishment of the expected increase in total cross-sectional area with reloading in Igf-1r-/- mice compared to the control mice. These results suggest that the Igf-1r in mature osteoblasts has a critical role in periosteal bone formation in the skeletal response to mechanical loading.

  7. Effects of Unloading and Reloading on Expressions of Skelatal Muscle Membrane Proteins in Mice

    NASA Astrophysics Data System (ADS)

    Ohno, Y.; Ikuta, A.; Goto, A.; Sugiura, T.; Ohira, Y.; Yoshioka, T.; Goto, K.

    2013-02-01

    Effects of unloading and reloading on the expression levels of tripartite motif-containing 72 (TRIM72) and caveolin-3 (Cav-3) of soleus muscle in mice were investigated. Male C57BL/6J mice (11-week old) were randomly assigned to control and hindlimb-suspended groups. Some of mice in hindlimb-suspended group were subjected to continuous hindlimb suspension (HS) for 2 weeks with or without 7 days of ambulation recovery. Following HS, the muscle weight and protein expression levels of TRIM72 and Cav-3 in soleus were decreased. On the other hand, the gradual increases in muscle mass, TRIM72 and Cav-3 were observed after reloading following HS. Therefore, it was suggested that mechanical loading played a key role in a regulatory system for protein expressions of TRIM72 and Cav-3.

  8. Real-Time Scheduling in Heterogeneous Systems Considering Cache Reload Time Using Genetic Algorithms

    NASA Astrophysics Data System (ADS)

    Miryani, Mohammad Reza; Naghibzadeh, Mahmoud

    Since optimal assignment of tasks in a multiprocessor system is, in almost all practical cases, an NP-hard problem, in recent years some algorithms based on genetic algorithms have been proposed. Some of these algorithms have considered real-time applications with multiple objectives, total tardiness, completion time, etc. Here, we propose a suboptimal static scheduler of nonpreemptable tasks in hard real-time heterogeneous multiprocessor systems considering time constraints and cache reload time. The approach makes use of genetic algorithm to minimize total completion time and number of processors used, simultaneously. One important issue which makes this research different from previous ones is cache reload time. The method is implemented and the results are compared against a similar method.

  9. Anodic behavior of carbon supported Cu@Ag core-shell nanocatalysts in direct borohydride fuel cells

    NASA Astrophysics Data System (ADS)

    Duan, Donghong; Liu, Huihong; You, Xiu; Wei, Huikai; Liu, Shibin

    2015-10-01

    Carbon-supported Cu@Ag core-shell nanoparticles are prepared by a successive reduction method in an aqueous solution and are used as an anode electrocatalyst for the direct borohydride-hydrogen peroxide fuel cell (DBHFC). The physical and electrochemical properties of the as-prepared electrocatalysts are investigated by transmission electron microscopy (TEM), X-ray diffraction (XRD), cyclic voltammetry (CV), chronopotentiometry (CP), and fuel cell tests. In situ Fourier transform infrared (FTIR) spectroscopy is employed in 2 M NaOH/0.1 M NaBH4 to understand the borohydride oxidation reaction (BOR) mechanism by studying the intermediate reactions occurring on the Cu@Ag/C electrode. The TEM images show that the average size of the Cu1@Ag1/C particles is approximately 18 nm. Among the as-prepared catalysts, the Cu2@Ag1/C catalyst presents the highest catalytic activity. As shown by in situ FTIR, the oxidation reaction mechanism of BH4- is similar to that of Ag/C: BHn(OH)4-n- + 2OH- → BHn-1(OH)5-n- +H2 O + 2e . At 25 °C, the DBHFC with Cu2@Ag1/C as the anode electrocatalyst and Pt mesh (1 cm2) as the cathode electrode exhibits a maximum anodic power density of 17.27 mW mg-1 at a discharge current density of 27.8 mA mg-1.

  10. Evaluation of accuracy of calculations of VVER-1000 core states with incomplete covering of fuel by the absorber

    SciTech Connect

    Tikhomirov, A. V.; Ponomarenko, G. L.

    2012-07-01

    An additional verification of bundled software (BS) SAPFIR-95 and amp;RC [1] and code KORSAR/GP [2] was performed. Both software products were developed in A.P. Alexandrov NITI and certified by ROSTEKHNADZOR of RF for numeric simulation of stationary, transitional and emergency conditions of VVER reactors. A benchmark model for neutronics calculations was created within the limits of this work. The cold subcritical state of VVER - 1000 reactor stationary fuelling was simulated on the basis of FA with an increased height of the fuel column (TVS-2M) considering detailed presentation of radial and front neutron reflectors. A case of passing of pure condensate slug through the core in initially deep subcritical state during start of the first RCP set after refueling was considered as an examined condition of reactor operation. A relatively small size of the slug, its spatial position near the reflectors (lower and lateral), as well as failure of the inserted control rods of the control and protection system (CPS CR) to reach the lower limit of the fuel column stipulate for methodical complexity of a correct calculation of the neutron multiplication constant (K{sub eff}) using engineering codes. Code RC was used as a test program in the process of reactor calculated 3-D modeling. Code MCNP5 [3] was used as the precision program, which solves the equation of neutrons transfer by Monte-Carlo method and which was developed in the US (Los-Alamos). As a result of comparative calculations dependency of K{sub eff} on two parameters was evaluated - boron acid concentration (Cb) and CPS CR position. Reactivity effect was evaluated, which is implemented as a result of failure of all CPS control rods to reach the lower fuel limit calculated using the engineering codes mentioned above. (authors)

  11. Action potential duration determines sarcoplasmic reticulum Ca2+ reloading in mammalian ventricular myocytes

    PubMed Central

    Bassani, Rosana A; Altamirano, Julio; Puglisi, José L; Bers, Donald M

    2004-01-01

    After sarcoplasmic reticulum (SR) Ca2+ depletion in intact ventricular myocytes, electrical activity promotes SR Ca2+ reloading and recovery of twitch amplitude. In ferret, recovery of twitch and caffeine-induced contracture required fewer twitches than in rabbit or rat. In rat, there was no difference in action potential duration at 90% repolarization (APD90) at steady state (SS) versus at the first post-depletion (PD) twitch. The SS APD90 was similar in ferret and rabbit (but longer than in rat). However, compared to SS, the PD APD90 was lengthened in ferret, but shortened in rabbit. When rabbit myocytes were subjected to AP-clamp patterns during SR Ca2+ reloading (ferret- or rabbit-type APs), reloading was much faster using the ferret AP templates. We conclude that the faster SR Ca2+ refilling in ferret is due to the increased Ca2+ influx during the longer PD AP. The PD versus SS APD90 difference was suppressed by thapsigargin in ferret (indicating Ca2+ dependence). In rabbit, the PD AP shortening depended on the preceding diastolic interval (rather than Ca2+), because rest produced the same AP shortening, and SS APD90 increased as a function of frequency (in contrast to ferret). Transient outward current (Ito) was larger and recovered from inactivation much faster in ferret than in rabbit. Moreover, slow Ito recovery (τ ∼ 3 s) in rabbit was a much larger fraction of Ito. Our data and a computational model (including two Ito components) suggest that in rabbit the slowly recovering Ito is responsible for short post-rest and PD APs, for the unusual frequency dependence of APD90, and ultimately for the slower post-depletion SR Ca2+ reloading. PMID:15243136

  12. Kinetics and Muscle Activity Patterns during Unweighting and Reloading Transition Phases in Running

    PubMed Central

    Sainton, Patrick; Nicol, Caroline; Cabri, Jan; Barthèlemy-Montfort, Joëlle; Chavet, Pascale

    2016-01-01

    Amongst reduced gravity simulators, the lower body positive pressure (LBPP) treadmill is emerging as an innovative tool for both rehabilitation and fundamental research purposes as it allows running while experiencing reduced vertical ground reaction forces. The appropriate use of such a treadmill requires an improved understanding of the associated neuromechanical changes. This study concentrates on the runner’s adjustments to LBPP-induced unweighting and reloading during running. Nine healthy males performed two running series of nine minutes at natural speed. Each series comprised three sequences of three minutes at: 100% bodyweight (BW), 60 or 80% BW, and 100% BW. The progressive unweighting and reloading transitions lasted 10 to 15 s. The LBPP-induced unweighting level, vertical ground reaction force and center of mass accelerations were analyzed together with surface electromyographic activity from 6 major lower limb muscles. The analyses of stride-to-stride adjustments during each transition established highly linear relationships between the LBPP-induced progressive changes of BW and most mechanical parameters. However, the impact peak force and the loading rate systematically presented an initial 10% increase with unweighting which could result from a passive mechanism of leg retraction. Another major insight lies in the distinct neural adjustments found amongst the recorded lower-limb muscles during the pre- and post-contact phases. The preactivation phase was characterized by an overall EMG stability, the braking phase by decreased quadriceps and soleus muscle activities, and the push-off phase by decreased activities of the shank muscles. These neural changes were mirrored during reloading. These neural adjustments can be attributed in part to the lack of visual cues on the foot touchdown. These findings highlight both the rapidity and the complexity of the neuromechanical changes associated with LBPP-induced unweighting and reloading during running

  13. Agonist-sensitive calcium pool in the pancreatic acinar cell. II. Characterization of reloading

    SciTech Connect

    Muallem, S.; Schoeffield, M.S.; Fimmel, C.J.; Pandol, S.J.

    1988-08-01

    45Ca2+ fluxes and free cytosolic Ca2+ measurements in guinea pig pancreatic acini indicated that after agonist stimulation and the release of Ca2+ from the agonist-sensitive pool at least part of the Ca2+ is extruded from the cell, resulting in 45Ca2+ efflux. In the continued presence of agonist, the pool remains permeable to Ca2+ but partially refills with Ca2+. This reloading is dependent on the concentration of extracellular Ca2+. In the absence of extracellular Ca2+, the pool is completely depleted of Ca2+. However, with increasing concentrations of CaCl2 in the incubation solution (from 0.5 to 2.0 mM) there is increasing repletion of the pool with Ca2+ during agonist stimulation. With termination of agonist stimulation, the Ca2+ permeability of the agonist-sensitive pool is rapidly reduced to that measured in the unstimulated cell. As a result, the Ca2+ incorporated into the pool during the stimulation period is rapidly trapped within the pool and exchanges poorly with medium Ca2+. Subsequently, the pool completely refills with Ca2+. The rate of Ca2+ reloading at the termination of agonist stimulation is slower than the conversion of the pool to the impermeable state. In incubation media containing 1.3 mM CaCl2, the half-time for reloading at the termination of stimulation is 5 min. These observations demonstrate the characteristics of Ca2+ reloading of the agonist-sensitive pool both during stimulation and at the termination of stimulation.

  14. Logical Reloading. What is it and What is a Profit from it?

    NASA Astrophysics Data System (ADS)

    Rylov, Yuri A.

    2014-07-01

    Logical reloading is a replacement of basic statements of a conception by equivalent statements of the same conception. The logical reloading does not change the conception, but it changes the mathematical formalism and changes results of this conception generalization. In the paper two examples of the logical reloading are considered. (1) Generalization of the deterministic particle dynamics on the case of the stochastic particle dynamics. As a result the unified formalism for description of particles of all kinds appears. This formalism admits one to explain freely quantum dynamics in terms of the classical particle dynamics. In particular, one discovers κ-field responsible for pair production. (2) Generalization of the proper Euclidean geometry which contains such space-time geometries, where free particles move stochastically. As a result such a conception of elementary particle dynamics arises, where one can investigate the elementary particles arrangement, but not only systematize elementary particles, ascribing quantum numbers to them. Besides, one succeeds to expand the general relativity on the non-Riemannian space-time geometries.

  15. Osteocyte-viability-based simulations of trabecular bone loss and recovery in disuse and reloading.

    PubMed

    Wang, Hong; Ji, Baohua; Liu, X Sherry; van Oers, René F M; Guo, X Edward; Huang, Yonggang; Hwang, Keh-Chih

    2014-01-01

    Osteocyte apoptosis is known to trigger targeted bone resorption. In the present study, we developed an osteocyte-viability-based trabecular bone remodeling (OVBR) model. This novel remodeling model, combined with recent advanced simulation methods and analysis techniques, such as the element-by-element 3D finite element method and the ITS technique, was used to quantitatively study the dynamic evolution of bone mass and trabecular microstructure in response to various loading and unloading conditions. Different levels of unloading simulated the disuse condition of bed rest or microgravity in space. The amount of bone loss and microstructural deterioration correlated with the magnitude of unloading. The restoration of bone mass upon the reloading condition was achieved by thickening the remaining trabecular architecture, while the lost trabecular plates and rods could not be recovered by reloading. Compared to previous models, the predictions of bone resorption of the OVBR model are more consistent with physiological values reported from previous experiments. Whereas osteocytes suffer a lack of loading during disuse, they may suffer overloading during the reloading phase, which hampers recovery. The OVBR model is promising for quantitative studies of trabecular bone loss and microstructural deterioration of patients or astronauts during long-term bed rest or space flight and thereafter bone recovery.

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

    SciTech Connect

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

    2009-02-23

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

  17. Development and Implementation of a Newton-BICGSTAB Iterative Solver in the FORMOSA-B BWR Core Simulator Code

    SciTech Connect

    Kastanya, Doddy Yozef Febrian; Turinsky, Paul J.

    2005-05-15

    A Newton-Krylov iterative solver has been developed to reduce the CPU execution time of boiling water reactor (BWR) core simulators implemented in the core simulator part of the Fuel Optimization for Reloads Multiple Objectives by Simulated Annealing for BWR (FORMOSA-B) code, which is an in-core fuel management optimization code for BWRs. This new solver utilizes Newton's method to explicitly treat strong nonlinearities in the problem, replacing the traditionally used nested iterative approach. Newton's method provides the solver with a higher-than-linear convergence rate, assuming that good initial estimates of the unknowns are provided. Within each Newton iteration, an appropriately preconditioned Krylov solver is utilized for solving the linearized system of equations. Taking advantage of the higher convergence rate provided by Newton's method and utilizing an efficient preconditioned Krylov solver, we have developed a Newton-Krylov solver to evaluate the three-dimensional, two-group neutron diffusion equations coupled with a two-phase flow model within a BWR core simulator. Numerical tests on the new solver have shown that speedups ranging from 1.6 to 2.1, with reference to the traditional approach of employing nested iterations to treat the nonlinear feedbacks, can be achieved. However, if a preconditioned Krylov solver is employed to complete the inner iterations of the traditional approach, negligible CPU time differences are noted between the Newton-Krylov and traditional (Krylov) approaches.

  18. Robust synthesis of green fuels from biomass-derived ethyl esters over a hierarchically core/shell-structured ZSM-5@(Co/SiO2) catalyst.

    PubMed

    Wang, Darui; Wang, Bo; Ding, Yu; Yuan, Qingqing; Wu, Haihong; Guan, Yejun; Wu, Peng

    2017-09-12

    A novel bifunctional ZSM-5@(Co/SiO2) material with a hierarchical core/shell structure was successfully prepared through a simple chemoselective interaction between the crystal surface silica species of zeolite and the external Co(2+) source in basic media, which served as an excellent catalyst in the synthesis of green fuels from biomass-derived ethyl esters.

  19. SAS2H input for computing core activities of 4.5, 5.0, and 5.5 weight % {sup 235}U fuel for Sequoyah Nuclear Plant

    SciTech Connect

    Hermann, O.W.

    1994-08-01

    Sequoyah Nuclear Plant core activities at initial fuel enrichments of 4.5, 5.0, and 5.5 wt% {sup 235}U, required in nuclear safety evaluations, were computed by the SAS2H analysis sequence and the ORIGEN-S code within the SCALE-4.2 code system.

  20. Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

    SciTech Connect

    Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh; Ba Vien Luong; Kien Cuong Nguyen

    2008-07-15

    The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configuration with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)

  1. Moving Spent Fuel and Wastes in the USA: Concentrating on Core Competencies

    SciTech Connect

    Roland, V. H.; Gallo, B.

    2003-02-26

    With the study progress toward a US national reporting at Yucca Mountain, organizing the transport system becomes a more pressing necessity. The present paper will try to provide operators whose core business is not transport a frame of reference through which a transport supplier/system can be selected, not only on the mere criterion of regulatory compliance, but also beyond, taking into account the resilience of the system, the ability for the operator to protect its image by fulfilling its general duty to society and sustainable development. For that purpose the paper will take us through the basics up to the ''post graduate level'' of what transport and associated services should look like, from a theoretical point of view and also from illustrations from the current European field will be presented, illustrating the evolution of practice and interfacing of the actors.

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

    NASA Astrophysics Data System (ADS)

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

    2017-07-01

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

  3. Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR using a U7Mo Dispersion Fuel

    SciTech Connect

    Hanson A. L.; Diamond D.

    2014-06-30

    A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The LEU fuel may be a monolithic foil (LEUm) of U10Mo (10% molybdenum by weight in an alloy with uranium) or a dispersion of U7Mo in aluminum (LEUd). A previous report provided neutronic calculations for the LEUm fuel and this report presents the neutronics parameters for the LEUd fuel. The neutronics parameters for the LEUd fuel are compared to those previously obtained for the present HEU fuel and the proposed LEUm fuel. The results show no significant differences between the LEUm and the LEUd other than the LEUd fuel requires slightly less uranium than the LEUm fuel due to less molybdenum being present. The calculations include kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions under normal operation and with misloaded fuel elements.

  4. Inhibition of inflammation mediates the protective effect of atorvastatin reload in patients with coronary artery disease undergoing noncardiac emergency surgery.

    PubMed

    Qu, Yang; Wei, Lixin; Zhang, Haiqing

    2014-12-01

    This study aimed to (a) investigate whether atorvastatin reload protects against acute heart failure (AHF) in patients with stable coronary artery disease (CAD) undergoing noncardiac emergency surgery and decreases the incidence of major adverse cardiac events (MACE) during hospitalization and (b) elucidate its possible mechanism of action. In total, 500 patients with stable CAD before noncardiac emergency surgery were randomized either to the atorvastatin reload or to the placebo group. All patients received atorvastatin treatment postoperatively. The primary end point was the incidence of AHF during hospitalization, and the secondary end point was the incidence of MACE during hospitalization. Preoperative and 72 h postoperative changes in high-sensitivity C-reactive protein and interleukin-6 levels were compared between the two groups. AHF during hospitalization occurred in 5.2% of patients in the atorvastatin reload group and 11.2% in the placebo group (P=0.0225). MACE during hospitalization occurred in 2.4% of patients in the atorvastatin reload group and 8.0% in the placebo group (P=0.0088). According to multivariable analysis, atorvastatin reload conferred a 50% reduction in the risk of AHF during hospitalization (odds ratio, 0.50; 95% confidence interval, 0.2-0.8; P=0.005). The median decrease in the high-sensitivity C-reactive protein and interleukin-6 levels was significantly greater in the atorvastatin reload group (P<0.001). Atorvastatin reload may improve the clinical outcome of patients with stable CAD undergoing noncardiac emergency surgery by decreasing the incidence of AHF and MACE during hospitalization. The mechanism of this protective effect may involve inhibition of inflammation.

  5. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    SciTech Connect

    John D. Bess; Leland M. Montierth

    2013-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

  6. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    SciTech Connect

    John D. Bess; Leland M. Montierth

    2014-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

  7. Solid oxide fuel cell matrix and modules

    DOEpatents

    Riley, B.

    1988-04-22

    Porous refractory ceramic blocks arranged in an abutting, stacked configuration and forming a three dimensional array provide a support structure and coupling means for a plurality of solid oxide fuel cells (SOFCs). The stack of ceramic blocks is self-supporting, with a plurality of such stacked arrays forming a matrix enclosed in an insulating refractory brick structure having an outer steel layer. The necessary connections for air, fuel, burnt gas, and anode and cathode connections are provided through the brick and steel outer shell. The ceramic blocks are so designed with respect to the strings of modules that by simple and logical design the strings could be replaced by hot reloading if one should fail. The hot reloading concept has not been included in any previous designs. 11 figs.

  8. Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region

    NASA Technical Reports Server (NTRS)

    Klann, P. G.; Lantz, E.

    1973-01-01

    A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

  9. Modifying scoping codes to accurately calculate TMI-cores with lifetimes greater than 500 effective full-power days

    SciTech Connect

    Bai, D.; Levine, S.L. ); Luoma, J.; Mahgerefteh, M. )

    1992-01-01

    The Three Mile Island unit 1 core reloads have been designed using fast but accurate scoping codes, PSUI-LEOPARD and ADMARC. PSUI-LEOPARD has been normalized to EPRI-CPM2 results and used to calculate the two-group constants, whereas ADMARC is a modern two-dimensional, two-group diffusion theory nodal code. Problems in accuracy were encountered for cycles 8 and higher as the core lifetime was increased beyond 500 effective full-power days. This is because the heavier loaded cores in both {sup 235}U and {sup 10}B have harder neutron spectra, which produces a change in the transport effect in the baffle reflector region, and the burnable poison (BP) simulations were not accurate enough for the cores containing the increased amount of {sup 10}B required in the BP rods. In the authors study, a technique has been developed to take into account the change in the transport effect in the baffle region by modifying the fast neutron diffusion coefficient as a function of cycle length and core exposure or burnup. A more accurate BP simulation method is also developed, using integral transport theory and CPM2 data, to calculate the BP contribution to the equivalent fuel assembly (supercell) two-group constants. The net result is that the accuracy of the scoping codes is as good as that produced by CASMO/SIMULATE or CPM2/SIMULATE when comparing with measured data.

  10. Copper-palladium core-shell as an anode in a multi-fuel membraneless nanofluidic fuel cell: toward a new era of small energy conversion devices.

    PubMed

    Maya-Cornejo, J; Ortiz-Ortega, E; Álvarez-Contreras, L; Arjona, N; Guerra-Balcázar, M; Ledesma-García, J; Arriaga, L G

    2015-02-14

    A membraneless nanofluidic fuel cell with flow-through electrodes that works with several fuels (individually or mixed): methanol, ethanol, glycerol and ethylene-glycol in alkaline media is presented. For this application, an efficient Cu@Pd electrocatalyst was synthesized and tested, resulting outstanding performance until now reported, opening the possibility of power nano-devices for multi-uses purposes, regardless of fuel re-charge employed.

  11. Effect of Pd loading in Pd-Pt bimetallic catalysts doped into hollow core mesoporous shell carbon on performance of proton exchange membrane fuel cells

    NASA Astrophysics Data System (ADS)

    Fıçıcılar, Berker; Bayrakçeken, Ayşe; Eroğlu, İnci

    A significantly active Pd-Pt/carbon electrocatalyst for polymer electrolyte membrane fuel cells was synthesized by microwave irradiation using a hollow core mesoporous shell (HCMS) carbon as catalyst support that was prepared by template replication of core/shell spherical silica particles and two different carbon precursors. Pt/Pd percent weight ratios on carbon support were varied as 20/0, 15/5, 10/10, 5/15 to 0/20. As the average pore diameter of the carbon support was increased from 3.02 nm to 3.90 nm by changing the type of the carbon precursor, fuel cell performances of the HCMS carbon based Pd-Pt bimetallic catalysts were improved significantly.

  12. Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3

    SciTech Connect

    Douglas W. Akers; Edwin A. Harvego

    2012-08-01

    This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1-3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and data on previously molten fuel characteristics from TMI- 2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the RPV will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues, such as the detector/collimator design, are included in the paper.

  13. Comet whole-core solution to a stylized 3-dimensional pressurized water reactor benchmark problem with UO{sub 2}and MOX fuel

    SciTech Connect

    Zhang, D.; Rahnema, F.

    2012-07-01

    A stylized pressurized water reactor (PWR) benchmark problem with UO{sub 2} and MOX fuel was used to test the accuracy and efficiency of the coarse mesh radiation transport (COMET) code. The benchmark problem contains 125 fuel assemblies and 44,000 fuel pins. The COMET code was used to compute the core eigenvalue and assembly and pin power distributions for three core configurations. In these calculations, a set of tensor products of orthogonal polynomials were used to expand the neutron angular phase space distribution on the interfaces between coarse meshes. The COMET calculations were compared with the Monte Carlo code MCNP reference solutions using a recently published an 8-group material cross section library. The comparison showed both the core eigenvalues and assembly and pin power distributions predicated by COMET agree very well with the MCNP reference solution if the orders of the angular flux expansion in the two spatial variables and the polar and azimuth angles on the mesh boundaries are 4, 4, 2 and 2. The mean and maximum differences in the pin fission density distribution ranged from 0.28%-0.44% and 3.0%-5.5%, all within 3-sigma uncertainty of the MCNP solution. These comparisons indicate that COMET can achieve accuracy comparable to Monte Carlo. It was also found that COMET's computational speed is 450 times faster than MCNP. (authors)

  14. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    SciTech Connect

    Petrovic, Bojan; Maldonado, Ivan

    2016-04-14

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  15. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    NASA Astrophysics Data System (ADS)

    Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

    2013-10-01

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  16. Exploration of the search space of the in-core fuel management problem by knowledge-based techniques

    SciTech Connect

    Galperin, A. . Dept. of Nuclear Engineering)

    1995-02-01

    The process of generating reload configuration patterns is presented as a search procedure. The search space of the problem is found to contain [approximately] 10[sup 12] possible problem states. If computational resources and execution time necessary to evaluate a single solution are taken into account, this problem may be described as a large space search problem.'' Understanding of the structure of the search space, i.e., distribution of the optimal (or nearly optimal) solutions, is necessary to choose an appropriate search method and to utilize adequately domain heuristic knowledge. A worth function is developed based on two performance parameters: cycle length and power peaking factor. A series of numerical experiments was carried out; 300,000 patterns were generated in 40 sessions. All these patterns were analyzed by simulating the power production cycle and by evaluating the two performance parameters. The worth function was calculated and plotted. Analysis of the worth function reveals quite a complicated search space structure. The fine structure shows an extremely large number of local peaks: about one peak per hundred configurations. The direct implication of this discovery is that within a search space of 10[sup 12] states, there are [approximately]10[sup 10] local optima. Further consideration of the worth function shape shows that the distribution of the local optima forms a contour with much slower variations, where better'' or worse'' groups of patterns are spaced within a few thousand or tens of thousands of configurations, and finally very broad subregions of the whole space display variations of the worth function, where optimal regions include tens of thousands of patterns and are separated by hundreds of thousands and millions.

