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Sample records for gas reactor systems

  1. Fundamental Considerations Of Gas Core Reactor Systems

    NASA Astrophysics Data System (ADS)

    Bussard, Robert W.

    1988-04-01

    Fundamental physics and engineering limitations on gas core reactors (GCR) have been found from coupled effects of reactor neutronics with oscillatory core fuel gas flows and with overall gas dynamics. These show allowable regimes for system operation as natural re-sults of the basiciphysics of the system. Cylindrical BeO-moderated systems, such as the acoustic GCR (AGCR ), are found to be well-suited for pressure wave oscillation at 100-Hz frequencies. These result in stable oscillations of core gas electrical conductivity which may be used for direct alternating current electric power production in magnetohydrodynamic (MHD) convertors. In contrast, single-cavity axial-flow spherical GCR (SGCR2) systems are inherently limited by core flow and fission energy-generation fluctuation phenomena (to continuous wave operation with mixed flows that cannot sustain high internal core gas tempera-ture gradients. Resulting low mixed-mean temperatures place upper limits on system MHD electrical performance.

  2. Gas-cooled reactor power systems for space

    NASA Astrophysics Data System (ADS)

    Walter, Carl E.

    Large amounts of electric power are required for some of the systems envisioned in support of SDI. Since various applications are being considered, and an overall power architecture study has not been completed, the required power levels and corresponding operating times for specific systems are not known. The characteristics of six designs for power levels of 2, 10 and 20 MWe for operating time of 1 and 7 yrs are described. The operating conditions for these arbitrary designs were chosen to minimize system specific mass. Both gas and liquid cooled reactors are considered. The designs discussed draw heavily on the Pluto project experience. Gas cooled thermal reactors coupled with Brayton cycle power conversion appear to provide reasonable multimegawatt space power systems. An advanced radiation design must be developed which can meet the mass limit assumed. The inherent high temperature capability of the reactors considered removes the reactor as a limiting condition on system performance.

  3. Gas-cooled reactor power systems for space

    SciTech Connect

    Walter, C.E.

    1987-01-01

    In this paper the characteristics of six designs for power levels of 2, 10, and 20 MWe for operating times of 1 and 7 y are described. The operating conditions for these arbitrary designs were chosen to minimize system specific mass. The designs are based on recent work which benefits from earlier analyses of nuclear space power systems conducted at our Laboratory. Both gas- and liquid-cooled reactors had been considered. Pitts and Walter (1970) reported on the results of a detailed study of a 10-MWe lithium-cooled reactor in a potassium Rankine system. Unpublished results (1966) of a computer analysis provide details of an argon-cooled reactor in an argon Brayton system. The gas-cooled reactor design was based on extensive development work on the 500-MWth reactor for the nuclear ramjet (Pluto) as described by Walter (1964). The designs discussed here draw heavily on the Pluto project experience, which culminated in a successful full-power ground test as reported by Reynolds (1964). At higher power levels gas-cooled reactors coupled with Brayton systems with advanced radiator designs become attractive.

  4. Gas-cooled reactor for space power systems

    SciTech Connect

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors.

  5. Gas-cooled reactor power systems for space

    SciTech Connect

    Walter, C.E.

    1987-01-01

    Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system.

  6. A Gas-Cooled Reactor Surface Power System

    SciTech Connect

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  7. A gas-cooled reactor surface power system

    NASA Astrophysics Data System (ADS)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  8. A gas-cooled reactor surface power system

    SciTech Connect

    Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1{percent}Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. {copyright} {ital 1999 American Institute of Physics.}

  9. A gas-cooled reactor surface power system

    SciTech Connect

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-22

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  10. Neural net controlled tag gas sampling system for nuclear reactors

    DOEpatents

    Gross, Kenneth C.; Laug, Matthew T.; Lambert, John D. B.; Herzog, James P.

    1997-01-01

    A method and system for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod.

  11. Neural net controlled tag gas sampling system for nuclear reactors

    DOEpatents

    Gross, K.C.; Laug, M.T.; Lambert, J.B.; Herzog, J.P.