  17. Detrimental effects of reloading recovery on force, shortening velocity, and power of soleus muscles from hindlimb-unloaded rats.

    PubMed

    Widrick, J J; Maddalozzo, G F; Hu, H; Herron, J C; Iwaniec, U T; Turner, R T

    2008-11-01

    To better understand how atrophied muscles recover from prolonged nonweight-bearing, we studied soleus muscles (in vitro at optimal length) from female rats subjected to normal weight bearing (WB), 15 days of hindlimb unloading (HU), or 15 days HU followed by 9 days of weight bearing reloading (HU-R). HU reduced peak tetanic force (P(o)), increased maximal shortening velocity (V(max)), and lowered peak power/muscle volume. Nine days of reloading failed to improve P(o), while depressing V(max) and intrinsic power below WB levels. These functional changes appeared intracellular in origin as HU-induced reductions in soleus mass, fiber cross-sectional area, and physiological cross-sectional area were partially or completely restored by reloading. We calculated that HU-induced reductions in soleus fiber length were of sufficient magnitude to overextend sarcomeres onto the descending limb of their length-tension relationship upon the resumption of WB activity. In conclusion, the force, shortening velocity, and power deficits observed after 9 days of reloading are consistent with contraction-induced damage to the soleus. HU-induced reductions in fiber length indicate that sarcomere hyperextension upon the resumption of weight-bearing activity may be an important mechanism underlying this response.

  18. Proposed Design and Operation of a Heat Pipe Reactor using the Sandia National Laboratories Annular Core Test Facility and Existing UZrH Fuel Pins

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Lipinski, Ronald J.; Pandya, Tara; Peters, Curtis

    2005-02-01

    Heat Pipe Reactors (HPR) for space power conversion systems offer a number of advantages not easily provided by other systems. They require no pumping, their design easily deals with freezing and thawing of the liquid metal, and they can provide substantial levels of redundancy. Nevertheless, no reactor has ever been operated and cooled with heat pipes, and the startup and other operational characteristics of these systems remain largely unknown. Signification deviations from normal reactor heat removal mechanisms exist, because the heat pipes have fundamental heat removal limits due to sonic flow issues at low temperatures. This paper proposes an early prototypic test of a Heat Pipe Reactor (using existing 20% enriched nuclear fuel pins) to determine the operational characteristics of the HPR. The proposed design is similar in design to the HOMER and SAFE-300 HPR designs (Elliot, Lipinski, and Poston, 2003; Houts, et. al, 2003). However, this reactor uses existing UZrH fuel pins that are coupled to potassium heat pipes modules. The prototype reactor would be located in the Sandia Annular Core Research Reactor Facility where the fuel pins currently reside. The proposed reactor would use the heat pipes to transport the heat from the UZrH fuel pins to a water pool above the core, and the heat transport to the water pool would be controlled by adjusting the pressure and gas type within a small annulus around each heat pipe. The reactor would operate as a self-critical assembly at power levels up to 200 kWth. Because the nuclear heated HPR test uses existing fuel and because it would be performed in an existing facility with the appropriate safety authorization basis, the test could be performed rapidly and inexpensively. This approach makes it possible to validate the operation of a HPR and also measure the feedback mechanisms for a typical HPR design. A test of this nature would be the world's first operating Heat Pipe Reactor. This reactor is therefore called "HPR-1".

  19. Rolling Reloaded

    ERIC Educational Resources Information Center

    Jones, Simon A.; Nieminen, John M.

    2008-01-01

    Not so long ago a new observation about rolling motion was described: for a rolling wheel, there is a set of points with instantaneous velocities directed at or away from the centre of the wheel; these points form a circle whose diameter connects the centre of the wheel to the wheel's point of contact with the ground (Sharma 1996 "Eur. J. Phys."…

  20. Jemboss reloaded.

    PubMed

    Mullan, Lisa

    2004-06-01

    Bioinformatics tools are freely available from websites all over the world. Often they are presented as web services, although there are many tools for download and use on a local machine. This tutorial section looks at Jemboss, a Java-based graphical user interface (GUI) for the EMBOSS bioinformatics suite, which combines the advantages of both web service and downloaded software.

  1. KPZ Reloaded

    NASA Astrophysics Data System (ADS)

    Gubinelli, Massimiliano; Perkowski, Nicolas

    2017-01-01

    We analyze the one-dimensional periodic Kardar-Parisi-Zhang equation in the language of paracontrolled distributions, giving an alternative viewpoint on the seminal results of Hairer. Apart from deriving a basic existence and uniqueness result for paracontrolled solutions to the KPZ equation we perform a thorough study of some related problems. We rigorously prove the links between the KPZ equation, stochastic Burgers equation, and (linear) stochastic heat equation and also the existence of solutions starting from quite irregular initial conditions. We also show that there is a natural approximation scheme for the nonlinearity in the stochastic Burgers equation. Interpreting the KPZ equation as the value function of an optimal control problem, we give a pathwise proof for the global existence of solutions and thus for the strict positivity of solutions to the stochastic heat equation. Moreover, we study Sasamoto-Spohn type discretizations of the stochastic Burgers equation and show that their limit solves the continuous Burgers equation possibly with an additional linear transport term. As an application, we give a proof of the invariance of the white noise for the stochastic Burgers equation that does not rely on the Cole-Hopf transform.

  2. Determination of neutron flux density distribution in the core with LEU fuel IRT-4M at the training reactor VR-1

    SciTech Connect

    Huml, O.

    2008-07-15

    The objective of this work was to determine the neutron flux density distribution in various places of the training reactor VR-1 Sparrow. This experiment was performed on the new core design C1, composed of the new low-enriched uranium fuel cells IRT-4M (19.7 %). This fuel replaced the old high-enriched uranium fuel IRT-3M (36 %) within the framework of the RERTR Program in September 2005. The measurement used the neutron activation analysis method with gold wires. The principle of this method consists in neutron capture in a nucleus of the material forming the activation detector. This capture can change the nucleus in a radioisotope, whose activity can be measured. The absorption cross-section values were evaluated by MCNP computer code. The gold wires were irradiated in seven different positions in the core C1. All irradiations were performed at reactor power level 1E8 (1 kW{sub therm}). The activity of segments of irradiated wires was measured by special automatic device called 'Drat' (Wire in English). (author)

  3. Short- and Long-Term Hindlimb Immobilization and Reloading: Profile of Epigenetic Events in Gastrocnemius.

    PubMed

    Chacon-Cabrera, Alba; Gea, Joaquim; Barreiro, Esther

    2017-06-01

    Skeletal muscle dysfunction and atrophy are characteristic features accompanying chronic conditions. Epigenetic events regulate muscle mass and function maintenance. We hypothesized that the pattern of epigenetic events (muscle-enriched microRNAs and histone acetylation) and acetylation of transcription factors known to signal muscle wasting may differ between early- and late-time points in skeletal muscles of mice exposed to hindlimb immobilization (I) and recovery following I. Body and muscle weights, grip strength, muscle-enriched microRNAs, histone deacetylases (HDACs), acetylation of proteins, histones, and transcription factors (TF), myogenic TF factors, and muscle phenotype were assessed in gastrocnemius of mice exposed to periods (1, 2, 3, 7, 15, and 30 days, I groups) of hindlimb immobilization, and in those exposed to reloading for different periods of time (1, 3, 7, 15, and 30 days, R groups) following 7-day immobilization. Compared to non-immobilized controls, muscle weight, limb strength, microRNAs, especially miR-486, SIRT1 levels, and slow- and fast-twitch cross-sectional areas were decreased in mice of I groups, whereas Pax7 and acetylated FoxO1 and FoxO3 levels were increased. Muscle reloading following splint removal improved muscle mass loss, strength, and fiber atrophy, by increasing microRNAs, particularly miR-486, and SIRT1 content, while decreasing acetylated FoxO1 and FoxO3 levels. In this mouse model of disuse muscle atrophy, muscle-enriched microRNAs, especially miR-486, through Pax7 regulation delayed muscle cell differentiation following unloading of gastrocnemius muscle. Acetylation of FoxO1 and 3 seemed to drive muscle mass loss and atrophy, while deacetylation of these factors through SIRT1 would enable the muscle fibers to regenerate. J. Cell. Physiol. 232: 1415-1427, 2017. © 2016 Wiley Periodicals, Inc.

  4. Logging of post-test and CCH record core samples for the Spent Fuel Test-Climax

    SciTech Connect

    Thorpe, R.; Qualheim, B.

    1985-02-01

    Over 1300 ft of core obtained from 50 boreholes are described, with emphasis on the nature and orientations of geologic discontinuities in the Climax Stock. Although most of the core was not oriented with respect to global coordinates, approximate orientations of planar features, such as joints, veins, and sheer fractures, were obtained by taking core measurements relative to a set of low-angle joints found throughout the rock mass. The method for determining the approximate core orientation is described in detail, and the associated uncertainty is discussed. A graphic fracture log for each hole is provided, along with the computer code used to generate it.

  5. Gas-chromatographic determination of aviation gasoline and JP-4 jet fuel in subsurface core samples (journal version)

    SciTech Connect

    Vandegrift, S.A.; Kampbell, D.H.

    1988-01-01

    A new gas-chromatographic procedure for quantifying levels of aviation gasoline (avgas) and JP-4 jet fuel contamination in soils is described. The fuel is extracted from a small quantity of soil or subsurface material, typically about 6 g, using 3 mL of methylene chloride. The extract is analyzed by wide-bore capillary-column gas chromatography with flame-ionization detection. Advantages of the method are a short analysis time of 20 minutes, the use of small amounts of solvent, detection limit in the low parts-per-million range, and determination of the type of fuel contamination (i.e. avgas or jet fuel) by the chromatographic pattern or fingerprint.

  6. Translational Signalling, Atrogenic and Myogenic Gene Expression during Unloading and Reloading of Skeletal Muscle in Myostatin-Deficient Mice

    PubMed Central

    Smith, Heather K.; Matthews, Kenneth G.; Oldham, Jenny M.; Jeanplong, Ferenc; Falconer, Shelley J.; Bass, James J.; Senna-Salerno, Mônica; Bracegirdle, Jeremy W.; McMahon, Christopher D.

    2014-01-01

    Skeletal muscles of myostatin null (Mstn(−/−)) mice are more susceptible to atrophy during hind limb suspension (HS) than are muscles of wild-type mice. Here we sought to elucidate the mechanism for this susceptibility and to determine if Mstn(−/−) mice can regain muscle mass after HS. Male Mstn(−/−) and wild-type mice were subjected to 0, 2 or 7 days of HS or 7 days of HS followed by 1, 3 or 7 days of reloading (n = 6 per group). Mstn(−/−) mice lost more mass from muscles expressing the fast type IIb myofibres during HS and muscle mass was recovered in both genotypes after reloading for 7 days. Concentrations of MAFbx and MuRF1 mRNA, crucial ligases regulating the ubiquitin-proteasome system, but not MUSA1, a BMP-regulated ubiquitin ligase, were increased more in muscles of Mstn(−/−) mice, compared with wild-type mice, during HS and concentrations decreased in both genotypes during reloading. Similarly, concentrations of LC3b, Gabarapl1 and Atg4b, key effectors of the autophagy-lysosomal system, were increased further in muscles of Mstn(−/−) mice, compared with wild-type mice, during HS and decreased in both genotypes during reloading. There was a greater abundance of 4E-BP1 and more bound to eIF4E in muscles of Mstn(−/−) compared with wild-type mice (P<0.001). The ratio of phosphorylated to total eIF2α increased during HS and decreased during reloading, while the opposite pattern was observed for rpS6. Concentrations of myogenic regulatory factors (MyoD, Myf5 and myogenin) mRNA were increased during HS in muscles of Mstn(−/−) mice compared with controls (P<0.001). We attribute the susceptibility of skeletal muscles of Mstn(−/−) mice to atrophy during HS to an up- and downregulation, respectively, of the mechanisms regulating atrophy of myofibres and translation of mRNA. These processes are reversed during reloading to aid a faster rate of recovery of muscle mass in Mstn(−/−) mice. PMID:24718581

  7. Pulse electrodeposition to prepare core-shell structured AuPt@Pd/C catalyst for formic acid fuel cell application

    NASA Astrophysics Data System (ADS)

    Lu, Xueyi; Luo, Fan; Song, Huiyu; Liao, Shijun; Li, Hualing

    2014-01-01

    A novel core-shell structured AuPt@Pd/C catalyst for the electrooxidation of formic acid is synthesized by a pulse electrodeposition process, and the AuPt core nanoparticles are obtained by a NaBH4 reduction method. The catalyst is characterized with X-ray powder diffraction and transmission electron microscopy, thermogravimetric analysis, cyclic voltammetry, CO stripping and X-ray photoelectron spectroscopy. The core-shell structure of the catalyst is revealed by the increase in particle size resulting from a Pd layer covering the AuPt core, and by a negative shift in the CO stripping peaks. The addition of a small amount of Pt improves the dispersion of Au and results in smaller core particles. The catalyst's activity is evaluated by cyclic voltammetry in formic acid solution. The catalyst shows excellent activity towards the anodic oxidation of formic acid, the mass activity reaches 4.4 A mg-1Pd and 0.83 A mg-1metal, which are 8.5 and 1.6 times that of commercial Pd/C. This enhanced electrocatalytic activity could be ascribed to the good dispersion of Au core particles resulting from the addition of Pt, as well as to the interaction between the Pd shell layer and the Au and Pt in the core nanoparticles.

  8. Advanced Fuels for LWRs: Fully-Ceramic Microencapsulated and Related Concepts FY 2012 Interim Report

    SciTech Connect

    R. Sonat Sen; Brian Boer; John D. Bess; Michael A. Pope; Abderrafi M. Ougouag

    2012-03-01

    less soluble boron by, for example, increasing the reactivity hold-down by burnable poisons. Then, the whole core analysis will be repeated until an acceptable design is found. Calculations of departure from nucleate boiling ratio (DNBR) will be included in the safety evaluation as well. Once a startup core is shown to be viable, subsequent reloads will be simulated by shuffling fuel and introducing fresh fuel. The PASTA code has been updated with material properties of UN fuel from literature and a model for the diffusion and release of volatile fission products from the SiC matrix material . Preliminary simulations have been performed for both normal conditions and elevated temperatures. These results indicated that the fuel performs well and that the SiC matrix has a good retention of the fission products. The path forward for fuel performance work includes improvement of metallic fission product release from the kernel. Results should be considered preliminary and further validation is required.

  9. Experimental investigations of heat transfer and temperature fields in models simulating fuel assemblies used in the core of a nuclear reactor with a liquid heavy-metal coolant

    NASA Astrophysics Data System (ADS)

    Belyaev, I. A.; Genin, L. G.; Krylov, S. G.; Novikov, A. O.; Razuvanov, N. G.; Sviridov, V. G.

    2015-09-01

    The aim of this experimental investigation is to obtain information on the temperature fields and heat transfer coefficients during flow of liquid-metal coolant in models simulating an elementary cell in the core of a liquid heavy metal cooled fast-neutron reactor. Two design versions for spacing fuel rods in the reactor core were considered. In the first version, the fuel rods were spaced apart from one another using helical wire wound on the fuel rod external surface, and in the second version spacer grids were used for the same purpose. The experiments were carried out on the mercury loop available at the Moscow Power Engineering Institute National Research University's Chair of Engineering Thermal Physics. Two experimental sections simulating an elementary cell for each of the fuel rod spacing versions were fabricated. The temperature fields were investigated using a dedicated hinged probe that allows temperature to be measured at any point of the studied channel cross section. The heat-transfer coefficients were determined using the wall temperature values obtained at the moment when the probe thermocouple tail end touched the channel wall. Such method of determining the wall temperature makes it possible to alleviate errors that are unavoidable in case of measuring the wall temperature using thermocouples placed in slots milled in the wall. In carrying out the experiments, an automated system of scientific research was applied, which allows a large body of data to be obtained within a short period of time. The experimental investigations in the first test section were carried out at Re = 8700, and in the second one, at five values of Reynolds number. Information about temperature fields was obtained by statistically processing the array of sampled probe thermocouple indications at 300 points in the experimental channel cross section. Reach material has been obtained for verifying the codes used for calculating velocity and temperature fields in channels with

  10. CF6 jet engine performance improvement program. Short core exhaust nozzle performance improvement concept. [specific fuel consumption reduction

    NASA Technical Reports Server (NTRS)

    Fasching, W. A.

    1979-01-01

    The short core exhaust nozzle was evaluated in CF6-50 engine ground tests including performance, acoustic, and endurance tests. The test results verified the performance predictions from scale model tests. The short core exhaust nozzle provides an internal cruise sfc reduction of 0.9 percent without an increase in engine noise. The nozzle hardware successfully completed 1000 flight cycles of endurance testing without any signs of distress.

  11. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    SciTech Connect

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  12. Review of primary spaceflight-induced and secondary reloading-induced changes in slow antigravity muscles of rats.

    PubMed

    Riley, D A

    1998-01-01

    We have examined the light and electron microscopic properties of hindlimb muscles of rats flown in space for 1-2 weeks on Cosmos biosatellite flights 1887 and 2044 and Space Shuttle missions Spacelab-3, Spacelab Life Sciences-1 and Spacelab Life Sciences-2. Tissues were obtained both inflight and postflight permitting definition of primary microgravity-induced changes and secondary reentry and gravity reloading-induced alterations. Spaceflight causes atrophy and expression of fast fiber characteristics in slow antigravity muscles. The stresses of reentry and reloading reveal that atrophic muscles show increased susceptibility to interstitial edema and ischemic-anoxic necrosis as well as muscle fiber tearing with disruption of contractile proteins. These results demonstrate that the effects of spaceflight on skeletal muscle are multifaceted, and major changes occur both inflight and following return to Earth's gravity.

  13. Review of primary spaceflight-induced and secondary reloading-induced changes in slow antigravity muscles of rats

    NASA Astrophysics Data System (ADS)

    Riley, D. A.

    We have examined the light and electron microscopic properties of hindlimb muscles of rats flown in space for 1-2 weeks on Cosmos biosatellite flights 1887 and 2044 and Space Shuttle missions Spacelab-3, Spacelab Life Sciences-1 and Spacelab Life Sciences-2. Tissues were obtained both inflight and postflight permitting definition of primary microgravity-induced changes and secondary reentry and gravity reloading-induced alterations. Spaceflight causes atrophy and expression of fast fiber characteristics in slow antigravity muscles. The stresses of reentry and reloading reveal that atrophic muscles show increased susceptibility to interstitial edema and ischemic-anoxic necrosis as well as muscle fiber tearing with disruption of contractile proteins. These results demonstrate that the effects of spaceflight on skeletal muscle are multifaceted, and major changes occur both inflight and following return to Earth's gravity.

  14. Wireless Participant Incentives Using Reloadable Bank Cards to Increase Clinical Trial Retention With Abused Women Drinkers: A Natural Experiment.

    PubMed

    Rodgers, Melissa; Meisel, Zachary; Wiebe, Douglas; Crits-Christoph, Paul; Rhodes, Karin V

    2016-08-07

    Retaining participants in longitudinal studies is a unique methodological challenge in many areas of investigation, and specifically for researchers aiming to identify effective interventions for women experiencing intimate partner violence (IPV). Individuals in abusive relationships are often transient and have logistical, confidentiality, and safety concerns that limit future contact. A natural experiment occurred during a large randomized clinical trial enrolling women in abusive relationships who were also heavy drinkers, which allowed for the comparison of two incentive methods to promote longitudinal retention: cash payment versus reloadable wireless bank cards. In all, 600 patients were enrolled in the overall trial, which aimed to incentivize participants using a reloadable bank card system to promote the completion of 11 weekly interactive voice response system (IVRS) phone surveys and 3-, 6-, and 12-month follow-up phone or in person interviews. The first 145 participants were paid with cash as a result of logistical delays in setting up the bank card system. At 12 weeks, participants receiving the bank card incentive completed significantly more IVRS phone surveys, odds ratio (OR) = 2.4, 95% confidence interval (CI) = [0.01, 1.69]. There were no significant differences between the two groups related to satisfaction or safety and/or privacy. The bank card system delivered lower administrative burden for tracking payments for study staff. Based on these and other results, our large medical research university is implementing reloadable bank card as the preferred method of participant incentive payments. © The Author(s) 2016.

  15. Reactor Physics Parametric and Depletion Studies in Support of TRISO Particle Fuel Specification for the Next Generation Nuclear Plant

    SciTech Connect

    James W. Sterbentz; Bren Phillips; Robert L. Sant; Gray S. Chang; Paul D. Bayless

    2003-09-01

    Reactor physics calculations were initiated to answer several major questions related to the proposed TRISO-coated particle fuel that is to be used in the prismatic Very High Temperature Reactor (VHTR) or the Next Generation Nuclear Plant (NGNP). These preliminary design evaluation calculations help ensure that the upcoming fuel irradiation tests will test appropriate size and type of fuel particles for a future NGNP reactor design. Conclusions from these calculations are expected to confirm and suggest possible modifications to the current particle fuel parameters specified in the evolving Fuel Specification. Calculated results dispel the need for a binary fuel particle system, which is proposed in the General Atomics GT-MHR concept. The GT-MHR binary system is composed of both a fissile and fertile particle with 350- and 500- micron kernel diameters, respectively. For the NGNP reactor, a single fissile particle system (single UCO kernel size) can meet the reactivity and power cycle length requirements demanded of the NGNP. At the same time, it will provide substantial programmatic cost savings by eliminating the need for dual particle fabrication process lines and dual fuel particle irradiation tests required of a binary system. Use of a larger 425-micron kernel diameter single fissile particle (proposed here), as opposed to the 350-micron GT-MHR fissile particle size, helps alleviate current compact particle packing fractions fabrication limitations (<35%), improves fuel block loading for higher n-batch reload options, and tracks the historical correlation between particle size and enrichment (10 and 14 wt% U-235 particle enrichments are proposed for the NGNP). Overall, the use of the slightly larger kernel significantly broadens the NGNP reactor core design envelope and provides increased design margin to accommodate the (as yet) unknown final NGNP reactor design. Maximum power-peaking factors are calculated for both the initial and equilibrium NGNP cores

  16. Adaptation of the proximal femur to skeletal reloading after long-duration spaceflight.

    PubMed

    Lang, Thomas F; Leblanc, Adrian D; Evans, Harlan J; Lu, Ying

    2006-08-01

    We studied the effect of re-exposure to Earth's gravity on the proximal femoral BMD and structure of astronauts 1 year after missions lasting 4-6 months. We observed that the readaptation of the proximal femur to Earth's gravity entailed an increase in bone size and an incomplete recovery of volumetric BMD. Bone loss is a well-known result of skeletal unloading in long-duration spaceflight, with the most severe losses occurring in the proximal femur. However, there is little information about the recovery of bone loss after mission completion and no information about effect of reloading on the structure of load-bearing bone. To address these questions, we carried out a study of the effect of re-exposure to Earth's gravity on the BMD and structure of the proximal femur 1 year after missions lasting 4-6 months. In 16 crew members of the International Space Station (ISS) making flights of 4.5-6 months, we used QCT imaging to measure the total, trabecular, and cortical volumetric BMD (vBMD) of the proximal femur. In addition to vBMD, we also quantified BMC, bone volume, femoral neck cross-sectional area (CSA), and femoral neck indices of compressive and bending strength at three time-points: preflight, postflight, and 1 year after mission. Proximal femoral bone mass was substantially recovered in the year after spaceflight, but measures of vBMD and estimated bone strength showed only partial recovery. The recovery of BMC, in the absence of a comparable increase in vBMD, was explained by increases in bone volume and CSA during the year after spaceflight. Adaptation of the proximal femur to reloading entailed an increase in bone size and an incomplete recovery of vBMD. The data indicate that recovery of skeletal density after long-duration space missions may exceed 1 year and supports the evidence in the aging literature for periosteal apposition as a compensatory response for bone loss. The extent to which this compensatory effect protects against fracture remains to be

  17. Fuel transfer system

    DOEpatents

    Townsend, H.E.; Barbanti, G.

    1994-03-01

    A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool. 6 figures.

  18. Fuel transfer system

    DOEpatents

    Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool.

  19. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    NASA Technical Reports Server (NTRS)

    Walton, James T.

    1992-01-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code.

  20. Glucose uptake in rat soleus - Effect of acute unloading and subsequent reloading

    NASA Technical Reports Server (NTRS)

    Henriksen, Eric J.; Tischler, Marc E.

    1988-01-01

    The effect of acutely reduced weight bearing (unloading) on the in vitro uptake of 2-1,2-H-3-deoxy-D-glucose was studied in the soleus muscle by tail casting and suspending rats. After just 4 h, the uptake of 2-deoxy-D-glucose fell (-19 percent) and declined further after an additional 20 h of unloading. This diminution at 24 h was associated with slower oxidation of C-14-glucose and incorporation of C-14-glucose into glycogen. At 3 days of unloading, basal uptake of 2-deoxy-D-glucose did not differ from control. Reloading of the soleus after 1 or 3 days of unloading increased uptake of 2-deoxy-D-glucose above control and returned it to normal within 6 h and 4 days, respectively. These effects of unloading and recovery were caused by local changes in the soleus, because the extensor digitorum longus from the same hindlimbs did not display any alterations in uptake of 2-deoxy-D-glucose or metabolism of glucose.

  1. Analysis of an open cycle gas core nuclear propulsion system using MHD driven vortices for fuel containment

    NASA Astrophysics Data System (ADS)

    Sedwick, Raymond John

    1998-12-01

    A novel method for containing gaseous uranium vapor in an open cycle nuclear space propulsion system is developed. In an attempt to increase the operating temperature of the nuclear reactor beyond the melting point of solid fuel rods (thus increasing specific impulse), the fuel is instead suspended as a vapor in the propellant using the pressure forces developed in a confined vortex flow. The introduction of the fuel as uranium hexafluoride is found to be effective in maintaining its vapor phase in the feed passages from the tank, but not in the main vortex. A mechanism by which the resulting condensation of the uranium may be tolerated is identified, and the electro- optical properties of the resulting mixture are investigated. Containment is modeled using a 1D- axisymmetric geometry, and radiative heat transfer is found to restrict the maximum specific impulse of the system to 1500 seconds using pumping pressures of 500 atm. The specific impulse is related to this pressure as pm1/4, allowing only marginal increases in Isp at increased pressure levels. Additional 2D- axisymmetric issues, such as non-uniform current distribution and bypass flows through the boundary layers, are investigated, with possible methods of solution cited. A two-group, two-region reactor analysis is performed, estimating the mass of the reactor to be about 10 metric tonnes, and establishing the thrust to weight ratio achievable by the system at about 50. To reduce the mass of the power system, a scheme for using cross-flow heat exchange with the propellant flow to minimize (and possibly eliminate) the need for radiators to reject waste heat is presented. (Copies available exclusively from MIT Libraries, Rm. 14-0551, Cambridge, MA 02139-4307. Ph. 617-253-5668; Fax 617-253-1690.)

  2. High-capacity carbon-coated titanium dioxide core-shell nanoparticles modified three dimensional anodes for improved energy output in microbial fuel cells

    NASA Astrophysics Data System (ADS)

    Tang, Jiahuan; Yuan, Yong; Liu, Ting; Zhou, Shungui

    2015-01-01

    Three-dimensional (3D) electrodes have been intensively investigated as alternatives to conventional plate electrodes in the development of high-performance microbial fuel cells (MFCs). However, the energy output of the MFCs with the 3D anodes is still limited for practical applications. In this study, a 3D anode modified with a nano-structured capacitive layer is prepared to improve the performance of an microbial fuel cell (MFC). The capacitive layer composes of titanium dioxide (TiO2) and egg white protein (EWP)-derived carbon assembled core-shell nanoparticles, which are integrated into loofah sponge carbon (LSC) to obtain a high-capacitive 3D electrode. The as-prepared 3D anode produces a power density of 2.59 ± 0.12 W m-2, which is 63% and 201% higher than that of the original LSC and graphite anodes, respectively. The increased energy output is contributed to the enhanced electrochemical capacitance of the 3D anodes as well as the synergetic effects between TiO2 and EWP-derived carbon due to their unique properties, such as relatively high surface area, good biocompatibility, and favorable surface functionalization for interfacial microbial electron transfer. The results obtained in this study will benefit the optimized design of new 3D materials to achieve enhanced performance in MFCs.

  3. Properties of reactor fuel rod materials at high temperatures: Final summary report: Severe Core Damage Property Tests Program

    SciTech Connect

    Prater, J.T.; Courtright, E.L.

    1987-07-01

    This report summarizes work sponsored by the US Nuclear Regulatory Commission Division of Accident Evaluation to investigate those physical properties that are needed to predict the behavior of fuel-rod assemblies during a loss-of-coolant accident. The results include a determination of the oxidation kinetics of Zircaloy and Zircaloy-uranium oxide mixtures in steam and steam-hydrogen gas mixtures at 1300 to 2400C, viscosity measurements of zirconium-oxide mixtures at 1800 to 2100C, an estimate of the heat of reaction for the dissolution of uranium oxide by molten zirconium at 2000C, and thermal diffusivity measurements on prereacted Zircaloy-uranium oxide mixtures at 800 to 1500C.

  4. ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

    NASA Technical Reports Server (NTRS)

    Walton, J. T.

    1994-01-01

    ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

  5. ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

    NASA Technical Reports Server (NTRS)

    Walton, J. T.

    1994-01-01

    ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

  6. Solid oxide fuel cell matrix and modules

    DOEpatents

    Riley, Brian

    1990-01-01

    Porous refractory ceramic blocks arranged in an abutting, stacked configuration and forming a three dimensional array provide a support structure and coupling means for a plurality of solid oxide fuel cells (SOFCs). Each of the blocks includes a square center channel which forms a vertical shaft when the blocks are arranged in a stacked array. Positioned within the channel is a SOFC unit cell such that a plurality of such SOFC units disposed within a vertical shaft form a string of SOFC units coupled in series. A first pair of facing inner walls of each of the blocks each include an interconnecting channel hole cut horizontally and vertically into the block walls to form gas exit channels. A second pair of facing lateral walls of each block further include a pair of inner half circular grooves which form sleeves to accommodate anode fuel and cathode air tubes. The stack of ceramic blocks is self-supporting, with a plurality of such stacked arrays forming a matrix enclosed in an insulating refractory brick structure having an outer steel layer. The necessary connections for air, fuel, burnt gas, and anode and cathode connections are provided through the brick and steel outer shell. The ceramic blocks are so designed with respect to the strings of modules that by simple and logical design the strings could be replaced by hot reloading if one should fail. The hot reloading concept has not been included in any previous designs.

  7. Nucleoprotein supplementation enhances the recovery of rat soleus mass with reloading after hindlimb unloading-induced atrophy via myonuclei accretion and increased protein synthesis.

    PubMed

    Nakanishi, Ryosuke; Hirayama, Yusuke; Tanaka, Minoru; Maeshige, Noriaki; Kondo, Hiroyo; Ishihara, Akihiko; Roy, Roland R; Fujino, Hidemi

    2016-12-01

    Hindlimb unloading results in muscle atrophy and a period of reloading has been shown to partially recover the lost muscle mass. Two of the mechanisms involved in this recovery of muscle mass are the activation of protein synthesis pathways and an increase in myonuclei number. The additional myonuclei are provided by satellite cells that are activated by the mechanical stress associated with the reloading of the muscles and eventually incorporated into the muscle fibers. Amino acid supplementation with exercise also can increase skeletal muscle mass through enhancement of protein synthesis and nucleotide supplements can promote cell cycle activity. Therefore, we hypothesized that nucleoprotein supplementation, a combination of amino acids and nucleotides, would enhance the recovery of muscle mass to a greater extent than reloading alone after a period of unloading. Adult rats were assigned to 4 groups: control, hindlimb unloaded (HU; 14 days), reloaded (5 days) after hindlimb unloading (HUR), and reloaded after hindlimb unloading with nucleoprotein supplementation (HUR + NP). Compared with the HUR group, the HUR + NP group had larger soleus muscles and fiber cross-sectional areas, higher levels of phosphorylated rpS6, and higher numbers of myonuclei and myogenin-positive cells. These results suggest that nucleoprotein supplementation has a synergistic effect with reloading in recovering skeletal muscle properties after a period of unloading via rpS6 activation and satellite cell differentiation and incorporation into the muscle fibers. Therefore, this supplement may be an effective therapeutic regimen to include in rehabilitative strategies for a variety of muscle wasting conditions such as aging, cancer cachexia, muscular dystrophy, bed rest, and cast immobilization. Copyright © 2016 Elsevier Inc. All rights reserved.

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

    SciTech Connect

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

    2007-10-01

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

  9. Effect of boric acid mass transfer on the accumulation thereof in a fuel core under emergency modes at NPPs with WMR

    NASA Astrophysics Data System (ADS)

    Morozov, A. V.; Sorokin, A. P.; Ragulin, S. V.; Pityk, A. V.; Sahipgareev, A. R.; Soshkina, A. S.; Shlepkin, A. S.

    2017-07-01

    Boric acid mass transfer processes in the reactor facilities with WMR are considered for the case of an emergency with breaking of the main circulation pipeline (MCP) and the operation of the passive safety systems, such as first-, second-, and third-stage accumulator tank systems, and a passive heat removal system (PHRS). Calculation results are presented for a change in the boric acid concentration in the fuel core (FC) of a water-moderated reactor (WMR) in the case of an emergency process. The calculations have been performed for different values of drop entrainment of boric acid from the reactor (0, 0.2, 2%). A substantial excess of the maximum concentration of boric acid has been found to occur 24 hours after an emergency event with a break of MCP. It is shown that increasing the droplet entrainment of boric acid causes the crystallization and accumulation thereof in the FC to become slower. The mass of boric acid deposits on the elements of internals is determined depending on the values of drop entrainment. These results allow one to draw a conclusion concerning the possibility of accumulation and crystallization of boric acid in the FC, because the latter event could lead to a blocking of the flow cross section and disturbance in the heat removal from fuel elements. A review of available literature data concerning the thermal properties of boric acid solution (density, viscosity, thermal conductivity) is presented. It is found that the available data are of quite a general character, but it does not cover the entire range of parameters (temperature, pressure, acid concentrations) inherent in a possible emergency situation at nuclear power plants with WMR. It is demonstrated that experimental study of boric acid drop entrainment at the parameters inherent in the emergency mode of WMR operation, as well as the studies of boric acid thermal properties in a wide range of concentrations, are required.

  10. Postseismic Reloading: A Mechanism for Temporal Clustering of Major Earthquakes on Individual Faults

    NASA Astrophysics Data System (ADS)

    Kenner, S. J.; Simons, M.

    2001-12-01

    On a single fault segment, geologic and paleoseismic evidence from locations such as the Basin and Range [Friedrich et al. JGR, submitted] and Dead Sea Transform [Marco et al., JGR, 1996] indicate that occurrence of major earthquakes in time is often extremely heterogeneous and may, in fact, exhibit temporal clustering. We consider major earthquake clustering as the occurrence of multiple event sequences with intra-cluster inter-event times much shorter than the average time between clusters. Many factors may contribute to temporal clustering of major earthquakes. Over multiple event time scales, time-dependent postseismic stress transfer may play an important role. After major earthquakes, time-varying deformation transients occur. These transients result from diffusion of stress away from zones of stress concentration generated during the coseismic rupture. As a consequence, the coseismic fault is reloaded at a rate that is initially much higher than the background rate derived from far-field plate motions. On a given fault, earthquake recurrence intervals are moderated by various sources of system noise, including stress perturbations due to neighboring earthquakes, crustal heterogeneity, and fault evolution. Depending on the relative timing and magnitude of earthquakes in a sequence, therefore, the postseismic stress available for transfer to the coseismic fault may be greater or less than average. This may lead to a situation in which postseismic stress transfer becomes a significant factor in controlling the time to the next event. To investigate these longer-term postseismic processes, we develop a spring-dashpost-slider model of time-dependent stress transfer in the earth. With this tool, we gain an understanding of how variations in rheology, fault slip-rate, and system noise affect a fault's behavior. In tectonic environments with a weak lower crust/upper mantle, we find that small random variations in the fault failure criteria generate temporally

  11. Structure and Functional Characteristics of Rat's Left Ventricle Cardiomyocytes under Antiorthostatic Suspension of Various Duration and Subsequent Reloading

    PubMed Central

    Ogneva, I. V.; Mirzoev, T. M.; Biryukov, N. S.; Veselova, O. M.; Larina, I. M.

    2012-01-01

    The goal of the research was to identify the structural and functional characteristics of the rat's left ventricle under antiorthostatic suspension within 1, 3, 7 and 14 days, and subsequent 3 and 7-day reloading after a 14-day suspension. The transversal stiffness of the cardiomyocyte has been determined by the atomic force microscopy, cell respiration—by polarography and proteins content—by Western blotting. Stiffness of the cortical cytoskeleton increases as soon as one day after the suspension and increases up to the 14th day, and starts decreasing during reloading, reaching the control level after 7 days. The stiffness of the contractile apparatus and the intensity of cell respiration also increases. The content of non-muscle isoforms of actin in the cytoplasmic fraction of proteins does not change during the whole experiment, as does not the beta-actin content in the membrane fraction. The content of gamma-actin in the membrane fraction correlates with the change in the transversal stiffness of the cortical cytoskeleton. Increased content of alpha-actinin-1 and alpha-actinin-4 in the membrane fraction of proteins during the suspension is consistent with increased gamma-actin content there. The opposite direction of change of alpha-actinin-1 and alpha-actinin-4 content suggests their involvement into the signal pathways. PMID:23093854

  12. Probability and consequences of misloading fuel in a PWR

    SciTech Connect

    Diamond, D.J.; Hsu, C.J.; Mubayi, V. )

    1991-08-01

    This report documents the results of a study into the frequency and consequences of misloading fresh fuel assemblies during the reloading of a pressurized water reactor. The consequences that were considered included (1) loss of required shutdown margin, (2) inadvertent criticality, and (3) worker exposure within the plant given inadvertent criticality. Neutronic calculations were performed for different patterns of fresh fuel clustered together in a Combustion Engineering rector. The fresh fuel considered had a high U-235 content and was assumed to be loaded without control element assemblies. The frequencies of misloading fresh fuel assemblies into these clustered patterns were calculated taking into account operator errors an equipment malfunctions that could occur during an offload/reload sequence. The study has improved our understanding of how difficult it is to misload fuel and has quantified the loss of shutdown margin and the frequency of occurrence for specific misloadings as well as the doses that might result from an inadvertent criticality. 15 refs., 18 figs., 13 tabs.

  13. FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Abbott, W.E.; Balent, R.

    1958-09-16

    A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

  14. In situ fabrication of high-performance Ni-GDC-nanocube core-shell anode for low-temperature solid-oxide fuel cells

    NASA Astrophysics Data System (ADS)

    Yamamoto, Kazuhiro; Qiu, Nan; Ohara, Satoshi

    2015-11-01

    A core-shell anode consisting of nickel-gadolinium-doped-ceria (Ni-GDC) nanocubes was directly fabricated by a chemical process in a solution containing a nickel source and GDC nanocubes covered with highly reactive {001} facets. The cermet anode effectively generated a Ni metal framework even at 500 °C with the growth of the Ni spheres. Anode fabrication at such a low temperature without any sintering could insert a finely nanostructured layer close to the interface between the electrolyte and the anode. The maximum power density of the attractive anode was 97 mW cm-2, which is higher than that of a conventional NiO-GDC anode prepared by an aerosol process at 55 mW cm-2 and 600 °C, followed by sintering at 1300 °C. Furthermore, the macro- and microstructure of the Ni-GDC-nanocube anode were preserved before and after the power-generation test at 700 °C. Especially, the reactive {001} facets were stabled even after generation test, which served to reduce the activation energy for fuel oxidation successfully.

  15. In situ fabrication of high-performance Ni-GDC-nanocube core-shell anode for low-temperature solid-oxide fuel cells.

    PubMed

    Yamamoto, Kazuhiro; Qiu, Nan; Ohara, Satoshi

    2015-11-30

    A core-shell anode consisting of nickel-gadolinium-doped-ceria (Ni-GDC) nanocubes was directly fabricated by a chemical process in a solution containing a nickel source and GDC nanocubes covered with highly reactive {001} facets. The cermet anode effectively generated a Ni metal framework even at 500 °C with the growth of the Ni spheres. Anode fabrication at such a low temperature without any sintering could insert a finely nanostructured layer close to the interface between the electrolyte and the anode. The maximum power density of the attractive anode was 97 mW cm(-2), which is higher than that of a conventional NiO-GDC anode prepared by an aerosol process at 55 mW cm(-2) and 600 °C, followed by sintering at 1300 °C. Furthermore, the macro- and microstructure of the Ni-GDC-nanocube anode were preserved before and after the power-generation test at 700 °C. Especially, the reactive {001} facets were stabled even after generation test, which served to reduce the activation energy for fuel oxidation successfully.

  16. Carbon nanotubes/heteroatom-doped carbon core-sheath nanostructures as highly active, metal-free oxygen reduction electrocatalysts for alkaline fuel cells.

    PubMed

    Sa, Young Jin; Park, Chiyoung; Jeong, Hu Young; Park, Seok-Hee; Lee, Zonghoon; Kim, Kyoung Taek; Park, Gu-Gon; Joo, Sang Hoon

    2014-04-14

    A facile, scalable route to new nanocomposites that are based on carbon nanotubes/heteroatom-doped carbon (CNT/HDC) core-sheath nanostructures is reported. These nanostructures were prepared by the adsorption of heteroatom-containing ionic liquids on the walls of CNTs, followed by carbonization. The design of the CNT/HDC composite allows for combining the electrical conductivity of the CNTs with the catalytic activity of the heteroatom-containing HDC sheath layers. The CNT/HDC nanostructures are highly active electrocatalysts for the oxygen reduction reaction and displayed one of the best performances among heteroatom-doped nanocarbon catalysts in terms of half-wave potential and kinetic current density. The four-electron selectivity and the exchange current density of the CNT/HDC nanostructures are comparable with those of a Pt/C catalyst, and the CNT/HDC composites were superior to Pt/C in terms of long-term durability and poison tolerance. Furthermore, an alkaline fuel cell that employs a CNT/HDC nanostructure as the cathode catalyst shows very high current and power densities, which sheds light on the practical applicability of these new nanocomposites.

  17. Nuclear core positioning system

    DOEpatents

    Garkisch, Hans D.; Yant, Howard W.; Patterson, John F.

    1979-01-01

    A structural support system for the core of a nuclear reactor which achieves relatively restricted clearances at operating conditions and yet allows sufficient clearance between fuel assemblies at refueling temperatures. Axially displaced spacer pads having variable between pad spacing and a temperature compensated radial restraint system are utilized to maintain clearances between the fuel elements. The core support plates are constructed of metals specially chosen such that differential thermal expansion produces positive restraint at operating temperatures.

  18. MCNP LWR Core Generator

    SciTech Connect

    Fischer, Noah A.

    2012-08-14

    The reactor core input generator allows for MCNP input files to be tailored to design specifications and generated in seconds. Full reactor models can now easily be created by specifying a small set of parameters and generating an MCNP input for a full reactor core. Axial zoning of the core will allow for density variation in the fuel and moderator, with pin-by-pin fidelity, so that BWR cores can more accurately be modeled. LWR core work in progress: (1) Reflectivity option for specifying 1/4, 1/2, or full core simulation; (2) Axial zoning for moderator densities that vary with height; (3) Generating multiple types of assemblies for different fuel enrichments; and (4) Parameters for specifying BWR box walls. Fuel pin work in progress: (1) Radial and azimuthal zoning for generating further unique materials in fuel rods; (2) Options for specifying different types of fuel for MOX or multiple burn assemblies; (3) Additional options for replacing fuel rods with burnable poison rods; and (4) Control rod/blade modeling.

  19. Fuel flexible fuel injector

    DOEpatents

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

    2015-02-03

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

  20. Time-Course of Muscle Mass Loss, Damage, and Proteolysis in Gastrocnemius following Unloading and Reloading: Implications in Chronic Diseases

    PubMed Central

    Chacon-Cabrera, Alba; Lund-Palau, Helena; Gea, Joaquim; Barreiro, Esther

    2016-01-01

    Background Disuse muscle atrophy is a major comorbidity in patients with chronic diseases including cancer. We sought to explore the kinetics of molecular mechanisms shown to be involved in muscle mass loss throughout time in a mouse model of disuse muscle atrophy and recovery following immobilization. Methods Body and muscle weights, grip strength, muscle phenotype (fiber type composition and morphometry and muscle structural alterations), proteolysis, contractile proteins, systemic troponin I, and mitochondrial content were assessed in gastrocnemius of mice exposed to periods (1, 2, 3, 7, 15 and 30 days) of non-invasive hindlimb immobilization (plastic splint, I cohorts) and in those exposed to reloading for different time-points (1, 3, 7, 15, and 30 days, R cohorts) following a seven-day period of immobilization. Groups of control animals were also used. Results Compared to non-exposed controls, muscle weight, limb strength, slow- and fast-twitch cross-sectional areas, mtDNA/nDNA, and myosin content were decreased in mice of I cohorts, whereas tyrosine release, ubiquitin-proteasome activity, muscle injury and systemic troponin I levels were increased. Gastrocnemius reloading following splint removal improved muscle mass loss, strength, fiber atrophy, injury, myosin content, and mtDNA/nDNA, while reducing ubiquitin-proteasome activity and proteolysis. Conclusions A consistent program of molecular and cellular events leading to reduced gastrocnemius muscle mass and mitochondrial content and reduced strength, enhanced proteolysis, and injury, was seen in this non-invasive mouse model of disuse muscle atrophy. Unloading of the muscle following removal of the splint significantly improved the alterations seen during unloading, characterized by a specific kinetic profile of molecular events involved in muscle regeneration. These findings have implications in patients with chronic diseases including cancer in whom physical activity may be severely compromised. PMID

  1. Akt-dependent and Akt-independent pathways are involved in protein synthesis activation during reloading of disused soleus muscle.

    PubMed

    Mirzoev, Timur M; Tyganov, Sergey A; Shenkman, Boris S

    2017-03-01

    The purpose of our study was to assess the contribution of insulin growth factor-1-dependent and phosphatidic acid-dependent signaling pathways to activation of protein synthesis (PS) in rat soleus muscle during early recovery from unloading. Wistar rats were divided into: Control, 14HS [14-day hindlimb suspension (HS)], 3R+placebo (3-day reloading + saline administration), 3R+Wort (3-day reloading + wortmannin administration), 3R+But (3-day reloading + 1-butanol administration). SUnSET and Western blot analyses were used in this study. Wortmannin and 1-butanol induced a decrease in protein kinase B (phospho-Akt) and the rate of PS (P < 0.05) versus Control. In 3R+placebo and 3R+Wort, phosphorylation of glycogen synthase kinase 3 beta (phospho-GSK-3β) was increased versus Control (P < 0.05). Wortmannin administration during reloading did not alter phospho-p70S6K (70 kDa ribosomal protein S6 kinase) versus 3R+placebo. In 3R+But, there was a decline in phospho-GSK-3β versus 3R+placebo and Control. In 3R+But, there was a decrease in phopho-p70S6K (P < 0.05) versus 3R+placebo. These results suggest that PS activation during 3-day reloading following 14HS involves both Akt-dependent and Akt-independent pathways. Muscle Nerve 55: 393-399, 2017. © 2016 Wiley Periodicals, Inc.

  2. FUEL ELEMENT CONSTRUCTION

    DOEpatents

    Zumwalt, L.R.

    1961-08-01

    Fuel elements having a solid core of fissionable material encased in a cladding material are described. A conversion material is provided within the cladding to react with the fission products to form stable, relatively non- volatile compounds thereby minimizing the migration of the fission products into the coolant. The conversion material is preferably a metallic fluoride, such as lead difluoride, and may be in the form of a coating on the fuel core or interior of the cladding, or dispersed within the fuel core. (AEC)

  3. Thermal breeder fuel enrichment zoning

    DOEpatents

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  4. Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    SciTech Connect

    Ade, Brian J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Martinez, J. S.

    2015-01-01

    [Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, and it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades

  5. Epigallocatechin-3-gallate increases autophagy signaling in resting and unloaded plantaris muscles but selectively suppresses autophagy protein abundance in reloaded muscles of aged rats.

    PubMed

    Takahashi, Hideyuki; Suzuki, Yutaka; Mohamed, Junaith S; Gotoh, Takafumi; Pereira, Suzette L; Alway, Stephen E

    2017-03-07

    We have previously found that Epigallocatechin-3-gallate (EGCg), an abundant catechin in green tea, reduced apoptotic signaling and improved muscle recovery in response to reloading after hindlimb suspension (HS). In this study, we investigated if EGCg altered autophagy signaling in skeletal muscle of old rats in response to HS or reloading after HS. Fischer 344×Brown Norway inbred rats (age 34months) were given 1ml/day of purified EGCg (50mg/kg body weight), or the same sample volume of the vehicle by gavage. One group of animals received HS for 14days and the second group of rats received 14days of HS, then the HS was removed and they were allowed to recover by ambulating normally around the cage for two weeks. EGCg decreased a small number of autophagy genes in control muscles, but it increased the expression of other autophagy genes (e.g., ATG16L2, SNCA, TM9SF1, Pink1, PIM-2) and HS did not attenuate these increases. HS increased Beclin1, ATG7 and LC3-II/I protein abundance in hindlimb muscles. Relative to vehicle treatment, EGCg treatment had greater ATG12 protein abundance (35.8%, P<0.05), but decreased Beclin1 protein levels (-101.1%, P<0.05) after HS. However, in reloaded muscles, EGCg suppressed Beclin1 and LC3-II/I protein abundance as compared to vehicle treated muscles. EGCg appeared to "prime" autophagy signaling before and enhance autophagy gene expression and protein levels during unloading in muscles of aged rats, perhaps to improve the clearance of damaged organelles. However, EGCg suppressed autophagy signaling after reloading, potentially to increase the recovery of hindlimb muscles mass and function after loading is restored.

  6. REACTOR FUEL SCAVENGING MEANS

    DOEpatents

    Coffinberry, A.S.

    1962-04-10

    A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

  7. Extended burnup core management for once-through uranium fuel cycles in LWRS. First annual report for the period 1 July 1979-30 June 1980

    SciTech Connect

    Sesonske, A.

    1980-08-01

    Detailed core management arrangements are developed requiring four operating cycles for the transition from present three-batch loading to an extended burnup four-batch plan for Zion-1. The ARMP code EPRI-NODE-P was used for core modeling. Although this work is preliminary, uranium and economic savings during the transition cycles appear of the order of 6 percent.

  8. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    SciTech Connect

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  9. Risk-based decision making for staggered bioterrorist attacks : resource allocation and risk reduction in "reload" scenarios.

    SciTech Connect

    Lemaster, Michelle Nicole; Gay, David M.; Ehlen, Mark Andrew; Boggs, Paul T.; Ray, Jaideep

    2009-10-01

    Staggered bioterrorist attacks with aerosolized pathogens on population centers present a formidable challenge to resource allocation and response planning. The response and planning will commence immediately after the detection of the first attack and with no or little information of the second attack. In this report, we outline a method by which resource allocation may be performed. It involves probabilistic reconstruction of the bioterrorist attack from partial observations of the outbreak, followed by an optimization-under-uncertainty approach to perform resource allocations. We consider both single-site and time-staggered multi-site attacks (i.e., a reload scenario) under conditions when resources (personnel and equipment which are difficult to gather and transport) are insufficient. Both communicable (plague) and non-communicable diseases (anthrax) are addressed, and we also consider cases when the data, the time-series of people reporting with symptoms, are confounded with a reporting delay. We demonstrate how our approach develops allocations profiles that have the potential to reduce the probability of an extremely adverse outcome in exchange for a more certain, but less adverse outcome. We explore the effect of placing limits on daily allocations. Further, since our method is data-driven, the resource allocation progressively improves as more data becomes available.

  10. LMFBR fuel assembly design for HCDA fuel dispersal

    DOEpatents

    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.

  11. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Stacy, J.T.

    1958-12-01

    A reactor fuel element having a core of molybdenum-uranium alloy jacketed in stainless steel is described. A barrier layer of tungsten, tantalum, molybdenum, columbium, or silver is interposed between the core and jacket to prevent formation of a low melting eutectic between uranium and the varlous alloy constituents of the stainless steel.

  12. JACKETED REACTOR FUEL ELEMENT

    DOEpatents

    Smith, K.F.; Van Thyne, R.J.