    1997-02-11

    A method and system are disclosed for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod. 12 figs.

  12. Modeling and performance of the MHTGR (Modular High-Temperature Gas-Cooled Reactor) reactor cavity cooling system

    SciTech Connect

    Conklin, J.C. )

    1990-04-01

    The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab.

  13. Acoustical gas core reactor with MHD power generation for burst power in a bimodal system

    NASA Astrophysics Data System (ADS)

    Dugan, E. T.; Jacobs, A. M.; Oliver, C. C.; Lear, W. E., Jr.

    Research is being conducted on gas core reactors for space nuclear power to establish the scientific feasibility and engineering validation of a reactor and energy conversion system that can significantly improve specific power, dynamic performance and system efficiency. Rapid achievement of burst mode (GWe) operation at core power densities of 1 kW/mL and reactor masses of a kg/MWt are research objectives; coupled with MHD conversion, system efficiencies of 40 percent for open cycle operation and heat rejection temperatures of 1500 K or higher for closed cycle operation are anticipated. The design of the gas core reactor/MHD generator configuration to directly produce pulsed electrical power, thereby alleviating external power conditioning requirements, is also a research objective.

  14. Ion transport membrane reactor systems and methods for producing synthesis gas

    SciTech Connect

    Repasky, John Michael

    2015-05-12

    Embodiments of the present invention provide cost-effective systems and methods for producing a synthesis gas product using a steam reformer system and an ion transport membrane (ITM) reactor having multiple stages, without requiring inter-stage reactant injections. Embodiments of the present invention also provide techniques for compensating for membrane performance degradation and other changes in system operating conditions that negatively affect synthesis gas production.

  15. Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing

    NASA Astrophysics Data System (ADS)

    Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; van Dyke, M. K.

    2003-01-01

    This paper describes the concept and preliminary component testing of a gas-cooled. UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.

  16. Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing

    NASA Technical Reports Server (NTRS)

    Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; VanDyke, M. K.

    2002-01-01

    This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet- sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.

  17. Response of a thermal barrier system to acoustic excitation in a gas turbine nuclear reactor

    SciTech Connect

    Betts, W.S. Jr.; Blevins, R.D.

    1980-11-01

    A gas turbine located within a High-Temperature Gas-Cooled Reactor (HTGR) induces high acoustic sound pressure levels into the primary coolant (helium). This acoustic loading induces high cycle fatigue stresses which may control the design of the thermal barrier system. This study examines the dynamic response of a thermal barrier configuration consisting of a fibrous insulation compressed against the reactor vessel by a coverplate which is held in position by a central attachment fixture. The results of dynamic vibration analyses indicate the effect of the plate size and curvature and the attachment size on the response of the thermal barrier.

  18. Improved vortex reactor system

    DOEpatents

    Diebold, James P.; Scahill, John W.

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  19. Air-lift reactor system for the treatment of waste-gas-containing monochlorobenzene.

    PubMed

    Joshi, Pradnya R; Deshmukh, Sharvari C; Morone, Amruta P; Kanade, Gajanan; Pandey, R A

    2013-01-01

    An air-lift bioreactor (ALR) system, applied for the treatment of waste-gas-containing monochlorobenzene (MCB) was seeded with pure culture of Acinetobacter calcoaceticus, isolated from soil as a starter seed. It was found that MCB was biologically converted to chloride as chloride was mineralized in the ALR. After the built up of the biomass in the ALR, the reactor parameters which have major influence on the removal efficiency and elimination capacity were studied using response surface methodology. The data generated by running the reactor for 150 days at varying conditions were fed to the model with a target to obtain the removal efficiency above 95% and the elimination capacity greater than 60%. The data analysis indicated that inlet loading was the major parameter affecting the elimination capacity and removal efficiency of >95%. The reactor when operated at optimized conditions resulted in enhanced performance of the reactor.

  20. Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging

    DOEpatents

    Gross, Kenny C.

    1994-01-01

    Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as "background" gases, further reducing the number of trial node combinations. Lastly, a "fuzzy" set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements.

  1. Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging

    DOEpatents

    Gross, K.C.