    1958-12-01

    A fuel element is described for fast reactors comprised of a core of uranium metal containing material and a jacket around the core, the jacket consisting of from 2.5 to 15 percent of titanium, from 1 to 5 percent of niobium, and from 80 to 96.5 percent of vanadium.

  13. Performance and fuel cycle cost study of the R2 reactor with HEU and LEU fuels

    SciTech Connect

    Pond, R.B.; Freese, K.E.; Matos, J.E.

    1984-01-01

    A systematic study of the experiment performance and fuel cycle costs of the 50 MW R2 reactor operated by Studsvik Energiteknik AB has been performed using the current R2 HEU fuel, a variety of LEU fuel element designs, and two core-box/reflector configurations. The results include the relative performance of both in-core and ex-core experiments, control rod worths, and relative annual fuel cycle costs.

  14. Mixed oxide fuel development

    SciTech Connect

    Leggett, R.D.; Omberg, R.P.

    1987-05-08

    This paper describes the success of the ongoing mixed-oxide fuel development program in the United States aimed at qualifying an economical fuel system for liquid metal cooled reactors. This development has been the cornerstone of the US program for the past 20 years and has proceeded in a deliberate and highly disciplined fashion with high emphasis on fuel reliability and operational safety as major features of an economical fuel system. The program progresses from feature testing in EBR-II to qualifying full size components in FFTF under fully prototypic conditions to establish a basis for extending allowable lifetimes. The development program started with the one year (300 EFPD) core, which is the FFTF driver fuel, continued with the demonstration of a two year (600 EFPD) core and is presently evaluating a three year (900 EFPD) fuel system. All three of these systems, consistent with other LMR fuel programs around the world, use fuel pellets gas bonded to a cladding tube that is assembled into a bundle and fitted into a wrapper tube or duct for ease of insertion into a core. The materials of construction progressed from austenitic CW 316 SS to lower swelling austenitic D9 to non swelling ferritic/martensitic HT9. 6 figs., 2 tabs.

  15. Fueling systems

    SciTech Connect

    Gorker, G.E.

    1987-01-01

    This report deals with concepts of the Tiber II tokamak reactor fueling systems. Contained in this report are the fuel injection requirement data, startup fueling requirements, intermediate range fueling requirements, power range fueling requirements and research and development considerations. (LSR)

  16. FUEL ASSAY REACTOR

    DOEpatents

    Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

    1962-12-25

    A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

  17. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Shackleford, M.H.

    1958-12-16

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

  18. Core design and performance of small inherently safe LMRs

    SciTech Connect

    Orechwa, Y.; Khalil, H.; Turski, R.B.; Fujita, E.K.

    1986-01-01

    Oxide and metal-fueled core designs at the 900 MWt level and constrained by a requirement for interchangeability are described. The physics parameters of the two cores studied here indicate that metal-fueled cores display attractive economic and safety features and are more flexible than are oxide cores in adapting to currently-changing deployment scenarios.

  19. A new alkali-resistant Ni/Al2O3-MSU-1 core-shell catalyst for methane steam reforming in a direct internal reforming molten carbonate fuel cell

    NASA Astrophysics Data System (ADS)

    Zhang, Jian; Zhang, Xiongfu; Liu, Weifeng; Liu, Haiou; Qiu, Jieshan; Yeung, King Lun

    2014-01-01

    An alkali-resistant catalyst for direct internal reforming molten carbonate fuel cell (DIR-MCFC) is prepared by growing a thin shell of mesoporous MSU-1 membrane on Ni/Al2O3 catalyst beads. The MSU-1 shell is obtained by first depositing a monolayer of colloidal silicalite-1 (Sil-1) on the catalyst bead as linkers and then using NaF stored in the beads to catalyze the growth of the MSU-1 layer. The resulting core-shell catalysts display excellent alkali-resistance and deliver stable methane conversion and hydrogen yield in an out-of-cell test simulating the operating conditions of an operating DIR-MCFC. Higher conversion and yield (i.e., up to over 70%) are obtained from the new core-shell catalyst with MSU-1 shell compared to the catalyst with microporous Sil-1 shell. A mathematical model of the reaction and poisoning of the core-shell catalyst is used to predict the optimum shell thickness for its reliable use in a DIR-MCFC.

  20. Singlet oxygen sensitizing materials based on porous silicone: photochemical characterization, effect of dye reloading and application to water disinfection with solar reactors.

    PubMed

    Manjón, Francisco; Santana-Magaña, Montserrat; García-Fresnadillo, David; Orellana, Guillermo

    2010-06-01

    Photogeneration of singlet molecular oxygen ((1)O(2)) is applied to organic synthesis (photooxidations), atmosphere/water treatment (disinfection), antibiofouling materials and in photodynamic therapy of cancer. In this paper, (1)O(2) photosensitizing materials containing the dyes tris(4,4'-diphenyl-2,2'-bipyridine)ruthenium(II) (1, RDB(2+)) or tris(4,7-diphenyl-1,10-phenanthroline)ruthenium(II) (2, RDP(2+)), immobilized on porous silicone (abbreviated RDB/pSil and RDP/pSil), have been produced and tested for waterborne Enterococcus faecalis inactivation using a laboratory solar simulator and a compound parabolic collector (CPC)-based solar photoreactor. In order to investigate the feasibility of its reuse, the sunlight-exposed RDP/pSil sensitizing material (RDP/pSil-a) has been reloaded with RDP(2+) (RDP/pSil-r). Surprisingly, results for bacteria inactivation with the reloaded material have demonstrated a 4-fold higher efficiency compared to those of either RDP/pSil-a, unused RDB/pSil and the original RDP/pSil. Surface and bulk photochemical characterization of the new material (RDP/pSil-r) has shown that the bactericidal efficiency enhancement is due to aggregation of the silicone-supported photosensitizer on the surface of the polymer, as evidenced by confocal fluorescence lifetime imaging microscopy (FLIM). Photogenerated (1)O(2) lifetimes in the wet sensitizer-doped silicone have been determined to be ten times longer than in water. These facts, together with the water rheology in the solar reactor and the interfacial production of the biocidal species, account for the more effective disinfection observed with the reloaded photosensitizing material. These results extend and improve the operational lifetime of photocatalytic materials for point-of-use (1)O(2)-mediated solar water disinfection.

  1. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    SciTech Connect

    Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N.

    2006-02-01

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

  2. High-performance 1D type-II TiO2@ZnO core-shell nanorods arrays photoanodes for photoelectrochemical solar fuel production

    NASA Astrophysics Data System (ADS)

    Hu, Sujuan; Wang, Baoling; Zhu, Mingshan; Ma, Yinhai; Lv, Zhonghao; Wang, Hai

    2017-05-01

    In this paper, vertically aligned one-dimensional (1D) TiO2@ZnO core-shell nanorods arrays (NRs) were prepared by a facile two-step hydrothermal method. The as-prepared TiO2@ZnO nanorods arrays were used as photoanode and exhibited high photoelectrochemical water splitting performance. A maximum photocurrent density of TiO2@ZnO nanorods arrays achieves to 2.02 mA cm-2 at 1.7 V vs. RHE, which is 2 times higher than that of pure TiO2 nanorods arrays. In addition, a significant cathodic shift in the onset potential and a stable photocurrent during 6000 s measurement are observed in TiO2@ZnO nanorods arrays. TiO2@ZnO core-shell NRs electrode shows a maximal conversion efficiency of 0.48% at 0.55 V vs. RHE, which is 2.7 times higher than that of the bare TiO2 NRs electrode. The improvement of photoelectrochemical properties can be attributed to the ingenious core-shell structure that enhances the interfacial interaction based on the type II heterojunction. An efficient interfacial charge separation and photoinduced charge transfer are achieved in 1D type-II TiO2@ZnO core-shell nanorods arrays.

  3. High flux research reactors based on particulate fuel

    SciTech Connect

    Powell, J.R.; Takahashi, H.; Horn, F.L.

    1986-02-01

    High Flux Particle Bed Reactor (HFPBR) designs based on High Temperature Gas Reactors (HTGR) particular fuel are described. The coated fuel particles, approx.500 microns in diameter, are packed between porous metal frits, and directly cooled by flowing D/sub 2/O. The large heat transfer surface area in the packed bed, approx.100 cm/sup 2//cm/sup 3/ of volume, allows high power densities, typically 10 MW/liter. Peak thermal fluxes in the HFPBR are 1 to 2 x 1/sup 16/ n/c/sup 2/ sec., depending on configuration and moderator choice with beryllium and D/sub 2/O Moderators yielding the best flux performance. Spent fuel particles can be hydraulically unloaded every day or two and fresh fuel reloaded. The short fuel cycle allows HFPBR fuel loading to be very low, approx.2 kg of /sup 235/U, with a fission product inventory one-tenth of that in present high flux research reactors. The HFPBR can use partially enriched fuel, 20% /sup 235/U, without degradation in flux reactivity. 8 refs., 12 figs., 2 tabs.

  4. FUEL ELEMENTS FOR NUCLEAR REACTORS

    DOEpatents

    Blainey, A.; Lloyd, H.

    1961-07-11

    A method of sheathing a tubular fuel element for a nuclear reactor is described. A low melting metal core member is centered in a die, a layer of a powdered sheathing substance is placed on the bottom of the die, the tubular fuel element is inserted in the die, the space between the tubular fuel element and the die walls and core member is filled with the same powdered sheathing substance, a layer of the same substance is placed over the fissile material, and the charge within the die is subjected to pressure in the direction of the axis of the fuel element at the sintering temperature of the protective substance.

  5. Nondestrucive analysis of fuel pins

    DOEpatents

    Stepan, I.E.; Allard, N.P.; Suter, C.R.

    1972-11-03

    Disclosure is made of a method and a correspondingly adapted facility for the nondestructive analysis of the concentation of fuel and poison in a nuclear reactor fuel pin. The concentrations of fuel and poison in successive sections along the entire length of the fuel pin are determined by measuring the reactivity of a thermal reactor as each successive small section of the fuel pin is exposed to the neutron flux of the reactor core and comparing the measured reactivity with the reactivities measured for standard fuel pins having various known concentrations. Only a small section of the length of the fuel pin is exposed to the neutron flux at any one time while the remainder of the fuel pin is shielded from the neutron flux. In order to expose only a small section at any one time, a boron-10-lined dry traverse tube is passed through the test region within the core of a low-power thermal nuclear reactor which has a very high fuel sensitivity. A narrow window in the boron-10 lining is positioned at the core center line. The fuel pins are then systematically traversed through the tube past the narrow window such that successive small sections along the length of the fuel pin are exposed to the neutron flux which passes through the narrow window.

  6. COVERING A CORE BY EXTRUSION

    DOEpatents

    Karnie, A.J.

    1963-07-16

    A method of covering a cylindrical fuel core with a cladding metal ms described. The metal is forced between dies around the core from both ends in two opposing skirts, and as these meet the ends turn outward into an annular recess in the dics. By cutting off the raised portion formed by the recess, oxide impurities are eliminated. (AEC)

  7. Nitrogen- and boron-co-doped core-shell carbon nanoparticles as efficient metal-free catalysts for oxygen reduction reactions in microbial fuel cells

    NASA Astrophysics Data System (ADS)

    Zhong, Shengkui; Zhou, Lihua; Wu, Ling; Tang, Lianfeng; He, Qiyi; Ahmed, Jalal

    2014-12-01

    The most severe bottleneck hindering the widespread application of fuel cell technologies is the difficulty in obtaining an inexpensive and abundant oxygen reduction reaction (ORR) catalyst. The concept of a heteroatom-doped carbon-based metal-free catalyst has recently attracted interest. In this study, a metal-free carbon nanoparticles-based catalyst hybridized with dual nitrogen and boron components was synthesized to catalyze the ORR in microbial fuel cells (MFCs). Multiple physical and chemical characterizations confirmed that the synthetic method enabled the incorporation of both nitrogen and boron dopants. The electrochemical measurements indicated that the co-existence of nitrogen and boron could enhance the ORR kinetics by reducing the overpotential and increasing the current density. The results from the kinetic studies indicated that the nitrogen and boron induced an oxygen adsorption mechanism and a four-electron-dominated reaction pathway for the as-prepared catalyst that was very similar to those induced by Pt/C. The MFC results showed that a maximum power density of ∼642 mW m-2 was obtained using the as-prepared catalyst, which is comparable to that obtained using expensive Pt catalyst. The prepared nitrogen- and boron-co-doped carbon nanoparticles might be an alternative cathode catalyst for MFC applications if large-scale applications and price are considered.

  8. Disruption of adaptor protein 2μ (AP-2μ) in cochlear hair cells impairs vesicle reloading of synaptic release sites and hearing.

    PubMed

    Jung, SangYong; Maritzen, Tanja; Wichmann, Carolin; Jing, Zhizi; Neef, Andreas; Revelo, Natalia H; Al-Moyed, Hanan; Meese, Sandra; Wojcik, Sonja M; Panou, Iliana; Bulut, Haydar; Schu, Peter; Ficner, Ralf; Reisinger, Ellen; Rizzoli, Silvio O; Neef, Jakob; Strenzke, Nicola; Haucke, Volker; Moser, Tobias

    2015-11-03

    Active zones (AZs) of inner hair cells (IHCs) indefatigably release hundreds of vesicles per second, requiring each release site to reload vesicles at tens per second. Here, we report that the endocytic adaptor protein 2μ (AP-2μ) is required for release site replenishment and hearing. We show that hair cell-specific disruption of AP-2μ slows IHC exocytosis immediately after fusion of the readily releasable pool of vesicles, despite normal abundance of membrane-proximal vesicles and intact endocytic membrane retrieval. Sound-driven postsynaptic spiking was reduced in a use-dependent manner, and the altered interspike interval statistics suggested a slowed reloading of release sites. Sustained strong stimulation led to accumulation of endosome-like vacuoles, fewer clathrin-coated endocytic intermediates, and vesicle depletion of the membrane-distal synaptic ribbon in AP-2μ-deficient IHCs, indicating a further role of AP-2μ in clathrin-dependent vesicle reformation on a timescale of many seconds. Finally, we show that AP-2 sorts its IHC-cargo otoferlin. We propose that binding of AP-2 to otoferlin facilitates replenishment of release sites, for example, via speeding AZ clearance of exocytosed material, in addition to a role of AP-2 in synaptic vesicle reformation.

  9. Fossil fuels -- future fuels

    SciTech Connect

    1998-03-01

    Fossil fuels -- coal, oil, and natural gas -- built America`s historic economic strength. Today, coal supplies more than 55% of the electricity, oil more than 97% of the transportation needs, and natural gas 24% of the primary energy used in the US. Even taking into account increased use of renewable fuels and vastly improved powerplant efficiencies, 90% of national energy needs will still be met by fossil fuels in 2020. If advanced technologies that boost efficiency and environmental performance can be successfully developed and deployed, the US can continue to depend upon its rich resources of fossil fuels.

  10. PROCESS FOR JACKETING A CORE

    DOEpatents

    Last, G.A.

    1960-07-19

    A process is given for enclosing the uranium core of a nuclear fuel element by placing the core in an aluminum cup and closing the open end of the cup over the core. As the metal of the cup is brought together in a weld over the center of the end of the core, it is extruded inwardly as internal projection into a central recess in the core and outwardly as an external projection. Thus oxide inclusions in the weld of the cup are spread out into the internal and external projections and do not interfere with the integrity of the weld.

  11. Core physics analysis of 100% MOX Core in IRIS

    SciTech Connect

    Franceschini, F.; Petrovic, B.

    2006-07-01

    International Reactor Innovative and Secure (IRIS) is an advanced small-to-medium-size (1000 MWt) Pressurized Water Reactor (PWR), targeting deployment around 2015. Its reference core design is based on the current Westinghouse UO{sub 2} fuel with less than 5% {sup 235}U, and the analysis has been previously completed confirming good performance. The full MOX fuel core is currently under evaluation as one of the alternatives for the second wave of IRIS reactors. A full 3-D neutronic analysis has been performed to examine main core performance parameters, such as critical boron concentration, peaking factors, discharge burnup, etc. The enhanced moderation of the IRIS fuel lattice facilitates MOX core design, and all the obtained results are within the requirements, confirming viability of this option from the reactor physics standpoint. (authors)

  12. Composite fuel cell membranes

    DOEpatents

    Plowman, K.R.; Rehg, T.J.; Davis, L.W.; Carl, W.P.; Cisar, A.J.; Eastland, C.S.

    1997-08-05

    A bilayer or trilayer composite ion exchange membrane is described suitable for use in a fuel cell. The composite membrane has a high equivalent weight thick layer in order to provide sufficient strength and low equivalent weight surface layers for improved electrical performance in a fuel cell. In use, the composite membrane is provided with electrode surface layers. The composite membrane can be composed of a sulfonic fluoropolymer in both core and surface layers.

  13. Vibrating fuel grapple. [LMFBR

    DOEpatents

    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.

  14. Vibrating fuel grapple

    DOEpatents

    Chertock, deceased, Alan J.; Fox, Jack N.; Weissinger, Robert B.

    1982-01-01

    A reactor refueling method utilizing a vibrating fuel grapple for removing spent fuel assemblies from a reactor core which incorporates a pneumatic vibrator in the grapple head, enabling 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.

  15. Composite fuel cell membranes

    DOEpatents

    Plowman, Keith R.; Rehg, Timothy J.; Davis, Larry W.; Carl, William P.; Cisar, Alan J.; Eastland, Charles S.

    1997-01-01

    A bilayer or trilayer composite ion exchange membrane suitable for use in a fuel cell. The composite membrane has a high equivalent weight thick layer in order to provide sufficient strength and low equivalent weight surface layers for improved electrical performance in a fuel cell. In use, the composite membrane is provided with electrode surface layers. The composite membrane can be composed of a sulfonic fluoropolymer in both core and surface layers.

  16. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    SciTech Connect

    John D. Bess

    2014-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  17. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    SciTech Connect

    John D. Bess

    2013-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  18. The technique and preliminary results of LEU U-Mo full-size IRT type fuel testing in the MIR reactor

    SciTech Connect

    Izhutov, A.L.; Starkov, V.A.; Pimenov, V.V.; Fedoseev, V.Ye.; Dobrikova, I.V.; Vatulin, A.V.; Suprun, V.B.; Kartashov, Ye.F.; Lukichev, V.A.; Troyanov, V.M.; Enin, A.A.; Tkachev, A.A.

    2008-07-15

    In March 2007 in-pile testing of LEU U-Mo full-size IRT type fuel elements was started in the MIR reactor. Four prototype fuel elements for Uzbekistan WWR SM reactor are being tested simultaneously - two of tube type design and two of pin type design. The dismountable irradiation devices were constructed for intermediate reloading and inspection of fuel elements during reactor testing. The objective of the test is to obtain the experimental results for determination of more reliable design and licensing fuel elements for conversion of the WWR SM reactor. The heat power of fuel elements is measured on-line by thermal balance method. The distribution of fission density and burn-up of uranium in the volume of elements are calculated by using the MIR reactor MCU code (Monte-Carlo) model. In this paper the design of fuel elements, the technique, main parameters and preliminary results are described. (author)

  19. Co-Pt core-shell nanostructured catalyst prepared by selective chemical vapor pulse deposition of Pt on Co as a cathode in polymer electrolyte fuel cells

    SciTech Connect

    Seo, Sang-Joon; Chung, Ho-Kyoon; Yoo, Ji-Beom; Chae, Heeyeop; Seo, Seung-Woo; Min Cho, Sung

    2014-01-15

    A new type of PtCo/C catalyst for use as a cathode in polymer electrolyte fuel cells was prepared by selective chemical vapor pulse deposition (CVPD) of Pt on the surface of Co. The activity of the prepared catalyst for oxygen reduction was higher than that of a catalyst prepared by sequential impregnation (IMP) with the two metallic components. This catalytic activity difference occurs because the former catalyst has smaller Pt crystallites that produce stronger Pt-Co interactions and have a larger Pt surface area. Consequently, the CVPD catalyst has a great number of Co particles that are in close contact with the added Pt. The Pt surface was also electronically modified by interactions with Co, which were stronger in the CVPD catalyst than in the IMP catalyst, as indicated by X-ray diffraction, X-ray photoemission spectroscopy, and cyclic voltammetry measurements of the catalysts.

  20. Emergency core cooling system

    DOEpatents

    Schenewerk, William E.; Glasgow, Lyle E.

    1983-01-01

    A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.

  1. Nuclear reactor fuel element

    DOEpatents

    Johnson, Carl E.; Crouthamel, Carl E.

    1980-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

  2. Nuclear reactor composite fuel assembly

    DOEpatents

    Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  3. FUEL ELEMENTS FOR NEUTRONIC REACTORS

    DOEpatents

    Foote, F.G.; Jette, E.R.

    1963-05-01

    A fuel element for a nuclear reactor is described that consists of a jacket containing a unitary core of fissionable material and a filling of a metal of the group consisting of sodium and sodium-potassium alloys. (AEC)

  4. NASA reload program

    NASA Technical Reports Server (NTRS)

    Byington, Marshall

    1993-01-01

    Atlantic Research Corporation (ARC) contracted with NASA to manufacture and deliver thirteen small scale Solid Rocket Motors (SRM). These motors, containing five distinct propellant formulations, will be used for plume induced radiation studies. The information contained herein summarizes and documents the program accomplishments and results. Several modifications were made to the scope of work during the course of the program. The effort was on hold from late 1991 through August, 1992 while propellant formulation changes were developed. Modifications to the baseline program were completed in late-August and Modification No. 6 was received by ARC on September 14, 1992. The modifications include changes to the propellant formulation and the nozzle design. The required motor deliveries were completed in late-December, 1992. However, ARC agreed to perform an additional mix and cast effort at no cost to NASA and another motor was delivered in March, 1993.

  5. Oral Insulin Reloaded

    PubMed Central

    Heinemann, Lutz; Plum-Mörschel, Leona

    2014-01-01

    Optimal coverage of insulin needs is the paramount aim of insulin replacement therapy in patients with diabetes mellitus. To apply insulin without breaking the skin barrier by a needle and/or to allow a more physiological provision of insulin are the main reasons triggering the continuous search for alternative routes of insulin administration. Despite numerous attempts over the past 9 decades to develop an insulin pill, no insulin for oral dosing is commercially available. By way of a structured approach, we aim to provide a systematic update on the most recent developments toward an orally available insulin formulation with a clear focus on data from clinical-experimental and clinical studies. Thirteen companies that claim to be working on oral insulin formulations were identified. However, only 6 of these companies published new clinical trial results within the past 5 years. Interestingly, these clinical data reports make up a mere 4% of the considerably high total number of publications on the development of oral insulin formulations within this time period. While this picture clearly reflects the rising research interest in orally bioavailable insulin formulations, it also highlights the fact that the lion’s share of research efforts is still allocated to the preclinical stages. PMID:24876606

  6. Plutonium age dating reloaded

    NASA Astrophysics Data System (ADS)

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

    2014-05-01

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

  7. The Heliogyro Reloaded

    NASA Technical Reports Server (NTRS)

    Wilkie, William K.; Warren, Jerry E.; Thompson, M. W.; Lisman, P. D.; Walkemeyer, P. E.; Guerrant, D. V.; Lawrence, D. A.

    2011-01-01

    The heliogyro is a high-performance, spinning solar sail architecture that uses long - order of kilometers - reflective membrane strips to produce thrust from solar radiation pressure. The heliogyro s membrane blades spin about a central hub and are stiffened by centrifugal forces only, making the design exceedingly light weight. Blades are also stowed and deployed from rolls; eliminating deployment and packaging problems associated with handling extremely large, and delicate, membrane sheets used with most traditional square-rigged or spinning disk solar sail designs. The heliogyro solar sail concept was first advanced in the 1960s by MacNeal. A 15 km diameter version was later extensively studied in the 1970s by JPL for an ambitious Comet Halley rendezvous mission, but ultimately not selected due to the need for a risk-reduction flight demonstration. Demonstrating system-level feasibility of a large, spinning heliogyro solar sail on the ground is impossible; however, recent advances in microsatellite bus technologies, coupled with the successful flight demonstration of reflectance control technologies on the JAXA IKAROS solar sail, now make an affordable, small-scale heliogyro technology flight demonstration potentially feasible. In this paper, we will present an overview of the history of the heliogyro solar sail concept, with particular attention paid to the MIT 200-meter-diameter heliogyro study of 1989, followed by a description of our updated, low-cost, heliogyro flight demonstration concept. Our preliminary heliogyro concept (HELIOS) should be capable of demonstrating an order-of-magnitude characteristic acceleration performance improvement over existing solar sail demonstrators (HELIOS target: 0.5 to 1.0 mm/s2 at 1.0 AU); placing the heliogyro technology in the range required to enable a variety of science and human exploration relevant support missions.

  8. AMS-02 antiprotons reloaded

    SciTech Connect

    Kappl, Rolf; Reinert, Annika; Winkler, Martin Wolfgang E-mail: areinert@th.physik.uni-bonn.de

    2015-10-01

    The AMS-02 collaboration has released preliminary data on the antiproton fraction in cosmic rays. The surprisingly hard antiproton spectrum at high rigidity has triggered speculations about a possible primary antiproton component originating from dark matter annihilations. In this note, we employ newly available AMS-02 boron to carbon data to update the secondary antiproton flux within the standard two-zone diffusion model. The new background permits a considerably better fit to the measured antiproton fraction compared to previous estimates. This is mainly a consequence of the smaller slope of the diffusion coefficient favored by the new AMS-02 boron to carbon data.

  9. Fragmentation trees reloaded.

    PubMed

    Böcker, Sebastian; Dührkop, Kai

    2016-01-01

    Untargeted metabolomics commonly uses liquid chromatography mass spectrometry to measure abundances of metabolites; subsequent tandem mass spectrometry is used to derive information about individual compounds. One of the bottlenecks in this experimental setup is the interpretation of fragmentation spectra to accurately and efficiently identify compounds. Fragmentation trees have become a powerful tool for the interpretation of tandem mass spectrometry data of small molecules. These trees are determined from the data using combinatorial optimization, and aim at explaining the experimental data via fragmentation cascades. Fragmentation tree computation does not require spectral or structural databases. To obtain biochemically meaningful trees, one needs an elaborate optimization function (scoring). We present a new scoring for computing fragmentation trees, transforming the combinatorial optimization into a Maximum A Posteriori estimator. We demonstrate the superiority of the new scoring for two tasks: both for the de novo identification of molecular formulas of unknown compounds, and for searching a database for structurally similar compounds, our method SIRIUS 3, performs significantly better than the previous version of our method, as well as other methods for this task. SIRIUS 3 can be a part of an untargeted metabolomics workflow, allowing researchers to investigate unknowns using automated computational methods.Graphical abstractWe present a new scoring for computing fragmentation trees from tandem mass spectrometry data based on Bayesian statistics. The best scoring fragmentation tree most likely explains the molecular formula of the measured parent ion.

  10. Corrosion resistant PEM fuel cell

    DOEpatents

    Li, Yang; Meng, Wen-Jin; Swathirajan, Swathy; Harris, Stephen J.; Doll, Gary L.

    1997-01-01

    The present invention contemplates a PEM fuel cell having electrical contact elements (including bipolar plates/septums) comprising a titanium nitride coated light weight metal (e.g., Al or Ti) core, having a passivating, protective metal layer intermediate the core and the titanium nitride. The protective layer forms a barrier to further oxidation/corrosion when exposed to the fuel cell's operating environment. Stainless steels rich in CR, Ni, and Mo are particularly effective protective interlayers.

  11. Corrosion resistant PEM fuel cell

    DOEpatents

    Li, Y.; Meng, W.J.; Swathirajan, S.; Harris, S.J.; Doll, G.L.

    1997-04-29

    The present invention contemplates a PEM fuel cell having electrical contact elements (including bipolar plates/septums) comprising a titanium nitride coated light weight metal (e.g., Al or Ti) core, having a passivating, protective metal layer intermediate the core and the titanium nitride. The protective layer forms a barrier to further oxidation/corrosion when exposed to the fuel cell`s operating environment. Stainless steels rich in Cr, Ni, and Mo are particularly effective protective interlayers. 6 figs.

  12. Corrosion resistant PEM fuel cell

    DOEpatents

    Li, Yang; Meng, Wen-Jin; Swathirajan, Swathy; Harris, Stephen Joel; Doll, Gary Lynn

    2001-07-17

    The present invention contemplates a PEM fuel cell having electrical contact elements (including bipolar plates/septums) comprising a titanium nitride coated light weight metal (e.g., Al or Ti) core, having a passivating, protective metal layer intermediate the core and the titanium nitride. The protective layer forms a barrier to further oxidation/corrosion when exposed to the fuel cell's operating environment. Stainless steels rich in CR, Ni, and Mo are particularly effective protective interlayers.

  13. Corrosion resistant PEM fuel cell

    DOEpatents

    Li, Yang; Meng, Wen-Jin; Swathirajan, Swathy; Harris, Stephen Joel; Doll, Gary Lynn

    2002-01-01

    The present invention contemplates a PEM fuel cell having electrical contact elements (including bipolar plates/septums) comprising a titanium nitride coated light weight metal (e.g., Al or Ti) core, having a passivating, protective metal layer intermediate the core and the titanium nitride. The protective layer forms a barrier to further oxidation/corrosion when exposed to the fuel cell's operating environment. Stainless steels rich in CR, Ni, and Mo are particularly effective protective interlayers.

  14. In situ fabrication of high-performance Ni-GDC-nanocube core-shell anode for low-temperature solid-oxide fuel cells

    PubMed Central

    Yamamoto, Kazuhiro; Qiu, Nan; Ohara, Satoshi

    2015-01-01

    A core–shell anode consisting of nickel–gadolinium-doped-ceria (Ni–GDC) nanocubes was directly fabricated by a chemical process in a solution containing a nickel source and GDC nanocubes covered with highly reactive {001} facets. The cermet anode effectively generated a Ni metal framework even at 500 °C with the growth of the Ni spheres. Anode fabrication at such a low temperature without any sintering could insert a finely nanostructured layer close to the interface between the electrolyte and the anode. The maximum power density of the attractive anode was 97 mW cm–2, which is higher than that of a conventional NiO–GDC anode prepared by an aerosol process at 55 mW cm–2 and 600 °C, followed by sintering at 1300 °C. Furthermore, the macro- and microstructure of the Ni–GDC-nanocube anode were preserved before and after the power-generation test at 700 °C. Especially, the reactive {001} facets were stabled even after generation test, which served to reduce the activation energy for fuel oxidation successfully. PMID:26615816

  15. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Dickson, J.J.

    1963-09-24

    A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

  16. Nuclear gas core propulsion research program

    NASA Technical Reports Server (NTRS)

    Diaz, Nils J.; Dugan, Edward T.; Anghaie, Samim

    1993-01-01

    Viewgraphs on the nuclear gas core propulsion research program are presented. The objectives of this research are to develop models and experiments, systems, and fuel elements for advanced nuclear thermal propulsion rockets. The fuel elements under investigation are suitable for gas/vapor and multiphase fuel reactors. Topics covered include advanced nuclear propulsion studies, nuclear vapor thermal rocket (NVTR) studies, and ultrahigh temperature nuclear fuels and materials studies.

  17. Monitoring arrangement for vented nuclear fuel elements

    DOEpatents

    Campana, Robert J.

    1981-01-01

    In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180.degree. rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements.

  18. Composite Cores

    NASA Technical Reports Server (NTRS)

    1990-01-01

    Spang & Company's new configuration of converter transformer cores is a composite of gapped and ungapped cores assembled together in concentric relationship. The net effect of the composite design is to combine the protection from saturation offered by the gapped core with the lower magnetizing requirement of the ungapped core. The uncut core functions under normal operating conditions and the cut core takes over during abnormal operation to prevent power surges and their potentially destructive effect on transistors. Principal customers are aerospace and defense manufacturers. Cores also have applicability in commercial products where precise power regulation is required, as in the power supplies for large mainframe computers.

  19. Nencki Affective Word List (NAWL): the cultural adaptation of the Berlin Affective Word List-Reloaded (BAWL-R) for Polish.

    PubMed

    Riegel, Monika; Wierzba, Małgorzata; Wypych, Marek; Żurawski, Łukasz; Jednoróg, Katarzyna; Grabowska, Anna; Marchewka, Artur

    2015-12-01

    In the present article, we introduce the Nencki Affective Word List (NAWL), created in order to provide researchers with a database of 2,902 Polish words, including nouns, verbs, and adjectives, with ratings of emotional valence, arousal, and imageability. Measures of several objective psycholinguistic features of the words (frequency, grammatical class, and number of letters) are also controlled. The database is a Polish adaptation of the Berlin Affective Word List-Reloaded (BAWL-R; Võ et al., Behavior Research Methods 41:534-538, 2009), commonly used to investigate the affective properties of German words. Affective normative ratings were collected from 266 Polish participants (136 women and 130 men). The emotional ratings and psycholinguistic indexes provided by NAWL can be used by researchers to better control the verbal materials they apply and to adjust them to specific experimental questions or issues of interest. The NAWL is freely accessible to the scientific community for noncommercial use as supplementary material to this article.

  20. Fissioning Plasma Core Reactor

    NASA Technical Reports Server (NTRS)

    Albright, Dennis; Butler, Carey; West, Nicole; Cole, John W. (Technical Monitor)

    2002-01-01

    Institute for Scientific Research, Inc. (ISR) research program consist of: 1.Study core physics by adapting existing codes: MCNP4C - Monte Carlo code; COMBINE/VENTURE - diffusion theory; SCALE4 - Monte Carlo, with many utility codes. 2. Determine feasibility and study major design parameters: fuel selection, temperature and reflector sizing. 3. Study reactor kinetics: develop QCALC1 to model point kinetics; study dynamic behavior of the power release.

  1. Alternative Fuels

    EPA Pesticide Factsheets

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  2. Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Third semiannual report, January-June 1980

    SciTech Connect

    Rosenbaum, H.S.

    1980-09-01

    Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the work scope of this program one of these concepts is to be selected for demonstration in a commercial power reactor. It was decided to demonstrate Zr-liner in 132 bundles which have liners of either crystal-bar zirconium or of low-oxygen sponge zirconium in the reload for Quad Cities Unit 2, Cycle 6. Irradiation testing or barrier fuel was continued, and the superior PCI resistance of Zr-liner fuel was further substantiated in the current report period. Furthermore, an irradiation experiment in which Zr-liner fuel, having a deliberately fabricated cladding perforation, was operated at a linear heat generation rate of 35 kW/m to a burnup of approx. 3 MWd/kg U showed no unusual signs of degradation compared with a similarly defected reference fuel rod. Four lead test assemblies of barrier fuel (two of Zr-liner and two of Cu-barrier), presently under irradiation in Quad Cities Unit 1, have achieved a burnup of 11 MWd/kg U.

  3. Cermet fuel reactors

    SciTech Connect

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  4. Fuel pin

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

    1989-10-03

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  5. Fuel pin

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

    1989-01-01

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  6. Fuel pin

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

    1987-11-24

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  7. Fuel pin

    SciTech Connect

    Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

    1987-11-24

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  8. Nuclear fuel element

    DOEpatents

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1980-04-29

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

  9. Fuel pump

    SciTech Connect

    Bellis, P.D.; Nesselrode, F.

    1991-04-16

    This patent describes a fuel pump. It includes: a fuel reservoir member, the fuel reservoir member being formed with fuel chambers, the chambers comprising an inlet chamber and an outlet chamber, means to supply fuel to the inlet chamber, means to deliver fuel from the outlet chamber to a point of use, the fuel reservoir member chambers also including a bypass chamber, means interconnecting the bypass chamber with the outlet chamber; the fuel pump also comprising pump means interconnecting the inlet chamber and the outlet chamber and adapted to suck fuel from the fuel supply means into the inlet chamber, through the pump means, out the outlet chamber, and to the fuel delivery means; the bypass chamber and the pump means providing two substantially separate paths of fuel flow in the fuel reservoir member, bypass plunger means normally closing off the flow of fuel through the bypass chamber one of the substantially separate paths including the fuel supply means and the fuel delivery means when the bypass plunger means is closed, the second of the substantially separate paths including the bypass chamber when the bypass plunger means is open, and all of the chambers and the interconnecting means therebetween being configured so as to create turbulence in the flow of any fuel supplied to the outlet chamber by the pump means and bypassed through the bypass chamber and the interconnecting means.

  10. NUCLEAR REACTOR CORE DESIGN

    DOEpatents

    Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

    1960-03-22

    An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

  11. Safety analysis of thorium-based fuels in the General Electric Standard BWR

    SciTech Connect

    Colby, M.J.; Townsend, D.B.; Kunz, C.L.

    1980-06-01

    A denatured (U-233/Th)O/sub 2/ fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO/sub 2/ fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O/sub 2/-fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O/sub 2/-fueled BWR should perform similar to a UO/sub 2/-fueled BWR under all operating conditions. A (Pu/Th)O/sub 2/-fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO/sub 2/-fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths.

  12. Molten core retention assembly

    DOEpatents

    Lampe, Robert F.

    1976-06-22

    Molten fuel produced in a core overheating accident is caught by a molten core retention assembly consisting of a horizontal baffle plate having a plurality of openings therein, heat exchange tubes having flow holes near the top thereof mounted in the openings, and a cylindrical, imperforate baffle attached to the plate and surrounding the tubes. The baffle assembly is supported from the core support plate of the reactor by a plurality of hanger rods which are welded to radial beams passing under the baffle plate and intermittently welded thereto. Preferably the upper end of the cylindrical baffle terminates in an outwardly facing lip to which are welded a plurality of bearings having slots therein adapted to accept the hanger rods.

  13. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2A, Physical descriptions of LWR (Light-Water Reactor) fuel assemblies

    SciTech Connect

    Not Available

    1987-12-01

    This appendix includes a four-page Physical Description report for each assembly type identified from the current data. Where available, a drawing of an assembly follows the appropriate Physical Description report. If no drawing is available for an assembly, a cross-reference to a similar assembly is provided if possible. For Advanced Nuclear Fuels, Babcock and Wilcox, Combustion Engineering, and Westinghouse assemblies, information was obtained via subcontracts with these fuel vendors. Data for some assembly types are not available. For such assemblies, the information shown in this report was obtained from the open literature and by inference from reload fuels made by other vendors. Efforts to obtain additional information are continuing. Individual Physical Description reports can be generated interactively through the menu-driven LWR Assemblies Data Base system. These reports can be viewed on the screen or directed to a printer. Special reports and compilations of specific data items can be produced on request.

  14. Fuel management studies of small metal and oxide LMR's

    SciTech Connect

    Khalil, H.; Fujita, E.K.; Yang, S.; Orechwa, Y.

    1986-01-01

    Fuel-cycle analyses performed at Argonne National Laboratory to evaluate and compare the neutronic performance characteristics of small oxide- and metal-fueled LMR's are described. Specific consideration is given to those analyses concerned with optimization of core and blanket configurations, selection of fuel residence time and refueling interval, determination of control rod worths and requirements, development of in-core fuel management strategy, and evaluation of performance characteristics both for startup cycles and for the equilibrium state reached via repeated recycle of discharged fuel. Differences in the computed performance parameters of oxide and metal cores, arising from basic differences in their neutronic characteristics, are identified and discussed. Metal-fueled cores are shown to offer some important performance advantages over oxide cores for small LMR's because of their harder spectrum, superior neutron economy, and greater breeding capacity. These advantages include smaller fissile and heavy metal loadings, lower control-system requirements, and greater adaptability to changes in fuel management scenarios.

  15. Energy Efficient Engine core design and performance report

    NASA Technical Reports Server (NTRS)

    Stearns, E. Marshall

    1982-01-01

    The Energy Efficient Engine (E3) is a NASA program to develop fuel saving technology for future large transport aircraft engines. Testing of the General Electric E3 core showed that the core component performance and core system performance necessary to meet the program goals can be achieved. The E3 core design and test results are described.

  16. Integral fast reactor concept. [Pool type; metal fuel; integral fuel cycle

    SciTech Connect

    Chang, Y.I.; Marchaterre, J.F.; Sevy, R.H.

    1984-01-01

    Key features of the IFR consist of a pool-type plant arrangement, a metal fuel-based core design, and an integral fuel cycle with colocated fuel cycle facility. Both the basic concept and the technology base have been demonstrated through actual integral cycle operation in EBR-II. This paper discusses the inherent safety characteristics of the IFR concept. (DLC)

  17. HTGR Fuel Technology Program. Semiannual report for the period ending March 31, 1983

    SciTech Connect

    Not Available

    1983-07-01

    This document reports the technical accomplishments of the HTGR Fuel Technology Program at GA Technologies Inc. during the first half of FY 83. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, core component verification, and core technology transfer tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR.

  18. Fuel property effects on fuel/air mixing in an experimental diesel engine

    SciTech Connect

    Browne, K.R.; Patridge, I.M.; Greeves, G.

    1986-01-01

    Fuels of widely varying properties are studied by injection of a single and well defined spray into an experimental diesel engine. Three optical techniques were developed to visualise liquid fuel, fuel vapour, flame, soot and individual droplets and their associated vapour trails. Liquid core length measurements are presented for diesel fuel, toluene, ethanol and sunflower oil. Computer model predictions show that an increase of the fuel mid-boiling point by 40/sup 0/C gives a similar effect on liquid core length to an increase of 0.03mm in the nozzle hole diameter.

  19. Extended life aluminide fuel. Final report

    SciTech Connect

    Miller, L.G.; Beeston, J.M.

    1986-06-01

    As the price of fuel fabrication, shipment of both new and spent fuel, and fuel reprocessing continue to rise at a rapid rate, researchers look for alternate methods to keep reactor fuel costs within their limited funding. Extended fuel element lifetimes, without jeopardizing reactor safety, can reduce fuel costs by up to a factor of two. The Extended Life Aluminide (ELAF) program was started at the Idaho National Engineering Laboratory (INEL) as a joint project of the United States Department of Energy (DOE), the University of Missouri, and the Massachusetts Institute of Technology research reactors. Fuel plates of Advanced Test Reactor (ATR) type construction were fabricated at Atomics International and irradiated in the ATR at the INEL. Four fuel matrix compositions were tested (i.e., 50 vol% UAl/sub x/ cores for reference, and 40, 45 and 50 vol% UAl/sub 2/ cores). The 50 vol% UAl/sub 2/ cores contained up to 3 grams U-235 per cm/sup 3/ of core. Three plates of each composition were irradiated to peak burnup levels of 3 x 10/sup 21/ fission/cm/sup 3/ of core. The only observed damage was due to external corrosion at similar rates experienced by UAl/sub x/ fuel elements in test reactors.

  20. The neutronic and fuel cycle performance of interchangeable 3500 MWth metal and oxide fueled LMRs

    SciTech Connect

    Fujita, E.K.; Wade, D.C.

    1989-03-01

    This study summarizes the neutronic and fuel cycle analysis performed at Argonne National Laboratory for an oxide and a metal fueled 3500 MWth LMR. The oxide and metal core designs were developed to meet reactor performance specifications that are constrained by requirements for core loading interchangeability and for small burnup reactivity swing. Differences in the computed performance parameters of the oxide and metal cores, arising from basic differences in their neutronic characteristics, were identified and discussed. It is shown that metal and oxide cores designed to the same ground rules exhibit many similar performance characteristics; however, they differ substantially in reactivity coefficients, control strategies, and fuel cycle options. 12 refs., 25 figs.

  1. NEUTRONIC REACTOR CORE

    DOEpatents

    Thomson, W.B.; Corbin, A. Jr.

    1961-07-18

    An improved core for a gas-cooled power reactor which admits gas coolant at high temperatures while affording strong integral supporting structure and efficient moderation of neutrons is described. The multiplicities of fuel elements constituting the critical amassment of fissionable material are supported and confined by a matrix of metallic structure which is interspersed therebetween. Thermal insulation is interposed between substantially all of the metallic matrix and the fuel elements; the insulation then defines the principal conduit system for conducting the coolant gas in heat-transfer relationship with the fuel elements. The metallic matrix itseif comprises a system of ducts through which an externally-cooled hydrogeneous liquid, such as water, is circulated to serve as the principal neutron moderant for the core and conjointly as the principal coolant for the insulated metallic structure. In this way, use of substantially neutron transparent metals, such as aluminum, becomes possible for the supporting structure, despite the high temperatures of the proximate gas. The Aircraft Nuclear Propulsion program's "R-1" reactor design is a preferred embodiment.

  2. Uranium droplet core nuclear rocket

    NASA Technical Reports Server (NTRS)

    Anghaie, Samim

    1991-01-01

    Uranium droplet nuclear rocket is conceptually designed to utilize the broad temperature range ofthe liquid phase of metallic uranium in droplet configuration which maximizes the energy transfer area per unit fuel volume. In a baseline system dissociated hydrogen at 100 bar is heated to 6000 K, providing 2000 second of Isp. Fission fragments and intense radian field enhance the dissociation of molecular hydrogen beyond the equilibrium thermodynamic level. Uranium droplets in the core are confined and separated by an axisymmetric vortex flow generated by high velocity tangential injection of hydrogen in the mid-core regions. Droplet uranium flow to the core is controlled and adjusted by a twin flow nozzle injection system.

  3. Uranium droplet core nuclear rocket

    NASA Technical Reports Server (NTRS)

    Anghaie, Samim

    1991-01-01

    Uranium droplet nuclear rocket is conceptually designed to utilize the broad temperature range ofthe liquid phase of metallic uranium in droplet configuration which maximizes the energy transfer area per unit fuel volume. In a baseline system dissociated hydrogen at 100 bar is heated to 6000 K, providing 2000 second of Isp. Fission fragments and intense radian field enhance the dissociation of molecular hydrogen beyond the equilibrium thermodynamic level. Uranium droplets in the core are confined and separated by an axisymmetric vortex flow generated by high velocity tangential injection of hydrogen in the mid-core regions. Droplet uranium flow to the core is controlled and adjusted by a twin flow nozzle injection system.

  4. Stability of Molten Core Materials

    SciTech Connect

    Layne Pincock; Wendell Hintze

    2013-01-01

    The purpose of this report is to document a literature and data search for data and information pertaining to the stability of nuclear reactor molten core materials. This includes data and analysis from TMI-2 fuel and INL’s LOFT (Loss of Fluid Test) reactor project and other sources.

  5. Core Practices: Fuel Superintendents' Equity Focus

    ERIC Educational Resources Information Center

    Thompson, Scott

    2016-01-01

    For eight years, more than a dozen district superintendents in New Jersey have joined together for a full day each month during the school year to listen to and learn from each other as a community of practice. Known as the New Jersey Network of Superintendents, this community of practice has a tight focus on advancing equity through improvement…

  6. Core Practices: Fuel Superintendents' Equity Focus

    ERIC Educational Resources Information Center

    Thompson, Scott

    2016-01-01

    For eight years, more than a dozen district superintendents in New Jersey have joined together for a full day each month during the school year to listen to and learn from each other as a community of practice. Known as the New Jersey Network of Superintendents, this community of practice has a tight focus on advancing equity through improvement…

  7. Experiments on the Distribution of Fuel in Fuel Sprays

    NASA Technical Reports Server (NTRS)

    Lee, Dana W

    1933-01-01

    The distribution of fuel in sprays for compression-ignition engines was investigated by taking high-speed spark photographs of fuel sprays reproduced under a wide variety of conditions, and also by injecting them against pieces of plasticine. A photographic study was made of sprays injected into evacuated chambers, into the atmosphere, into compressed air, and into transparent liquids. Pairs of identical sprays were injected counter to each other and their behavior analyzed. Small high velocity air jets were directed normally to the axes of fuel sprays, with the result that the envelope of spray which usually obscures the core was blown aside, leaving the core exposed on one side. The results showed that the distribution of the fuel within the sprays was very uneven.

  8. Impact of fuel fabrication and fuel management technologies on uranium utilization

    SciTech Connect

    Arnsberger, P.L.; Stucker, D.L.

    1994-12-31

    Uranium utilization in commercial pressurized water reactors is a complex function of original NSSS design, utility energy requirements, fuel assembly design, fuel fabrication materials and fuel management optimization. Fuel design and fabrication technologies have reacted to the resulting market forcing functions with a combination of design and material changes. The technologies employed have included ever-increasing fuel discharge burnup, non-parasitic structural materials, burnable absorbers, and fissile material core zoning schemes (both in the axial and radial direction). The result of these technological advances has improved uranium utilization by roughly sixty percent from the infancy days of nuclear power to present fuel management. Fuel management optimization technologies have also been developed in recent years which provide fuel utilization improvements due to core loading pattern optimization. This paper describes the development and impact of technology advances upon uranium utilization in modem pressurized water reactors.

  9. Synthetic Fuel

    SciTech Connect

    Idaho National Laboratory - Steve Herring, Jim O'Brien, Carl Stoots

    2008-03-26

    Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhouse gass Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhous

  10. Synthetic Fuel

    ScienceCinema

    Idaho National Laboratory - Steve Herring, Jim O'Brien, Carl Stoots

    2016-07-12

    Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhouse gass Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhous

  11. Fuel cells

    NASA Astrophysics Data System (ADS)

    1984-12-01

    The US Department of Energy (DOE), Office of Fossil Energy, has supported and managed a fuel cell research and development (R and D) program since 1976. Responsibility for implementing DOE's fuel cell program, which includes activities related to both fuel cells and fuel cell systems, has been assigned to the Morgantown Energy Technology Center (METC) in Morgantown, West Virginia. The total United States effort of the private and public sectors in developing fuel cell technology is referred to as the National Fuel Cell Program (NFCP). The goal of the NFCP is to develop fuel cell power plants for base-load and dispersed electric utility systems, industrial cogeneration, and on-site applications. To achieve this goal, the fuel cell developers, electric and gas utilities, research institutes, and Government agencies are working together. Four organized groups are coordinating the diversified activities of the NFCP. The status of the overall program is reviewed in detail.

  12. Proceedings of the 2007 LWR Fuel Performance Meeting / TopFuel 2007 'Zero by 2010'

    SciTech Connect

    2007-07-01

    ANS, ENS, AESJ and KNS are jointly organizing the 2007 International LWR Fuel Performance Meeting following the successful ENS TopFuel meeting held during 22-26 October, 2006 in Salamaca, Spain. Merging three premier nuclear fuel design and performance meetings: the ANS LWR Fuel Performance Meeting, the ENS TopFuel and Asian Water Reactor Fuel Performance Meeting (WRFPM) created this international meeting. The meeting will be held annually on a tri-annual rotational basis in USA, Asia, and Europe. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as performance experience in commercial and test reactors. The meeting excludes front end and back end fuel issues, however, it covers all front and/or back issues that impact fuel designs and performance.

  13. Numerical Investigation of the Macroscopic Mechanical Behavior of NiTi-Hybrid Composites Subjected to Static Load-Unload-Reload Path

    NASA Astrophysics Data System (ADS)

    Taheri-Behrooz, Fathollah; Kiani, Ali

    2017-04-01

    Shape memory alloys (SMAs) are a type of shape memory materials that recover large deformation and return to their primary shape by rising temperature. In the current research, the effect of embedding SMA wires on the macroscopic mechanical behavior of glass-epoxy composites is investigated through finite element simulations. A perfect interface between SMA wires and the host composite is assumed. Effects of various parameters such as SMA wires volume fraction, SMA wires pre-strain and temperature are investigated during loading-unloading and reloading steps by employing ANSYS software. In order to quantify the extent of induced compressive stress in the host composite and residual tensile stress in the SMA wires, a theoretical approach is presented. Finally, it was shown that smart structures fabricated using composite layers and pre-strained SMA wires exhibited overall stiffness reduction at both ambient and elevated temperatures which were increased by adding SMA volume fraction. Also, the induced compressive stress on the host composite was increased remarkably using 4% pre-strained SMA wires at elevated temperature. Results obtained by FE simulations were in good correlation with the rule of mixture predictions and available experimental data in the literature.

  14. Numerical Investigation of the Macroscopic Mechanical Behavior of NiTi-Hybrid Composites Subjected to Static Load-Unload-Reload Path

    NASA Astrophysics Data System (ADS)

    Taheri-Behrooz, Fathollah; Kiani, Ali

    2017-02-01

    Shape memory alloys (SMAs) are a type of shape memory materials that recover large deformation and return to their primary shape by rising temperature. In the current research, the effect of embedding SMA wires on the macroscopic mechanical behavior of glass-epoxy composites is investigated through finite element simulations. A perfect interface between SMA wires and the host composite is assumed. Effects of various parameters such as SMA wires volume fraction, SMA wires pre-strain and temperature are investigated during loading-unloading and reloading steps by employing ANSYS software. In order to quantify the extent of induced compressive stress in the host composite and residual tensile stress in the SMA wires, a theoretical approach is presented. Finally, it was shown that smart structures fabricated using composite layers and pre-strained SMA wires exhibited overall stiffness reduction at both ambient and elevated temperatures which were increased by adding SMA volume fraction. Also, the induced compressive stress on the host composite was increased remarkably using 4% pre-strained SMA wires at elevated temperature. Results obtained by FE simulations were in good correlation with the rule of mixture predictions and available experimental data in the literature.