    1994-07-26

    Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as background'' gases, further reducing the number of trial node combinations. Lastly, a fuzzy'' set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements. 14 figs.

  2. Loop system with gas control of the power production in the MR reactor

    SciTech Connect

    Andreev, V.I.; Kolyadin, V.I.; Smirnov, A.I.; Yakovlev, V.V.

    1987-01-01

    An unsolved problem in reactor design is premature fuel-pin failure on account of mechanical interaction between the fuel and the sheath under nonstationary operating conditions. To examine the effects of this interaction on the viability, the authors have built an experimental system with the MR reactor. To provide for varying the power production over wide ranges, gas regulation based on /sup 3/He as neutron absorber is used. A map of core loading in the MR reactor is provided and variation in power in the experimental fuel assembly in accordance with /sup 3/He pressure and control location is shown. A structural diagram shows the reactor apparatus with gas power control in the experimental pin assembly. The relative changes in channel power in relation to neutron absorber pressure in GCU in channel 1-4 are presented. The results are offered on the power variation in the experimental assembly and reactivity as functions of /sup 3/He pressure in the GCU, together with the calculated data.

  3. Reactor vessel support system

    DOEpatents

    Golden, Martin P.; Holley, John C.

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  4. Improved vortex reactor system

    DOEpatents

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  5. Gas Evolution Measurements on Reactor Irradiated Advanced Fusion Magnet Insulation Systems

    NASA Astrophysics Data System (ADS)

    Humer, K.; Seidl, E.; Weber, H. W.; Fabian, P. E.; Feucht, S. W.; Munshi, N. A.

    2006-03-01

    Glass-fiber reinforced plastics (GFRPs) are used as insulation materials for the superconducting magnet coils of the International Thermonuclear Experimental Reactor (ITER). The radiation environment present at the magnet location will lead to gas production, swelling and weight loss of the laminate, which may result in a pressure rise combined with undefined stresses on the magnet coil casing. Consequently, these effects are important parameters for the engineering and design criteria of superconducting magnet coil structures. In this study, newly developed epoxy and cyanate-ester (CE) based S2-glass fiber reinforced insulation systems were irradiated at ambient temperature in the TRIGA-Mark II reactor (Vienna) to a fast neutron fluence of 1 and 5×1021 m-2 (E>0.1 MeV) prior to measurements of gas evolution, swelling and weight loss. The CE based laminates show increased radiation resistance, i.e. less gas evolution. The highest radiation hardness up to the highest dose was observed in a pure CE system. In addition, the effects of swelling and weight loss are either negligible or less pronounced for all systems. The results prove that the newly developed CE based composites are serious candidate insulation systems for ITER.

  6. KEY DESIGN REQUIREMENTS FOR THE HIGH TEMPERATURE GAS-COOLED REACTOR NUCLEAR HEAT SUPPLY SYSTEM

    SciTech Connect

    L.E. Demick

    2010-09-01

    Key requirements that affect the design of the high temperature gas-cooled reactor nuclear heat supply system (HTGR-NHSS) as the NGNP Project progresses through the design, licensing, construction and testing of the first of a kind HTGR based plant are summarized. These requirements derive from pre-conceptual design development completed to-date by HTGR Suppliers, collaboration with potential end users of the HTGR technology to identify energy needs, evaluation of integration of the HTGR technology with industrial processes and recommendations of the NGNP Project Senior Advisory Group.

  7. GAS COOLED NUCLEAR REACTORS

    DOEpatents

    Long, E.; Rodwell, W.

    1958-06-10

    A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

  8. Control rod system useable for fuel handling in a gas-cooled nuclear reactor

    DOEpatents

    Spurrier, Francis R.

    1976-11-30

    A control rod and its associated drive are used to elevate a complete stack of fuel blocks to a position above the core of a gas-cooled nuclear reactor. A fuel-handling machine grasps the control rod and the drive is unlatched from the rod. The stack and rod are transferred out of the reactor, or to a new location in the reactor, by the fuel-handling machine.