  15. Future Fuels

    DTIC Science & Technology

    2005-10-04

    tactical ground mobility and increasing operational reach • Identify, review, and assess – Technologies for reducing fuel consumption, including...T I O N S A C T I O N S TOR Focus - Tactical ground mobility - Operational reach - Not A/C, Ships, or troops Hybrid Electric Vehicle Fuel Management...Fuel Management During Combat Operations Energy Fundamentals • Energy Density • Tactical Mobility • Petroleum Use • Fuel Usage (TWV) • TWV OP TEMPO TOR

  16. Alternative Fuels

    DTIC Science & Technology

    2009-06-11

    JP-8 BACK-UP SLIDES Unclassified 19 What Are Biofuels ? Cellulose “first generation”“second generation” C18:0 C16:1 Triglycerides (fats, oils ...equipment when supplying jet fuel not practicable or cost effective Unclassified 5 erna ve ue s ocus Petroleum Crude Oil (declining discovery / production...on Jet A/A-1 Approved fuels, DXXXX Unclassified 6 JP-8/5 (Commercial Jet Fuel, ASTM Spec) DARPA Alternative Jet Fuels • Agricultural crop oils

  17. Methodology for embedded transport core calculation

    NASA Astrophysics Data System (ADS)

    Ivanov, Boyan D.

    The progress in the Nuclear Engineering field leads to developing new generations of Nuclear Power Plants (NPP) with complex rector core designs, such as cores loaded partially with mixed-oxide (MOX) fuel, high burn-up loadings, and cores with advanced designs of fuel assemblies and control rods. Such heterogeneous cores introduce challenges for the diffusion theory that has been used for several decades for calculations of the current Pressurized Water Rector (PWR) cores. To address the difficulties the diffusion approximation encounters new core calculation methodologies need to be developed by improving accuracy, while preserving efficiency of the current reactor core calculations. In this thesis, an advanced core calculation methodology is introduced, based on embedded transport calculations. Two different approaches are investigated. The first approach is based on embedded finite element (FEM), simplified P3 approximation (SP3), fuel assembly (FA) homogenization calculation within the framework of the diffusion core calculation with NEM code (Nodal Expansion Method). The second approach involves embedded FA lattice physics eigenvalue calculation based on collision probability method (CPM) again within the framework of the NEM diffusion core calculation. The second approach is superior to the first because most of the uncertainties introduced by the off-line cross-section generation are eliminated.

  18. Fossil Fuels.

    ERIC Educational Resources Information Center

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  19. Fossil Fuels.

    ERIC Educational Resources Information Center

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  20. Alternative fuels

    NASA Technical Reports Server (NTRS)

    Grobman, J. S.; Butze, H. F.; Friedman, R.; Antoine, A. C.; Reynolds, T. W.

    1977-01-01

    Potential problems related to the use of alternative aviation turbine fuels are discussed and both ongoing and required research into these fuels is described. This discussion is limited to aviation turbine fuels composed of liquid hydrocarbons. The advantages and disadvantages of the various solutions to the problems are summarized. The first solution is to continue to develop the necessary technology at the refinery to produce specification jet fuels regardless of the crude source. The second solution is to minimize energy consumption at the refinery and keep fuel costs down by relaxing specifications.

  1. Fuel compositions

    SciTech Connect

    Mekonen, K.

    1989-10-31

    This patent describes a hydrosol fuel. It comprises: from about 67% to 94% by weight of a hydrocarbon combustible fuel selected from the group consisting of the gasolines, diesel fuels and heavy fuel oils, from 5 to 25% by weight of water, at least one surfactant operable to create a hydrosol with the fuel and water present in the range of 0.1 up to about 3.4% by weight of an additive selected from the group consisting of alpha (mono) olefins and alkyl benzenes, each of the former having 7 to 15 recurring C{sub 2} monomers therein.

  2. Accident tolerant fuels for LWRs: A perspective

    NASA Astrophysics Data System (ADS)

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

    2014-05-01

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

  3. Core Optimization of a Deep-Burn Pebble Bed Reactor

    SciTech Connect

    Brian Boer; Abderrafi M. Ougouag

    2010-06-01

    Achieving a high fuel burnup in the Deep-Burn (DB) pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum as compared to a ’standard’ UO2 fueled core. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. The DB concept focuses on the destruction of spent fuel transuranics in TRISO coated particle fueled gas-cooled reactors with the aim of a fractional fuel burnup of 60-70% in fissions per initial metal atom (FIMA), using a single-pass, multi in-core fuel (re)cycling scheme. In principle, the DB pebble bed concept employs the same reactor designs as the present low enriched uranium core designs, i.e. the 400 MWth Pebble Bed Modular Reactor (PBMR-400). A Pu and Minor Actinide fueled PBMR-400 design serves as the starting point for a core optimization study. The fuel temperature, power peak, temperature reactivity coefficients, and burnup capabilities of the modified designs are analyzed with the PEBBED code. A code-to-code coupling with the PASTA code allows for the analysis of the TRISO fuel performance for both normal and Loss Of Forced Cooling conditions. An improved core design is sought, maximizing the fuel discharge burnup, while retaining negative temperature reactivity feedback coefficients for the entire temperature range and avoiding high fuel temperatures (fuel failure probabilities).

  4. Radioactivity of spent TRIGA fuel

    SciTech Connect

    Usang, M. D. Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-04-29

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  5. Radioactivity of spent TRIGA fuel

    NASA Astrophysics Data System (ADS)

    Usang, M. D.; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-04-01

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  6. 24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES ...

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

    24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES FOR A BRASS GATE VALVE BODY MADE ON A CORE BOX, CA. 1950. - Stockham Pipe & Fittings Company, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL

  7. Double-clad nuclear fuel safety rod

    DOEpatents

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  8. Fuel handling system for a nuclear reactor

    DOEpatents

    Saiveau, James G.; Kann, William J.; Burelbach, James P.

    1986-01-01

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  9. Fuel handling system for a nuclear reactor

    DOEpatents

    Saiveau, James G.; Kann, William J.; Burelbach, James P.

    1986-12-02

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  10. Plasma core reactor applications

    NASA Technical Reports Server (NTRS)

    Latham, T. S.; Rodgers, R. J.

    1976-01-01

    Analytical and experimental investigations were conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration. Axial working fluid channels are located along a fraction of each cavity peripheral wall. Results of calculations for outward-directed radiant energy fluxes corresponding to radiating temperatures of 2000 to 5000 K indicate total operating pressures from 80 to 650 atm, centerline temperatures from 6900 to 30,000 K, and total radiated powers from 25 to 2500 MW, respectively. Applications are described for this type of reactor such as (1) high-thrust, high specific impulse space propulsion, (2) highly efficient systems for generation of electricity, and (3) hydrogen or synthetic fuel production systems using the intense radiant energy fluxes.

  11. Plasma core reactor applications

    NASA Technical Reports Server (NTRS)

    Latham, T. S.; Rodgers, R. J.

    1976-01-01

    Analytical and experimental investigations were conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration. Axial working fluid channels are located along a fraction of each cavity peripheral wall. Results of calculations for outward-directed radiant energy fluxes corresponding to radiating temperatures of 2000 to 5000 K indicate total operating pressures from 80 to 650 atm, centerline temperatures from 6900 to 30,000 K, and total radiated powers from 25 to 2500 MW, respectively. Applications are described for this type of reactor such as (1) high-thrust, high specific impulse space propulsion, (2) highly efficient systems for generation of electricity, and (3) hydrogen or synthetic fuel production systems using the intense radiant energy fluxes.

  12. Affective Norms for 4900 Polish Words Reload (ANPW_R): Assessments for Valence, Arousal, Dominance, Origin, Significance, Concreteness, Imageability and, Age of Acquisition.

    PubMed

    Imbir, Kamil K

    2016-01-01

    In studies that combine understanding of emotions and language, there is growing demand for good-quality experimental materials. To meet this expectation, a large number of 4905 Polish words was assessed by 400 participants in order to provide a well-established research method for everyone interested in emotional word processing. The Affective Norms for Polish Words Reloaded (ANPW_R) is designed as an extension to the previously introduced the ANPW dataset and provides assessments for eight different affective and psycholinguistic measures of Valence, Arousal, Dominance, Origin, Significance, Concreteness, Imageability, and subjective Age of Acquisition. The ANPW_R is now the largest available dataset of affective words for Polish, including affective scores that have not been measured in any other dataset (concreteness and age of acquisition scales). Additionally, the ANPW_R allows for testing hypotheses concerning dual-mind models of emotion and activation (origin and subjective significance scales). Participants in the current study assessed all 4905 words in the list within 1 week, at their own pace in home sessions, using eight different Self-assessment Manikin (SAM) scales. Each measured dimension was evaluated by 25 women and 25 men. The ANPW_R norms appeared to be reliable in split-half estimation and congruent with previous normative studies in Polish. The quadratic relation between valence and arousal was found to be in line with previous findings. In addition, nine other relations appeared to be better described by quadratic instead of linear function. The ANPW_R provides well-established research materials for use in psycholinguistic and affective studies in Polish-speaking samples.

  13. Affective Norms for 4900 Polish Words Reload (ANPW_R): Assessments for Valence, Arousal, Dominance, Origin, Significance, Concreteness, Imageability and, Age of Acquisition

    PubMed Central

    Imbir, Kamil K.

    2016-01-01

    In studies that combine understanding of emotions and language, there is growing demand for good-quality experimental materials. To meet this expectation, a large number of 4905 Polish words was assessed by 400 participants in order to provide a well-established research method for everyone interested in emotional word processing. The Affective Norms for Polish Words Reloaded (ANPW_R) is designed as an extension to the previously introduced the ANPW dataset and provides assessments for eight different affective and psycholinguistic measures of Valence, Arousal, Dominance, Origin, Significance, Concreteness, Imageability, and subjective Age of Acquisition. The ANPW_R is now the largest available dataset of affective words for Polish, including affective scores that have not been measured in any other dataset (concreteness and age of acquisition scales). Additionally, the ANPW_R allows for testing hypotheses concerning dual-mind models of emotion and activation (origin and subjective significance scales). Participants in the current study assessed all 4905 words in the list within 1 week, at their own pace in home sessions, using eight different Self-assessment Manikin (SAM) scales. Each measured dimension was evaluated by 25 women and 25 men. The ANPW_R norms appeared to be reliable in split-half estimation and congruent with previous normative studies in Polish. The quadratic relation between valence and arousal was found to be in line with previous findings. In addition, nine other relations appeared to be better described by quadratic instead of linear function. The ANPW_R provides well-established research materials for use in psycholinguistic and affective studies in Polish-speaking samples. PMID:27486423

  14. Fuel compositions

    SciTech Connect

    Zaweski, E.F.; Niebylski, L.M.

    1986-08-05

    This patent describes distillate fuel for indirect injection compression ignition engines containing, in an amount sufficient to minimize coking, especially throttling nozzle coking in the prechambers or swirl chambers of indirect injection compression ignition engines operated on such fuel, at least the combination of (i) organic nitrate ignition accelerator and (ii) an esterified cycle dehydration product of sorbitol which, when added to the fuel in combination with the organic nitrate ignition accelerator minimizes the coking.

  15. Fuel cells 101

    SciTech Connect

    Hirschenhofer, J.H.

    1999-07-01

    This paper discusses the various types of fuel cells, the importance of cell voltage, fuel processing for natural gas, cell stacking, fuel cell plant description, advantages and disadvantages of the types of fuel cells, and applications. The types covered include: polymer electrolyte fuel cell, alkaline fuel cell, phosphoric acid fuel cell; molten carbonate fuel cell, and solid oxide fuel cell.

  16. Convection, nucleosynthesis, and core collapse

    NASA Technical Reports Server (NTRS)

    Bazan, Grant; Arnett, David

    1994-01-01

    We use a piecewise parabolic method hydrodynamics code (PROMETHEUS) to study convective burning in two dimensions in an oxygen shell prior to core collapse. Significant mixing beyond convective boundaries determined by mixing-length theory brings fuel (C-12) into the convective regon, causing hot spots of nuclear burning. Plumes dominate the velocity structure. Finite perturbations arise in a region in which O-16 will be explosively burned to Ni-56 when the star explodes; the resulting instabilities and mixing are likely to distribute Ni-56 throughout the supernova envelope. Inhomogeneities in Y(sub e) may be large enough to affect core collapse and will affect explosive nucleosynthesis. The nature of convective burning is dramatically different from that assumed in one-dimensional simulations; quantitative estimates of nucleosynthetic yields, core masses, and the approach to core collapse will be affected.

  17. Motor fuel

    SciTech Connect

    Burns, L.D.

    1982-07-13

    Liquid hydrocarbon fuel compositions are provided containing antiknock quantities of ashless antiknock agents comprising selected furyl compounds including furfuryl alcohol, furfuryl amine, furfuryl esters, and alkyl furoates.

  18. Fuel dehazers

    SciTech Connect

    Lyons, W.R.

    1986-03-01

    Hazy fuels can be caused by the emulsification of water into the fuel during refining, blending, or transportation operations. Detergent additive packages used in gasoline tend to emulsify water into the fuel. Fuels containing water haze can cause corrosion and contamination, and support microbiological growth. This results in problems. As the result of these problems, refiners, marketers, and product pipeline companies customarily have haze specifications. The haze specification may be a specific maximum water content or simply ''bright and clear'' at a specified temperature.

  19. Alternative fuels

    SciTech Connect

    Not Available

    1991-07-01

    This paper presents the preliminary results of a review, of the experiences of Brazil, Canada, and New Zealand, which have implemented programs to encourage the use of alternative motor fuels. It will also discuss the results of a separate completed review of the Department of Energy's (DOE) progress in implementing the Alternative Motor Fuels Act of 1988. The act calls for, among other things, the federal government to use alternative-fueled vehicles in its fleet. The Persian Gulf War, environmental concerns, and the administration's National Energy Strategy have greatly heightened interest in the use of alternative fuels in this country.

  20. Updated NGNP Fuel Acquisition Strategy

    SciTech Connect

    David Petti; Tim Abram; Richard Hobbins; Jim Kendall

    2010-12-01

    A Next Generation Nuclear Plant (NGNP) fuel acquisition strategy was first established in 2007. In that report, a detailed technical assessment of potential fuel vendors for the first core of NGNP was conducted by an independent group of international experts based on input from the three major reactor vendor teams. Part of the assessment included an evaluation of the credibility of each option, along with a cost and schedule to implement each strategy compared with the schedule and throughput needs of the NGNP project. While credible options were identified based on the conditions in place at the time, many changes in the assumptions underlying the strategy and in externalities that have occurred in the interim requiring that the options be re-evaluated. This document presents an update to that strategy based on current capabilities for fuel fabrication as well as fuel performance and qualification testing worldwide. In light of the recent Pebble Bed Modular Reactor (PBMR) project closure, the Advanced Gas Reactor (AGR) fuel development and qualification program needs to support both pebble and prismatic options under the NGNP project. A number of assumptions were established that formed a context for the evaluation. Of these, the most important are: • Based on logistics associated with the on-going engineering design activities, vendor teams would start preliminary design in October 2012 and complete in May 2014. A decision on reactor type will be made following preliminary design, with the decision process assumed to be completed in January 2015. Thus, no fuel decision (pebble or prismatic) will be made in the near term. • Activities necessary for both pebble and prismatic fuel qualification will be conducted in parallel until a fuel form selection is made. As such, process development, fuel fabrication, irradiation, and testing for pebble and prismatic options should not negatively influence each other during the period prior to a decision on reactor type

  1. Fuel elements of research reactor CM

    SciTech Connect

    Kozlov, A.V.; Morozov, A.V.; Vatulin, A.V.; Ershov, S.A.

    2013-07-01

    In 1961 the CM research reactor was commissioned at the Research Institute of Atomic Reactors (Dimitrovgrad, Russia), it was intended to carry on investigations and the production of transuranium nuclides. The reactor is of a tank type. Original fuel assembly contained plate fuels that were spaced with vanes and corrugated bands. Nickel was used as a cladding material, fuel meat was produced from UO{sub 2} + electrolytic nickel composition. Fuel plates have been replaced by self-spacing cross-shaped dispersion fuels clad in stainless steel. In 2005 the reactor was updated. The purpose of this updating was to increase the quantity of irradiation channels in the reactor core and to improve the neutron balance. The updating was implemented at the expense of 20 % reduction in the quantity of fuel elements in the core which released a space for extra channels and decreased the mass of structural materials in the core. The updated reactor is loaded with modified standard fuel elements with 20 % higher uranium masses. At the same time stainless steel in fuel assembly shrouds was substituted by zirconium alloy. Today in progress are investigations and work to promote the second stage of reactor updating that involve developments of cross-shaped fuel elements having low neutron absorption matrix materials. This article gives an historical account of the design and main technical changes that occurred for the CM reactor since its commissioning.

  2. Lunar Polar Coring Lander

    NASA Technical Reports Server (NTRS)

    Angell, David; Bealmear, David; Benarroche, Patrice; Henry, Alan; Hudson, Raymond; Rivellini, Tommaso; Tolmachoff, Alex

    1990-01-01

    Plans to build a lunar base are presently being studied with a number of considerations. One of the most important considerations is qualifying the presence of water on the Moon. The existence of water on the Moon implies that future lunar settlements may be able to use this resource to produce things such as drinking water and rocket fuel. Due to the very high cost of transporting these materials to the Moon, in situ production could save billions of dollars in operating costs of the lunar base. Scientists have suggested that the polar regions of the Moon may contain some amounts of water ice in the regolith. Six possible mission scenarios are suggested which would allow lunar polar soil samples to be collected for analysis. The options presented are: remote sensing satellite, two unmanned robotic lunar coring missions (one is a sample return and one is a data return only), two combined manned and robotic polar coring missions, and one fully manned core retrieval mission. One of the combined manned and robotic missions has been singled out for detailed analysis. This mission proposes sending at least three unmanned robotic landers to the lunar pole to take core samples as deep as 15 meters. Upon successful completion of the coring operations, a manned mission would be sent to retrieve the samples and perform extensive experiments of the polar region. Man's first step in returning to the Moon is recommended to investigate the issue of lunar polar water. The potential benefits of lunar water more than warrant sending either astronauts, robots or both to the Moon before any permanent facility is constructed.

  3. Core transfer

    NASA Astrophysics Data System (ADS)

    Good news for all petroleum geoscientists, mining and environmental scientists, university researchers, and the like: Shell Oil Company has deeded its Midland core and sample repository to the Bureau of Economic Geology (BEG) at the University of Texas at Austin. The Midland repository includes more than 1 million linear meters of slab, whole core, and prepared cuttings. Data comprising one of the largest U.S. core collections—the geologic samples from wells drilled in Texas and 39 other states—are now public data and will be incorporated into the existing BEG database. Both Shell and the University of Texas at Austin are affiliated with the American Geological Institute, which assisted in arranging the transfer as part of its goal to establish a National Geoscience Data Repository System at regional centers across the United States.

  4. BWR fuel design options for self-sustainable Th-{sup 233}U fuel cycle

    SciTech Connect

    Shaposhnik, Y.; Shwageraus, E.; Elias, E.

    2012-07-01

    In this work, we investigate a number of fuel assembly design options for a BWR core operating in a closed self-sustainable Th-{sup 233}U fuel cycle. The designs rely on axially heterogeneous fuel assembly structure in order to improve fertile to fissile conversion ratio. One of the main assumptions of the current study was to restrict the fuel assembly geometry to a single axial fissile zone 'sandwiched' between two fertile blanket zones. The main objective was to study the effect of the most important design parameters, such as dimensions of fissile and fertile zones and average void fraction, on the net breeding of {sup 233}U. The main design challenge in this respect is that the fuel breeding potential is at odds with axial power peaking and therefore limits the maximum achievable core power rating. The calculations were performed with BGCore system, which consists of MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly with reflective radial boundaries was modeled applying simplified restrictions on maximum central line fuel temperature and Critical Power Ratio. It was found that axially heterogeneous fuel assembly design with single fissile zone can potentially achieve net breeding. In this case however, the achievable core power density is roughly one third of the reference BWR core. (authors)

  5. CFD Analysis of Core Bypass Phenomena

    SciTech Connect

    Richard W. Johnson; Hiroyuki Sato; Richard R. Schultz

    2010-03-01

    The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the sector grid can be set as a symmetry boundary

  6. CFD Analysis of Core Bypass Phenomena

    SciTech Connect

    Richard W. Johnson; Hiroyuki Sato; Richard R. Schultz

    2009-11-01

    The U.S. Department of Energy is exploring the potential for the VHTR which will be either of a prismatic or a pebble-bed type. One important design consideration for the reactor core of a prismatic VHTR is coolant bypass flow which occurs in the interstitial regions between fuel blocks. Such gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The existence of the gaps induces a flow bias in the fuel blocks and results in unexpected increase of maximum fuel temperature. Traditionally, simplified methods such as flow network calculations employing experimental correlations are used to estimate flow and temperature distributions in the core design. However, the distribution of temperature in the fuel pins and graphite blocks as well as coolant outlet temperatures are strongly coupled with the local heat generation rate within fuel blocks which is not uniformly distributed in the core. Hence, it is crucial to establish mechanistic based methods which can be applied to the reactor core thermal hydraulic design and safety analysis. Computational Fluid Dynamics (CFD) codes, which have a capability of local physics based simulation, are widely used in various industrial fields. This study investigates core bypass flow phenomena with the assistance of commercial CFD codes and establishes a baseline for evaluation methods. A one-twelfth sector of the hexagonal block surface is modeled and extruded down to whole core length of 10.704m. The computational domain is divided vertically with an upper reflector, a fuel section and a lower reflector. Each side of the one-twelfth grid can be set as a symmetry boundary

  7. Fuel injector

    DOEpatents

    Lambeth, Malcolm David Dick

    2001-02-27

    A fuel injector comprises first and second housing parts, the first housing part being located within a bore or recess formed in the second housing part, the housing parts defining therebetween an inlet chamber, a delivery chamber axially spaced from the inlet chamber, and a filtration flow path interconnecting the inlet and delivery chambers to remove particulate contaminants from the flow of fuel therebetween.

  8. Future Fuels

    DTIC Science & Technology

    2006-04-01

    financing, push technology and help motivate the building of the necessary manufacturing and distribution infrastructure. Hybrid Electric Vehicles , Tether...conclusions in three major areas: Hybrid Electric Vehicles (HEVs), fuel management during combat operations and manufactured fuels to address the...payoffs in the relatively near term, are: • Hybrid Electric Vehicles : The development of and commitment to hybrid electric architecture for TWVs

  9. Microyielding of core-shell crystal dendrites in a bulk-metallic-glass matrix composite

    SciTech Connect

    Huang, E. -Wen; Qiao, Junwei; Winiarski, Bartlomiej; Lee, Wen -Jay; Scheel, Mario; Chuang, Chih -Pin; Liaw, Peter K.; Lo, Yu -Chieh; Zhang, Yong; Di Michiel, Marco

    2014-03-18

    In-situ synchrotron x-ray experiments have been used to follow the evolution of the diffraction peaks for crystalline dendrites embedded in a bulk metallic glass matrix subjected to a compressive loading-unloading cycle. We observe irreversible diffraction-peak splitting even though the load does not go beyond half of the bulk yield strength. The chemical analysis coupled with the transmission electron microscopy mapping suggests that the observed peak splitting originates from the chemical heterogeneity between the core (major peak) and the stiffer shell (minor peak) of the dendrites. A molecular dynamics model has been developed to compare the hkl-dependent microyielding of the bulk metallic-glass matrix composite. As a result, the complementary diffraction measurements and the simulation results suggest that the interfaces between the amorphous matrix and the (211) crystalline planes relax under prolonged load that causes a delay in the reload curve which ultimately catches up with the original path.

  10. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    SciTech Connect

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  11. Mechanical Tests and Analyses on the FBR Ductless Fuel Assembly

    SciTech Connect

    Itoh, K.; Sato, T.; Ogura, M.; Ohkubo, Y.; Moro, S.; Madarame, H.

    2002-07-01

    Fast Breeder Reactor (FBR) cores, which are composed of ductless fuel assemblies, have many potentialities of cost reduction through whole fuel cycle procedures with safety features, and of reduced high-level-waste disposal. The mechanistic aspects of the ductless fuel assembly have been investigated by mechanical tests and analyses, and the core static behaviors under the irradiation and the seismic occasion have been clarified in this study. (authors)

  12. Yankee Rowe simulator core model validation

    SciTech Connect

    Napolitano, M.E.

    1990-01-01

    This paper presents the validation of the Yankee Rowe simulator core model. Link-Miles Simulation Corporation is developing the Yankee Rowe simulator and Yankee Atomic Electric Company is involved in input and benchmark data generation, as well as simulator validation. Core model validation by Yankee comprises three tasks: (1) careful generation of fuel reactivity characteristics (B constants); (2) nonintegrated core model testing; and (3) fully integrated core model testing. Simulator core model validation and verification is a multistage process involving input and benchmark data generation as well as interactive debugging. Core characteristics were brought within acceptable criteria by this process. This process was achieved through constant communication between Link-Miles and Yankee engineers. Based on this validation, the Yankee Rowe simulator core model is found to be acceptable for training purposes.

  13. Core layering

    NASA Astrophysics Data System (ADS)

    Jacobson, S. A.; Rubie, D. C.; Hernlund, J. W.; Morbidelli, A.

    2015-12-01

    We have created a planetary accretion and differentiation model that self-consistently builds and evolves Earth's core. From this model, we show that the core grows stably stratified as the result of rising metal-silicate equilibration temperatures and pressures, which increases the concentrations of light element impurities into each newer core addition. This stable stratification would naturally resist convection and frustrate the onset of a geodynamo, however, late giant impacts could mechanically mix the distinct accreted core layers creating large homogenous regions. Within these regions, a geodynamo may operate. From this model, we interpret the difference between the planetary magnetic fields of Earth and Venus as a difference in giant impact histories. Our planetary accretion model is a numerical N-body integration of the Grand Tack scenario [1]—the most successful terrestrial planet formation model to date [2,3]. Then, we take the accretion histories of Earth-like and Venus-like planets from this model and post-process the growth of each terrestrial planet according to a well-tested planetary differentiation model [4,5]. This model fits Earth's mantle by modifying the oxygen content of the pre-cursor planetesimals and embryos as well as the conditions of metal-silicate equilibration. Other non-volatile major, minor and trace elements included in the model are assumed to be in CI chondrite proportions. The results from this model across many simulated terrestrial planet growth histories are robust. If the kinetic energy delivered by larger impacts is neglected, the core of each planet grows with a strong stable stratification that would significantly impede convection. However, if giant impact mixing is very efficient or if the impact history delivers large impacts late, than the stable stratification can be removed. [1] Walsh et al. Nature 475 (2011) [2] O'Brien et al. Icarus 223 (2014) [3] Jacobson & Morbidelli PTRSA 372 (2014) [4] Rubie et al. EPSL 301

  14. Fuel cell-fuel cell hybrid system

    DOEpatents

    Geisbrecht, Rodney A.; Williams, Mark C.

    2003-09-23

    A device for converting chemical energy to electricity is provided, the device comprising a high temperature fuel cell with the ability for partially oxidizing and completely reforming fuel, and a low temperature fuel cell juxtaposed to said high temperature fuel cell so as to utilize remaining reformed fuel from the high temperature fuel cell. Also provided is a method for producing electricity comprising directing fuel to a first fuel cell, completely oxidizing a first portion of the fuel and partially oxidizing a second portion of the fuel, directing the second fuel portion to a second fuel cell, allowing the first fuel cell to utilize the first portion of the fuel to produce electricity; and allowing the second fuel cell to utilize the second portion of the fuel to produce electricity.

  15. TMI-2 core shipping preparations

    SciTech Connect

    Ball, L.J.; ); Barkanic, R.J. ); Conaway, W.T. II ); Schmoker, D.S. )

    1988-01-01

    Shipping the damaged core from the Unit 2 reactor of Three Mile Island Nuclear Power Station near Harrisburg, PA, to the Idaho National Engineering Laboratory near Idaho Falls, ID, required development and implementation of a completely new spent fuel transportation system. This paper describes the equipment developed, the planning and activities used to implement the hardware systems into the facilities, and the planning involved in making the rail shipments. It also includes a summary of recommendations resulting from this experience.

  16. The closed fuel cycle

    SciTech Connect

    Froment, Antoine; Gillet, Philippe

    2007-07-01

    Available in abstract form only. Full text of publication follows: The fast growth of the world's economy coupled with the need for optimizing use of natural resources, for energy security and for climate change mitigation make energy supply one of the 21. century most daring challenges. The high reliability and efficiency of nuclear energy, its competitiveness in an energy market undergoing a new oil shock are as many factors in favor of the 'renaissance' of this greenhouse gas free energy. Over 160,000 tHM of LWR1 and AGR2 Used Nuclear Fuel (UNF) have already been unloaded from the reactor cores corresponding to 7,000 tons discharged per year worldwide. By 2030, this amount could exceed 400,000 tHM and annual unloading 14,000 tHM/year. AREVA believes that closing the nuclear fuel cycle through the treatment and recycling of Used Nuclear Fuel sustains the worldwide nuclear power expansion. It is an economically sound and environmentally responsible choice, based on the preservation of natural resources through the recycling of used fuel. It furthermore provides a safe and secure management of wastes while significantly minimizing the burden left to future generations. (authors)

  17. TMI-2 core damage: a summary of present knowledge

    SciTech Connect

    Owen, D.E.; Mason, R.E.; Meininger, R.D.; Franz, W.A.

    1983-01-01

    Extensive fuel damage (oxidation and fragmentation) has occurred and the top approx. 1.5 m of the center portion of the TMI-2 core has relocated. The fuel fragmentation extends outward to slightly beyond one-half the core radius in the direction examined by the CCTV camera. While the radial extent of core fragmentation in other directions was not directly observed, control and spider drop data and in-core instrument data suggest that the core void is roughly symmetrical, although there are a few indications of severe fuel damage extending to the core periphery. The core material fragmented into a broad range of particle sizes, extending down to a few microns. APSR movement data, the observation of damaged fuel assemblies hanging unsupported from the bottom of the reactor upper plenum structure, and the observation of once-molten stainless steel immediately above the active core indicate high temperatures (up to at least 1720 K) extended to the very top of the core. The relative lack of damage to the underside of the plenum structure implies a sharp temperature demarcation at the core/plenum interface. Filter debris and leadscrew deposit analyses indicate extensive high temperature core materials interaction, melting of the Ag-In-Cd control material, and transport of particulate control material to the plenum and out of the vessel.

  18. Nuclear fuel element

    DOEpatents

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1983-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

  19. NEUTRONIC REACTOR OPERATIONAL METHOD AND CORE SYSTEM

    DOEpatents

    Winters, C.E.; Graham, C.B.; Culver, J.S.; Wilson, R.H.

    1960-07-19

    Homogeneous neutronic reactor systems are described wherein an aqueous fuel solution is continuously circulated through a spherical core tank. The pumped fuel solution-is injected tangentially into the hollow spherical interior, thereby maintaining vigorous rotation of the solution within the tank in the form of a vortex; gaseous radiolytic decomposition products concentrate within the axial vortex cavity. The evolved gas is continuously discharged through a gas- outlet port registering with an extremity of the vortex cavity. and the solution stream is discharged through an annular liquid outlet port concentrically encircling the gas outlet by virtue of which the vortex and its cavity are maintained precisely axially aligned with the gas outlet. A primary heat exchanger extracts useful heat from the hot effluent fuel solution before its recirculation into the core tank. Hollow cylinders and other alternative core- tank configurations defining geometric volumes of revolution about a principal axis are also covered. AEC's Homogeneous Reactor Experiment No. 1 is a preferred embodiment.

  20. Fuels research: Fuel thermal stability overview

    NASA Technical Reports Server (NTRS)

    Cohen, S. M.

    1980-01-01

    Alternative fuels or crude supplies are examined with respect to satisfying aviation fuel needs for the next 50 years. The thermal stability of potential future fuels is discussed and the effects of these characteristics on aircraft fuel systems are examined. Advanced fuel system technology and design guidelines for future fuels with lower thermal stability are reported.

  1. Gaseous fuel nuclear reactor research

    NASA Technical Reports Server (NTRS)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  2. Gaseous fuel nuclear reactor research

    NASA Technical Reports Server (NTRS)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  3. Axial grading of inert matrix fuels

    SciTech Connect

    Recktenwald, G. D.; Deinert, M. R.

    2012-07-01

    Burning actinides in an inert matrix fuel to 750 MWd/kg IHM results in a significant reduction in transuranic isotopes. However, achieving this level of burnup in a standard light water reactor would require residence times that are twice that of uranium dioxide fuels. The reactivity of an inert matrix assembly at the end of life is less than 1/3 of its beginning of life reactivity leading to undesirable radial and axial power peaking in the reactor core. Here we show that axial grading of the inert matrix fuel rods can reduce peaking significantly. Monte Carlo simulations are used to model the assembly level power distributions in both ungraded and graded fuel rods. The results show that an axial grading of uranium dioxide and inert matrix fuels with erbium can reduces power peaking by more than 50% in the axial direction. The reduction in power peaking enables the core to operate at significantly higher power. (authors)

  4. FUEL ELEMENT

    DOEpatents

    Bean, R.W.

    1963-11-19

    A ceramic fuel element for a nuclear reactor that has improved structural stability as well as improved cooling and fission product retention characteristics is presented. The fuel element includes a plurality of stacked hollow ceramic moderator blocks arranged along a tubular raetallic shroud that encloses a series of axially apertured moderator cylinders spaced inwardly of the shroud. A plurality of ceramic nuclear fuel rods are arranged in the annular space between the shroud and cylinders of moderator and appropriate support means and means for directing gas coolant through the annular space are also provided. (AEC)

  5. Fuel ethanol

    SciTech Connect

    Not Available

    1989-02-01

    This report discusses the Omnibus Trade and Competitiveness Act of 1988 which requires GAO to examine fuel ethanol imports from Central America and the Caribbean and their impact on the U.S. fuel ethanol industry. Ethanol is the alcohol in beverages, such as beer, wine, and whiskey. It can also be used as a fuel by blending with gasoline. It can be made from renewable resources, such as corn, wheat, grapes, and sugarcane, through a process of fermentation. This report finds that, given current sugar and gasoline prices, it is not economically feasible for Caribbean ethanol producers to meet the current local feedstock requirement.

  6. POWER GENERATION FROM LIQUID METAL NUCLEAR FUEL

    DOEpatents

    Dwyer, O.E.

    1958-12-23

    A nuclear reactor system is described wherein the reactor is the type using a liquid metal fuel, such as a dispersion of fissile material in bismuth. The reactor is designed ln the form of a closed loop having a core sectlon and heat exchanger sections. The liquid fuel is clrculated through the loop undergoing flssion in the core section to produce heat energy and transferrlng this heat energy to secondary fluids in the heat exchanger sections. The fission in the core may be produced by a separate neutron source or by a selfsustained chain reaction of the liquid fuel present in the core section. Additional auxiliary heat exchangers are used in the system to convert water into steam which drives a turbine.

  7. RECONDITIONING FUEL ELEMENTS

    DOEpatents

    Brandt, H.L.

    1962-02-20

    A process is given for decanning fuel elements that consist of a uranium core, an intermediate section either of bronze, silicon, Al-Si, and uranium silicide layers or of lead, Al-Si, and uranium silicide layers around said core, and an aluminum can bonded to said intermediate section. The aluminum can is dissolved in a solution of sodium hydroxide (9 to 20 wt%) and sodium nitrate (35 to 12 wt %), and the layers of the intermediate section are dissolved in a boiling sodium hydroxide solution of a minimum concentration of 50 wt%. (AEC) A method of selectively reducing plutonium oxides and the rare earth oxides but not uranium oxides is described which comprises placing the oxides in a molten solvent of zinc or cadmium and then adding metallic uranium as a reducing agent. (AEC)

  8. Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

    SciTech Connect

    Heidet, F.; Kim, T.; Grandy, C.

    2012-07-30

    Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium is more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the

  9. Droplet Core Nuclear Rocket (DCNR)

    NASA Technical Reports Server (NTRS)

    Anghaie, Samim

    1991-01-01

    The most basic design feature of the droplet core nuclear reactor is to spray liquid uranium into the core in the form of droplets on the order of five to ten microns in size, to bring the reactor to critical conditions. The liquid uranium fuel ejector is driven by hydrogen, and more hydrogen is injected from the side of the reactor to about one and a half meters from the top. High temperature hydrogen is expanded through a nozzle to produce thrust. The hydrogen pressure in the system can be somewhere between 50 and 500 atmospheres; the higher pressure is more desirable. In the lower core region, hydrogen is tangentially injected to serve two purposes: (1) to provide a swirling flow to protect the wall from impingement of hot uranium droplets: (2) to generate a vortex flow that can be used for fuel separation. The reactor is designed to maximize the energy generation in the upper region of the core. The system can result in and Isp of 2000 per second, and a thrust-to-weight ratio of 1.6 for the shielded reactor. The nuclear engine system can reduce the Mars mission duration to less than 200 days. It can reduce the hydrogen consumption by a factor of 2 to 3, which reduces the hydrogen load by about 130 to 150 metric tons.

  10. Fuel composition

    SciTech Connect

    Badger, S.L.

    1983-09-20

    A composition useful, inter alia, as a fuel, is based on ethyl alcohol denatured with methylisobutyl alcohol and kerosene, which is mixed with xylenes and isopropyl alcohol. The xylenes and isopropyl alcohol act with the denaturizing agents to raise the flash point above that of ethyl alcohol alone and also to mask the odor and color the flame, thus making the composition safer for use as a charcoal lighter or as a fuel for e.g. patio lamps.

  11. PRIZMA predictions of in-core detection indications in the VVER-1000 reactor

    NASA Astrophysics Data System (ADS)

    Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.

    2014-06-01

    The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.

  12. Wire core reactor for nuclear thermal propulsion

    NASA Astrophysics Data System (ADS)

    Harty, Richard B.; Brengle, Robert G.

    1993-01-01

    Studies have been performed of a compact high-performance nuclear rocket reactor that incorporates a tungsten alloy wire fuel element. This reactor, termed the wire core reactor, can deliver a specific impulse of 1,000 s using an expander cycle and a nozzle expansion ratio of 500 to 1. The core is constructed of layers of 0.8-mm-dia fueled tungsten wires wound over alternate layers of spacer wires, which forms a rugged annular lattice. Hydrogen flow in the core is annular, flowing from inside to outside. In addition to the concepts compact size and good heat transfer, the core has excellent power-flow matching features and can resist vibration and thermal stresses during star-up and shutdown.

  13. MODULAR CORE UNITS FOR A NEUTRONIC REACTOR

    DOEpatents

    Gage, J.F. Jr.; Sherer, D.B.

    1964-04-01

    A modular core unit for use in a nuclear reactor is described. Many identical core modules can be placed next to each other to make up a complete core. Such a module includes a cylinder of moderator material surrounding a fuel- containing re-entrant coolant channel. The re-entrant channel provides for the circulation of coolant such as liquid sodium from one end of the core unit, through the fuel region, and back out through the same end as it entered. Thermal insulation surrounds the moderator exterior wall inducing heat to travel inwardly to the coolant channel. Spaces between units may be used to accommodate control rods and support structure, which may be cooled by a secondary gas coolant, independently of the main coolant. (AEC)

  14. Core/Shell semiconductor nanocrystals.

    PubMed

    Reiss, Peter; Protière, Myriam; Li, Liang

    2009-02-01

    Colloidal core/shell nanocrystals contain at least two semiconductor materials in an onionlike structure. The possibility to tune the basic optical properties of the core nanocrystals, for example, their fluorescence wavelength, quantum yield, and lifetime, by growing an epitaxial-type shell of another semiconductor has fueled significant progress on the chemical synthesis of these systems. In such core/shell nanocrystals, the shell provides a physical barrier between the optically active core and the surrounding medium, thus making the nanocrystals less sensitive to environmental changes, surface chemistry, and photo-oxidation. The shell further provides an efficient passivation of the surface trap states, giving rise to a strongly enhanced fluorescence quantum yield. This effect is a fundamental prerequisite for the use of nanocrystals in applications such as biological labeling and light-emitting devices, which rely on their emission properties. Focusing on recent advances, this Review discusses the fundamental properties and synthesis methods of core/shell and core/multiple shell structures of II-VI, IV-VI, and III-V semiconductors.

  15. Characterization plan for Fort St. Vrain and Peach Bottom graphite fuels

    SciTech Connect

    Maarschman, S.C.; Berting, F.M.; Clemmer, R.G.; Gilbert, E.R.; Guenther, R.J.; Morgan, W.C.; Sliva, P.

    1993-09-01

    Part of Fort St. Vrain (FSV) and most of the Peach Bottom (PB) reactor spent fuels are currently stored at INEL and may remain in storage for many years before disposal. Three disposal pathways have been proposed: intact disposal, fuels partially disassembled and the high-level waste fraction conditioned prior to disposal, and fuels completed disassembled and conditioned prior to disposal. Many options exist within each of these pathways. PNL evaluated the literature and other reference to develop a fuels characterization plan for these fuels. This plan provides guidance for the characteristics of the fuel which will be needed to pursue any of the storage or disposal pathways. It also provides a suggested fuels monitoring program for the current storage facilities. This report recommends a minimum of 7 fuel elements be characterized: PB Core 1 fuel: one Type II nonfailed element, one Type II failed element, and one Type III nonfailed element; PB Core 2 fuel: two Type II nonfailed fuel elements; and FSV fuel: at least two fuel blocks from regions of high temperature and fluence and long in-reactor performance (preferably at reactor end-of- life). Selection of PB fuel elements should focus on these between radial core position 8 and 14 and on compacts between compact numbers 10 and 20. Selection of FSV fuel elements should focus on these from Fuel Zones II and III, located in Core Layers 6, 7, and possibly 8.

  16. Heat exchanger for fuel cell power plant reformer

    DOEpatents

    Misage, Robert; Scheffler, Glenn W.; Setzer, Herbert J.; Margiott, Paul R.; Parenti, Jr., Edmund K.

    1988-01-01

    A heat exchanger uses the heat from processed fuel gas from a reformer for a fuel cell to superheat steam, to preheat raw fuel prior to entering the reformer and to heat a water-steam coolant mixture from the fuel cells. The processed fuel gas temperature is thus lowered to a level useful in the fuel cell reaction. The four temperature adjustments are accomplished in a single heat exchanger with only three heat transfer cores. The heat exchanger is preheated by circulating coolant and purge steam from the power section during startup of the latter.

  17. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    SciTech Connect

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-12-31

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR`s) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design.

  18. Stationary Liquid Fuel Fast Reactor

    SciTech Connect

    Yang, Won Sik; Grandy, Andrew; Boroski, Andrew; Krajtl, Lubomir; Johnson, Terry

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  19. Analyses of eigenvalue bias and control rod worths in FFTF (Fast Flux Test Facility)

    SciTech Connect

    Nelson, J.V.; Dobbin, K.D.; Wootan, D.W.; Campbell, L.R.

    1987-01-01

    The Fast Flux Test Facility (FFTF) core loading during its ninth operating cycle was significantly different from that of previous cycles because of the presence of the Core Demonstration Experiment (CDE). The CDE consists of a number of axially blanketed fuel assemblies and internal blankets prototypic of advanced oxide cores in Liquid Metal Reactors (LMR). In preparation for the Cycle 9 reload design effort, a careful assessment of control rod worth and reactivity calculations for Cycles 1 through 8 was made. The goal of this study was to establish calculational biases and reduce uncertainties factored into the reload design calculations. These analyses helped assure that the operational objectives for Cycle 9 were met.

  20. Seismic analysis of freestanding fuel racks

    SciTech Connect

    Gilmore, C.B.

    1982-01-01

    This paper presents a nonlinear transient dynamic time-history analysis of freestanding spent fuel storage racks subjected to seismic excitation. This type of storage rack is structurally unrestrained and submerged in water in the spent fuel pool of a nuclear power complex, holds (spent) fuel assemblies which have been removed from the reactor core. Nonlinearities in the fuel rack system include impact between the fuel assembly and surrounding cell due to clearances between them, friction due to sliding between the fuel rack support structure and spent fuel pool floor, and the lift-off of the fuel rack support structure from the spent fuel pool floor. The analysis of the fuel rack system includes impacting due to gap closures, energy losses due to impacting bodies, Coulomb damping between sliding surfaces, and hydrodynamic mass effects. Acceleration time history excitation development is discussed. Modeling considerations, such as the initial status of nonlinear elements, number of mode shapes to include in the analysis, modal damping, and integration time-step size are presented. The response of the fuel rack subjected to two-dimensional seismic excitation is analyzed by the modal superposition method, which has resulted in significant computer cost savings when compared to that of direct integration.

  1. COMPARISON OF METHODS TO DETERMINE OXYGEN DEMAND FOR BIOREMEDIATION OF A FUEL CONTAMINATED AQUIFER

    EPA Science Inventory

    Four analytical methods were compared for estimating concentrations of fuel contaminants in subsurface core samples. The methods were total organic carbon, chemical oxygen demand, oil and grease, and a solvent extraction of fuel hydrocarbons combined with a gas chromatographic te...

  2. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  3. Twilight reloaded: the peptide experience

    PubMed Central

    Weichenberger, Christian X.; Pozharski, Edwin; Rupp, Bernhard

    2017-01-01

    The de facto commoditization of biomolecular crystallography as a result of almost disruptive instrumentation automation and continuing improvement of software allows any sensibly trained structural biologist to conduct crystallo­graphic studies of biomolecules with reasonably valid outcomes: that is, models based on properly interpreted electron density. Robust validation has led to major mistakes in the protein part of structure models becoming rare, but some depositions of protein–peptide complex structure models, which generally carry significant interest to the scientific community, still contain erroneous models of the bound peptide ligand. Here, the protein small-molecule ligand validation tool Twilight is updated to include peptide ligands. (i) The primary technical reasons and potential human factors leading to problems in ligand structure models are presented; (ii) a new method used to score peptide-ligand models is presented; (iii) a few instructive and specific examples, including an electron-density-based analysis of peptide-ligand structures that do not contain any ligands, are discussed in detail; (iv) means to avoid such mistakes and the implications for database integrity are discussed and (v) some suggestions as to how journal editors could help to expunge errors from the Protein Data Bank are provided. PMID:28291756

  4. Diuretics in primary hypertension - Reloaded.

    PubMed

    Mishra, Sundeep

    Diuretics have long been cherished as drugs of choice for uncomplicated primary hypertension. Robust mortality and morbidity data is available for diuretics to back this strategy. Off-late the interest for diuretics has waned off perhaps due to availability of more effective drugs but more likely due to perceived lack of tolerance and side-effect profile of high-dose of diuretics required for mortality benefit. Low-dose diuretics particularly thiazide diuretics are safer but lack the mortality benefit shown by high-dose. However, indapamide and low dose chlorthalidone have fewer side-effects but continue to provide mortality benefit. Copyright © 2016. Published by Elsevier B.V.

  5. Composite Reliability Enhancement Via Reloading

    DTIC Science & Technology

    1988-09-01

    the design of composite structures , which in turn adds weight and size and causes other related problems that reduce design efficiency. As the...such large structures . This is due to the the lower weak tail of the strength distributions of the con- stituent fibers. This effect has been...tion was run on an IBM Personal Computer using Microsoft Fortran 4.01 for source code and Lotus 1-2-3 for graphing. When the simula- tion program had

  6. Alternative jet aircraft fuels

    NASA Technical Reports Server (NTRS)

    Grobman, J.

    1979-01-01

    Potential changes in jet aircraft fuel specifications due to shifts in supply and quality of refinery feedstocks are discussed with emphasis on the effects these changes would have on the performance and durability of aircraft engines and fuel systems. Combustion characteristics, fuel thermal stability, and fuel pumpability at low temperature are among the factors considered. Combustor and fuel system technology needs for broad specification fuels are reviewed including prevention of fuel system fouling and fuel system technology for fuels with higher freezing points.

  7. Alternative jet aircraft fuels

    NASA Technical Reports Server (NTRS)

    Grobman, J.

    1979-01-01

    Potential changes in jet aircraft fuel specifications due to shifts in supply and quality of refinery feedstocks are discussed with emphasis on the effects these changes would have on the performance and durability of aircraft engines and fuel systems. Combustion characteristics, fuel thermal stability, and fuel pumpability at low temperature are among the factors considered. Combustor and fuel system technology needs for broad specification fuels are reviewed including prevention of fuel system fouling and fuel system technology for fuels with higher freezing points.

  8. Nuclear fuel elements having a composite cladding

    DOEpatents

    Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

    1983-09-20

    An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

  9. Energy efficient engine core design and performance report. Report, January 1978-December 1982

    SciTech Connect

    Stearns, E.M.

    1982-12-01

    The Energy Efficient Engine (E3) is a NASA program to develop fuel saving technology for future large transport aircraft engines. Testing of the General Electric E3 core showed that the core component performance and core system performance necessary to meet the program goals can be achieved. The E3 core design and test results are described.

  10. Reforming of fuel inside fuel cell generator

    DOEpatents

    Grimble, Ralph E.

    1988-01-01

    Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream I and spent fuel stream II. Spent fuel stream I is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream I and exhaust stream II, and exhaust stream I is vented. Exhaust stream II is mixed with spent fuel stream II to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells.

  11. Reforming of fuel inside fuel cell generator

    DOEpatents

    Grimble, R.E.

    1988-03-08

    Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream 1 and spent fuel stream 2. Spent fuel stream 1 is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream 1 and exhaust stream 2, and exhaust stream 1 is vented. Exhaust stream 2 is mixed with spent fuel stream 2 to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells. 1 fig.

  12. FUEL ELEMENT

    DOEpatents

    Fortescue, P.; Zumwalt, L.R.

    1961-11-28

    A fuel element was developed for a gas cooled nuclear reactor. The element is constructed in the form of a compacted fuel slug including carbides of fissionable material in some cases with a breeder material carbide and a moderator which slug is disposed in a canning jacket of relatively impermeable moderator material. Such canned fuel slugs are disposed in an elongated shell of moderator having greater gas permeability than the canning material wherefore application of reduced pressure to the space therebetween causes gas diffusing through the exterior shell to sweep fission products from the system. Integral fission product traps and/or exterior traps as well as a fission product monitoring system may be employed therewith. (AEC)

  13. Fuel bundle

    SciTech Connect

    Lui, C.K.

    1989-04-04

    This patent describes a method of forming a fuel bundle of a nuclear reactor. The method consists of positioning the fuel rods in the bottom plate, positioning the tie rod in the bottom plate with the key passed through the receptacle to the underside of the bottom plate and, after the tie rod is so positioned, turning the tie rod so that the key is in engagement with the underside of the bottom plate. Thereafter mounting the top plate is mounted in engagement with the fuel rods with the upper end of the tie rod extending through the opening in the top plate and extending above the top plate, and the tie rod is secured to the upper side of sid top plate thus simultaneously securing the key to the underside of the bottom plate.

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

    SciTech Connect

    Richard, R.F.

    1997-10-28

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

  15. FUEL ELEMENT

    DOEpatents

    Howard, R.C.; Bokros, J.C.

    1962-03-01

    A fueled matrlx eontnwinlng uncomblned carbon is deslgned for use in graphlte-moderated gas-cooled reactors designed for operatlon at temperatures (about 1500 deg F) at which conventional metallic cladding would ordlnarily undergo undesired carburization or physical degeneratlon. - The invention comprlses, broadly a fuel body containlng uncombined earbon, clad with a nickel alloy contalning over about 28 percent by' weight copper in the preferred embodlment. Thls element ls supporirted in the passageways in close tolerance with the walls of unclad graphite moderator materlal. (AEC)

  16. Dosimetry Evaluation of In-Core and Above-Core Zirconium Alloy Samples in a PWR

    NASA Astrophysics Data System (ADS)

    Amiri, Benjamin W.; Foster, John P.; Greenwood, Larry R.

    2016-02-01

    A description of the neutron fluence analysis of activated zirconium alloys samples at a Westinghouse 3-loop reactor is presented. These samples were irradiated in the core and in the fuel plenum region, where dosimetry measurements are relatively rare compared with regions radially outward of the core. Dosimetry measurements performed by Batelle/PNNL are compared to the calculational models. Good agreement is shown with the in-core measurements when using analysis conditions expected to best represent this region, such as an assembly-specific axial power distribution. However, the use of these conditions to evaluate dosimetry in the fuel plenum region can lead to significant underestimation of the fluence. The use of a flat axial power distribution, however, does not underestimate the fluence in the fuel plenum region.

  17. LWR Core Materials Program: progress in 1983-1984

    SciTech Connect

    Franklin, D.G.; Gehl, S.M.; Machiels, A.J.; Santucci, J.

    1985-10-01

    The years 1983 and 1984 marked the completion of the power reactor extended-burnup demonstrations, several multinational Ramp programs, the prerelease version of the ESCORE fuel performance code, and the BWR channel extended-life program. Research results gathered under the overall program showed an improvement in core component reliability and an increase in nuclear fuel burnup.

  18. First Core and Refueling Options for IRIS

    SciTech Connect

    Petrovic, Bojan; Carelli, Mario D.; Greenspan, Ehud; Milosevic, Miodrag; Vujic, Jasmina; Padovani, Enrico; Ganda, Francesco

    2002-07-01

    The International Reactor Innovative and Secure (IRIS) is being developed by an international consortium of industry, laboratory, university and utility establishments, led by Westinghouse. The IRIS design addresses key requirements associated with advanced reactors, including improved safety, enhanced proliferation resistance, competitive electricity production cost, and improved waste management. IRIS is a modular, small/medium size (100 to 335 MWe) PWR with integral vessel configuration. Its design is based on proven LWR technology, so that no new technology development is needed and near term deployment is possible. At the same time, aim was to introduce improvements as compared to present PWRs. These opposing requirements resulted in an evolutionary approach to fuel and core design, balancing new features and the need to avoid extensive testing and demonstration programs. A path forward was devised by selecting the current fuel technology for the first IRIS core, but keeping future upgrades possible through the variable moderation fuel assembly design. This paper describes this approach and discusses core fueling options that enable achieving four-year and eight-year core lifetime. (authors)

  19. Fuels characterization studies. [jet fuels

    NASA Technical Reports Server (NTRS)

    Seng, G. T.; Antoine, A. C.; Flores, F. J.

    1980-01-01

    Current analytical techniques used in the characterization of broadened properties fuels are briefly described. Included are liquid chromatography, gas chromatography, and nuclear magnetic resonance spectroscopy. High performance liquid chromatographic ground-type methods development is being approached from several directions, including aromatic fraction standards development and the elimination of standards through removal or partial removal of the alkene and aromatic fractions or through the use of whole fuel refractive index values. More sensitive methods for alkene determinations using an ultraviolet-visible detector are also being pursued. Some of the more successful gas chromatographic physical property determinations for petroleum derived fuels are the distillation curve (simulated distillation), heat of combustion, hydrogen content, API gravity, viscosity, flash point, and (to a lesser extent) freezing point.

  20. PWR cores with silicon carbide cladding

    SciTech Connect

    Dobisesky, J. P.; Carpenter, D.; Pilat, E.; Kazimi, M. S.

    2012-07-01

    The feasibility of using silicon carbide rather than Zircaloy cladding, to reach higher power levels and higher discharge burnups in PWRs has been evaluated. A preliminary fuel design using fuel rods with the same dimensions as in the Westinghouse Robust Fuel Assembly but with fuel pellets having 10 vol% central void has been adopted to mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide and to the persistence of the open clad-pellet gap over most of the fuel life. With this modified fuel design, it is possible to achieve 18 month cycles that meet present-day operating constraints on peaking factor, boron concentration, reactivity coefficients and shutdown margin, while allowing batch average discharge burnups up to 80 MWD/kgU and peak rod burnups up to 100 MWD/kgU. Power uprates of 10% and possibly 20% also appear feasible. For non-uprated cores, the silicon carbide-clad fuel has a clear advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a savings in enriched uranium. Control rod configuration modifications may be required to meet the shutdown margin criterion for the 20% up-rate. Silicon carbide's ability to sustain higher burnups than Zircaloy also allows the design of a licensable two year cycle with only 96 fresh assemblies, avoiding the enriched uranium penalty incurred with use of larger batch sizes due to their excessive leakage. (authors)

  1. Future Fuel.

    ERIC Educational Resources Information Center

    Stover, Del

    1991-01-01

    Tough new environmental laws, coupled with fluctuating oil prices, are likely to prompt hundreds of school systems to examine alternative fuels. Literature reviews and interviews with 45 government, education, and industry officials provided data for a comparative analysis of gasoline, diesel, natural gas, methanol, and propane. (MLF)

  2. Future Fuel.

    ERIC Educational Resources Information Center

    Stover, Del

    1991-01-01

    Tough new environmental laws, coupled with fluctuating oil prices, are likely to prompt hundreds of school systems to examine alternative fuels. Literature reviews and interviews with 45 government, education, and industry officials provided data for a comparative analysis of gasoline, diesel, natural gas, methanol, and propane. (MLF)

  3. Nuclear Fuels.

    ERIC Educational Resources Information Center

    Nash, J. Thomas

    1983-01-01

    Trends in and factors related to the nuclear industry and nuclear fuel production are discussed. Topics addressed include nuclear reactors, survival of the U.S. uranium industry, production costs, budget cuts by the Department of Energy and U.S. Geological survey for resource studies, mining, and research/development activities. (JN)

  4. Nuclear Fuels.

    ERIC Educational Resources Information Center

    Nash, J. Thomas

    1983-01-01

    Trends in and factors related to the nuclear industry and nuclear fuel production are discussed. Topics addressed include nuclear reactors, survival of the U.S. uranium industry, production costs, budget cuts by the Department of Energy and U.S. Geological survey for resource studies, mining, and research/development activities. (JN)

  5. Fuel Cells

    ERIC Educational Resources Information Center

    Hawkins, M. D.

    1973-01-01

    Discusses the theories, construction, operation, types, and advantages of fuel cells developed by the American space programs. Indicates that the cell is an ideal small-scale power source characterized by its compactness, high efficiency, reliability, and freedom from polluting fumes. (CC)

  6. Fuel Cells

    ERIC Educational Resources Information Center

    Hawkins, M. D.

    1973-01-01

    Discusses the theories, construction, operation, types, and advantages of fuel cells developed by the American space programs. Indicates that the cell is an ideal small-scale power source characterized by its compactness, high efficiency, reliability, and freedom from polluting fumes. (CC)

  7. Pebble-bed core design option for VHTRs - Core configuration flexibility and potential applications

    SciTech Connect

    Pritchard, M. L.; Tsvetkov, P. V.

    2006-07-01

    Gas-cooled nuclear reactors have been receiving specific attention for Generation IV possibilities due to desired characteristics such as relatively low cost, short construction period, and inherent safety. Attractive inherent characteristics include an inert, single phase helium coolant, refractory coated fuel with high temperature capability and low fission product release, and graphite moderator with high temperature stability and long response times. The passively safe design has a relatively low power density, annular core, large negative temperature coefficient, and passive decay heat removal system. The objective of the U.S. DOE NERI Project is to assess the possibility, advantages and limitations of VHTRs with fuel loadings containing minor actinides. This paper presents the analysis of pebble-bed core configurations. Whole-core 3D models for pebble-bed design with multi-heterogeneity treatments in SCALE 5.0 are developed to compare computational results with experiments. Obtained results are in agreement with the available HTR-10 data. Actinide fueled VHTR configurations reveal promising performance. With an optimized pebble-bed model, the spectrum shifting abilities become more apparent. Effects of altered moderator to fuel ratio, Dancoff factor, and core and reflector configurations are investigated. This effort is anticipated to contribute to a facilitated development of new fuel cycles in support of future operation of Generation IV nuclear energy systems. (authors)

  8. Transversal Stiffness and Beta-Actin and Alpha-Actinin-4 Content of the M. Soleus Fibers in the Conditions of a 3-Day Reloading after 14-Day Gravitational Unloading

    PubMed Central

    Ogneva, I. V.

    2011-01-01

    The aim of the work was to analyze the structural changes in different parts of the sarcolemma and contractile apparatus of muscle fibers by measuring their transversal stiffness by atomic force microscopy in a three-day reloading after a 14-day gravity disuse, which was carried out by hind-limbs suspension. The object of the study was the soleus muscle of the Wistar rat. It was shown that after 14 days of disuse, there was a reduction of transversal stiffness of all points of the sarcolemma and contractile apparatus. Readaptation for 3 days leads to complete recovery of the values of the transversal stiffness of the sarcolemma and to partial value recovery of the contractile apparatus. The changes in transversal stiffness of sarcolemma correlate with beta-actin and alpha-actinin-4 in membrane protein fractions. PMID:21941432

  9. Thermionic fuel element technology status

    NASA Technical Reports Server (NTRS)

    Holland, J. W.; Horner, M. W.; Yang, L.

    1985-01-01

    The results of research, conducted between the mid-1960s and 1973, on the multiconverter thermionic fuel elements (TFEs) that comprise the reactor core of an SP-100 thermionic reactor system are presented. Fueled-emitter technology, insulator technology and cell and TFE assembly technology of the prototypical TFEs which were tested in-pile and out-of-pile during these years are described. The proto-TFEs have demonstrated reproducible performance within 5 percent and no premature failures within the 1.5 yr of operation (with projected 3-yr lifetimes). The two primary life-limiting factors had been identified as thermionic emitter dimensional increase due to interactions with the fuel and electrical insulator structural damage from fast neutrons. Multiple options for extending TFE lifetimes to 7 yr or longer are available and will be investigated in the 1984-1985 SP-100 program for resolution of critical technology issues. Design diagrams and test graphs are included.

  10. Full Core 3-D Simulation of a Partial MOX LWR Core

    SciTech Connect

    S. Bays; W. Skerjanc; M. Pope

    2009-05-01

    A comparative analysis and comparison of results obtained between 2-D lattice calculations and 3-D full core nodal calculations, in the frame of MOX fuel design, was conducted. This study revealed a set of advantages and disadvantages, with respect to each method, which can be used to guide the level of accuracy desired for future fuel and fuel cycle calculations. For the purpose of isotopic generation for fuel cycle analyses, the approach of using a 2-D lattice code (i.e., fuel assembly in infinite lattice) gave reasonable predictions of uranium and plutonium isotope concentrations at the predicted 3-D core simulation batch average discharge burnup. However, it was found that the 2-D lattice calculation can under-predict the power of pins located along a shared edge between MOX and UO2 by as much as 20%. In this analysis, this error did not occur in the peak pin. However, this was a coincidence and does not rule out the possibility that the peak pin could occur in a lattice position with high calculation uncertainty in future un-optimized studies. Another important consideration in realistic fuel design is the prediction of the peak axial burnup and neutron fluence. The use of 3-D core simulation gave peak burnup conditions, at the pellet level, to be approximately 1.4 times greater than what can be predicted using back-of-the-envelope assumptions of average specific power and irradiation time.

  11. Pellet fueling technology development leading to efficient fueling of ITER burning plasmas

    SciTech Connect

    Baylor, L.R.; Combs, S.K.; Jernigan, T.C.; Houlberg, W.A.; Owen, L.W.; Rasmussen, D.A.; Maruyama, S.; Parks, P.B.

    2005-05-15

    Pellet injection is the primary fueling technique planned for core fueling of ITER [ITER Technical Basis 2002 ITER EDA Documentation Series (Vienna: IAEA)] burning plasmas. Efficient core plasma fueling with deuterium and tritium D-T is a requirement for achieving high fusion gain and it cannot be achieved with gas fueling. Injection of pellets from the inner wall has been shown on present day tokamaks to provide efficient fueling and is planned for use on ITER. Modeling of the fueling deposition from inner wall pellet injection using the Parks ExB drift model indicates that pellets have the capability to fuel well inside the separatrix. Gas fueling calculations show very poor neutral penetration due to the high density and wide scrape off layer. Isotopically mixed D-T pellets can provide efficient tritium fueling that will minimize tritium wall loading when compared to gas puffing. Currently the performance of the ITER inner wall guide tube design is under test with initial results indicating that pellet speeds in excess of 300 m/s will lead to fragmented pellets. The ITER pellet injection technology requirements and remaining development issues are discussed along with a plan to reach the design goal for employment on ITER.

  12. The effects of core geometry on core power for passively safe HTGRs

    SciTech Connect

    Kubali, V.C.; Lidsky, L.M. )

    1993-01-01

    Decay heat removal under accident conditions is one of the major concerns of nuclear reactor safety. By imposing a limit on the core power density, reactor designers can ensure that the energy generated after shutdown would not pose any problem even when the active cooling agent is lost. On the other hand, it is usually desirable to have a high power density for the competitiveness of the reactor design. Therefore, the design approach must make maximum use of passive means of heat transfer for obtaining the highest achievable core power density without sacrificing passive safety. The strong temperature dependence of radiative heat transfer combined with the high-temperature operability of coated-particle-fueled gas-cooled reactors can provide a valuable opportunity in this regard. To enhance radiative heat transfer within the core, an optimization of the design parameters, such as core geometry, fuel element configuration, and a proper choice of the materials, would be necessary in addition to neutronics considerations.

  13. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    NASA Astrophysics Data System (ADS)

    Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

    2016-05-01

    The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data.

  14. Information Is a Common Core Dish Best Served First

    ERIC Educational Resources Information Center

    Walkowiak, Temple A.

    2015-01-01

    There has been a spike in negative comments about mathematics and standards since the implementation of the Common Core State Standards. Regardless of whether the comments are fueled by social media or traditional media, educators need to be armed with strategies for helping parents understand, navigate, and embrace the Common Core's mathematics…

  15. Information Is a Common Core Dish Best Served First

    ERIC Educational Resources Information Center

    Walkowiak, Temple A.

    2015-01-01

    There has been a spike in negative comments about mathematics and standards since the implementation of the Common Core State Standards. Regardless of whether the comments are fueled by social media or traditional media, educators need to be armed with strategies for helping parents understand, navigate, and embrace the Common Core's mathematics…

  16. HTR-Proteus Pebble Bed Experimental Program Cores 5,6,7,&8: Columnar Hexagonal Point-on-Point Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    SciTech Connect

    Bess, John D.; Sterbentz, James W.; Snoj, Luka; Lengar, Igor; Koberl, Oliver

    2015-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  17. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

    SciTech Connect

    John D. Bess

    2013-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  18. Critical Design Parameters for Pylon-Aided Gaseous Fuel Injection Upstream of a Flameholding Cavity

    DTIC Science & Technology

    2009-03-01

    cavity depth (in) FPP flammable plume percentage h pylon height (in) penetration of plume core (in) total fuel plume penetration (in) l pylon...metrics include: mixing efficiency (η), fuel plume area (Ap), flammable fuel plume area (Af) , flammable fuel plume percentage ( FPP ), species...combustion. Another key metric is the flammable plume percentage ( FPP ), as seen in Equation 10. This metric represents the percentage of the fuel

  19. Analyses of Greek Research Reactor with mixed HEU-LEU Be reflected core

    SciTech Connect

    Deen, J.R.; Snelgrove, J.L.; Papastergiou, K.

    1993-12-31

    The fuel-cycle analyses presented in this paper provide specific steps to be taken in the transition from a 36-element water-reflected HEU core to a 33-element LEU equilibrium core with a Be reflector on two faces. The first step will be to install the Be reflector and remove the highest burnup HEU fuel. The smaller Be-reflected core will be refueled with LEU fuel. All analyses were performed using a planar 5-group REBUS3 model benchmarked to VIM Monte Carlo. In addition to fuel cycle results, the control rod worth, reactivity response to increased fuel and water temperature and decreased water density were compared for the transition core and the reference HEU core.

  20. Sodium fast reactor evaluation: Core materials

    NASA Astrophysics Data System (ADS)

    Cheon, Jin Sik; Lee, Chan Bock; Lee, Byoung Oon; Raison, J. P.; Mizuno, T.; Delage, F.; Carmack, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.

  1. Fuel cell

    SciTech Connect

    Struthers, R.C.

    1983-06-28

    An improved fuel cell comprising an anode section including an anode terminal, an anode fuel, and an anolyte electrolyte, a cathode section including a cathode terminal, an electron distributor and a catholyte electrolyte, an ion exchange section between the anode and cathode sections and including an ionolyte electrolyte, ion transfer membranes separating the ionolyte from the anolyte and the catholyte and an electric circuit connected with and between the terminals conducting free electrons from the anode section and delivering free electrons to the cathode section, said ionolyte receives ions of one polarity moving from the anolyte through the membrane related thereto preventing chemical equilibrium in the anode section and sustaining chemical reaction and the generating of free electrons therein, said ions received by the ionolyte from the anolyte release different ions from the ionolyte which move through the membrane between the ionolyte and catholyte and which add to the catholyte.

  2. The effect of fuel rod oxidation on PCMI-induced fuel failure

    NASA Astrophysics Data System (ADS)

    Kim, Kyu-Tae

    2011-11-01

    It was found in a one of the PWRs operating in Korea that a few three cycle-burned Zry-4 fuel assemblies which were loaded in a core center region at control bank positions were leaking. The leaking cycle has experienced a few reactor trips and some fuel rods started to leak at about a month after a power ramp following the second reactor trip. To investigate a root cause of such fuel failure as well as to examine intact and leak rod oxidation behaviors, one intact and one leaking fuel rods were selected from one intact and one failed three cycle-burned fuel assemblies, respectively, and in parallel one intact fuel rod was selected from a two cycle-burned fuel assembly to examine the effect of burnup on fuel rod oxidation and cladding stress during the power ramp. The maximum oxide thicknesses for the intact two cycle-burned and three cycle-burned fuel rods were measured to be about 70 and 140 μm, respectively, whereas that for the leaking three cycle-burned fuel rod to be about 200 μm. The leaking fuel rods generated a very sharp increase in oxide thickness in the fuel rod upper region having a relatively high axial power, resulting in through-wall axial cracks. The root cause of the fuel rod leaks was evaluated to be the pellet-clad mechanical interaction (PCMI)-induced failure combined with excessive Zry-4 oxidation and cladding stress, based on the evaluations of pellet-clad friction coefficient-dependent cladding hoop stresses after the power ramp following the second trip, measured oxide thicknesses and axial cracks on the cladding surface, a fuel leak initiation time and failed fuel rod locations at the control bank positions.

  3. Critical experiments with mixed oxide fuel

    SciTech Connect

    Harris, D.R.

    1997-06-01

    This paper very briefly outlines technical considerations in performing critical experiments on weapons-grade plutonium mixed oxide fuel assemblies. The experiments proposed would use weapons-grade plutonium and Er{sub 2}O{sub 3} at various dissolved boron levels, and for specific fuel assemblies such as the ABBCE fuel assembly with five large water holes. Technical considerations described include the core, the measurements, safety, security, radiological matters, and licensing. It is concluded that the experiments are feasible at the Rensselaer Polytechnic Institute Reactor Critical Facility. 9 refs.

  4. Electrometallurgical treatment of TMI-2 fuel debris

    SciTech Connect

    Karell, E.J.; Gourishankar, K.V.; Johnson, G.K.

    1997-08-01

    Argonne National Laboratory (ANL) has developed an electrometallurgical treatment process suitable for conditioning DOE oxide spent fuel for long-term storage or disposal. The process consists of an initial oxide reduction step that converts the actinide oxides to a metallic form, followed by an electrochemical separation of uranium from the other fuel constituents. The final product of the process is a uniform set of stable waste forms suitable for long-term storage or disposal. The suitability of the process for treating core debris from the Three Mile Island-2 (TMI-2) reactor is being evaluated. This paper reviews the results of preliminary experimental work performed using simulated TMI-2 fuel debris.

  5. Fuel conditioner

    SciTech Connect

    Nelson, M.L.; Nelson, O.L. Jr.

    1988-06-28

    A fuel conditioner is described comprising 10 to 80% of a polar oxygenated hydrocarbon having an average molecular weight from about 250 to about 500, an acid acid number from about 25 to about 125, and a saponification number from about 30 to about 250; and 5 to 50% of an oxygenated compatibilizing agent having a solubility parameter of from about 8.8 to about 11.5 and moderate to strong hydrogen-bonding capacity.

  6. Fuel additives

    SciTech Connect

    Gheysens, J.L.G.

    1990-11-27

    This patent describes a composition for the improvement of hydrocarbon fuels exhibiting a boiling range of gasoline being suitable for use in spark ignition-type engines. It comprises an aromatic amine; a polyaminated detergent; a catalyst comprising a colloidal suspension or amine salt of transition/alkali/alkaline earth metal organic coordinations having at least one metal oxidehydroxide linked to an alkyl chain via a carboxyl group; and a solvent comprising an alkanol-aliphatic ether oxygenated hydrocarbon.

  7. Alcohol fuels

    SciTech Connect

    Not Available

    1990-07-01

    Ethanol is an alcohol made from grain that can be blended with gasoline to extend petroleum supplies and to increase gasoline octane levels. Congressional proposals to encourage greater use of alternative fuels could increase the demand for ethanol. This report evaluates the growth potential of the ethanol industry to meet future demand increases and the impacts increased production would have on American agriculture and the federal budget. It is found that ethanol production could double or triple in the next eight years, and that American farmers could provide the corn for this production increase. While corn growers would benefit, other agricultural segments would not; soybean producers, for example could suffer for increased corn oil production (an ethanol byproduct) and cattle ranchers would be faced with higher feed costs because of higher corn prices. Poultry farmers might benefit from lower priced feed. Overall, net farm cash income should increase, and consumers would see slightly higher food prices. Federal budget impacts would include a reduction in federal farm program outlays by an annual average of between $930 million (for double current production of ethanol) to $1.421 billion (for triple production) during the eight-year growth period. However, due to an partial tax exemption for ethanol blended fuels, federal fuel tax revenues could decrease by between $442 million and $813 million.

  8. Experimental evaluation of new NEU cores in the UVAR

    SciTech Connect

    Farrar, P.; Hosticka, B.; Krause, D.; Mulder, R.; Rydin, R.

    1997-08-01

    The University of Virginia began working on converting the UVAR reactor to LEU fuel in the Spring of 1986. The Safety Analysis Report was completed and submitted to the NRC in late 1989. After review, the DOE order to manufacture LEU fuel was placed at B&W in March 1992, and the new fuel was received in January 1994. The 4-by-4 fully-graphite-reflected LEU-1 core went critical on April 20, 1994, and the 4-by-5 partially-graphite-reflected operational LEU-2 core went critical on April 29. Full power was achieved on May 12, 1994. Both cores behaved very much as originally predicted. All of the old HEU fuel has been shipped to Savannah River.

  9. Innovative concepts for fuel plate fabrication

    SciTech Connect

    Domagala, R.F.; Wiencek, T.C.; Thresh, H.R.

    1987-10-01

    A number of fabrication concepts have been and are being explored at ANL. Although specific processes were addressed with silicide fuels in mind, most are applicable to fabrication with any fuel type. Processes include improved comminution procedures for converting U-Si alloy ingots to powder using a roll crusher and an impact mill. Aluminizing of core compacts by ion vapor deposition techniques in vacuum offers prospects for improved plate quality. Other items examined include the possible use of coatings on fuel particles, matrices different from pure Al, and ductile fuel alloys which might be used to produce fuel plates with uranium loadings higher than possible with conventional dispersed-phase powder metallurgy technology.

  10. Waste Stream Analyses for Nuclear Fuel Cycles

    SciTech Connect

    N. R. Soelberg

    2010-08-01

    A high-level study was performed in Fiscal Year 2009 for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) Advanced Fuel Cycle Initiative (AFCI) to provide information for a range of nuclear fuel cycle options (Wigeland 2009). At that time, some fuel cycle options could not be adequately evaluated since they were not well defined and lacked sufficient information. As a result, five families of these fuel cycle options are being studied during Fiscal Year 2010 by the Systems Analysis Campaign for the DOE NE Fuel Cycle Research and Development (FCRD) program. The quality and completeness of data available to date for the fuel cycle options is insufficient to perform quantitative radioactive waste analyses using recommended metrics. This study has been limited thus far to qualitative analyses of waste streams from the candidate fuel cycle options, because quantitative data for wastes from the front end, fuel fabrication, reactor core structure, and used fuel for these options is generally not yet available.

  11. Nuclear reactor fuel element having improved heat transfer

    DOEpatents

    Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

    1982-03-03

    A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

  12. IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

    SciTech Connect

    M.K. Meyer; J. Gan; J.-F. Jue; D.D. Keiser; E. Perez; A. Robinson; D.M. Wachs; N. Woolstenhulme; G.L. Hofman; Y.-S. Kim

    2014-04-01

    High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  13. Irradiation performance of U-Mo monolithic fuel

    SciTech Connect

    Meyer, M. K.; Gan, J.; Jue, J. F.; Keiser, D. D.; Perez, E.; Robinson, A.; Wachs, D. M.; Woolstenhulme, N.; Hofman, G. L.; Kim, Y. S.

    2014-04-01

    High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  14. Proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the integral fast reactor. [Metal fuel

    SciTech Connect

    Burris, L.; Steindler, M.; Miller, W.

    1984-01-01

    The pool-type Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps: a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented.

  15. 77 FR 137 - Applications and Amendments to Facility Operating Licenses Involving Proposed No Significant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-03

    ... recirculation loop MCPR SL from >= 1.12 to >= 1.15. The change is required to support the LSCS, Cycle 15, operation. Cycle 15 will be the first cycle of operation with a mixed core containing the following fuel... reevaluated for each reload using NRC- approved methodologies. The analyses for LSCS, Unit 1, Cycle 15 have...

  16. Double-clad nuclear-fuel safety rod

    DOEpatents

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  17. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOEpatents

    Flint, O.

    1961-01-10

    Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

  18. Neutronics calculation of RTP core

    NASA Astrophysics Data System (ADS)

    Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

    2017-01-01

    Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

  19. Gas Test Loop Booster Fuel Hydraulic Testing

    SciTech Connect

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

  20. Feasibility study on nuclear core design for soluble boron free small modular reactor

    SciTech Connect

    Rabir, Mohamad Hairie Hah, Chang Joo; Ju, Cho Sung

    2015-04-29

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.