  9. Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept

    NASA Technical Reports Server (NTRS)

    Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyke, L. C.

    1973-01-01

    A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

  10. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    SciTech Connect

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  11. System Engineering Program Applicability for the High Temperature Gas-Cooled Reactor (HTGR) Component Test Capability (CTC)

    SciTech Connect

    Jeffrey Bryan

    2009-06-01

    This white paper identifies where the technical management and systems engineering processes and activities to be used in establishing the High Temperature Gas-cooled Reactor (HTGR) Component Test Capability (CTC) should be addressed and presents specific considerations for these activities under each CTC alternative

  12. Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System

    SciTech Connect

    Frisani, Angelo; Hassan, Yassin A; Ugaz, Victor M

    2010-11-02

    The design of passive heat removal systems is one of the main concerns for the modular very high temperature gas-cooled reactors (VHTR) vessel cavity. The reactor cavity cooling system (RCCS) is a key heat removal system during normal and off-normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The computational fluid dynamics (CFD) STAR-CCM+/V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. A CFD model was developed to analyze heat exchange in the RCCS. The model incorporates a 180-deg section resembling the VHTR RCCS experimentally reproduced in a laboratory-scale test facility at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations. The objective of the present work was to benchmark CFD tools against experimental data addressing the behavior of the RCCS following accident conditions. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls' temperature below design limits. Different temperature profiles at the reactor pressure vessel (RPV) wall obtained from the experimental facility were used as boundary conditions in the numerical analyses to simulate VHTR transient evolution during accident scenarios. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The comparison among the different turbulence models analyzed showed satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. For such a complicated geometry and flow conditions, the tested turbulence models demonstrated that the

  13. Nuclear reactor sealing system

    DOEpatents

    McEdwards, James A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  14. A modular gas-cooled cermet reactor system for planetary base power

    NASA Astrophysics Data System (ADS)

    Jahshan, S. N.; Borkowski, J. A.

    1992-10-01

    Fission nuclear power is foreseen as the source for electricity in planetary colonization and exploration. This report discusses a six module, gas-cooled, cermet-fueled reactor which is proposed to meet the design objectives. The highly enriched core is compact and can operate at high temperatures for long periods of time. The helium coolant powers six modular Brayton cycles that compare favorably with the SP-100-based Brayton cycle.

  15. A modular gas-cooled cermet reactor system for planetary base power

    SciTech Connect

    Jahshan, S.N.; Borkowski, J.A. )

    1993-01-15

    Fission nuclear power is foreseen as the source for electricity in planetary colonization and exploration. A six module gas-cooled, cermet-fueled reactor is proposed that can meet the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers six modular Brayton cycles that compare favorably with the SP-100-based Brayton cycle.

  16. Moon base reactor system

    NASA Technical Reports Server (NTRS)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  17. Reactor vessel support system. [LMFBR

    DOEpatents

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  18. Hot Gas Desulfurization Using Transport Reactors

    SciTech Connect

    Moorehead, E.L.

    1996-12-31

    Sierra Pacific Power Company is building a 100 MW, IGCC power plant based on KRW fluid bed gasifier technology that utilizes transport reactors for hot gas desulfurization and sorbent regeneration. Use of a transport absorber avoids the need for pre-filtration of dust-laden gasifier effluent, while a transport regenerator allows for the use of 100% air without the need for heat exchange equipment. Selection of transport reactors for hot gas desulfurization using a proprietary sorbent, based on testing performed in a transport reactor test unit (TRTU) at the M. W. Kellogg Technology Development Center and in a fixed bed reactor at Morgantown Energy Technology Center (METC), is outlined. The results obtained in these two test facilities and reasons for selecting transport reactors for the IGCC power plant in preference to either fixed bed or fluidized bed reactors are discussed. This paper reviews the evolution of the hot gas desulfurization system designs and includes selected results on H{sub 2}S absorption and regeneration of sulfided sorbent over several absorption/regeneration cycles conducted in the TRTU and the METC fixed bed reactor. The original design for the Sierra Pacific Project was based on fixed bed reactors with zinc ferrite as the sorbent. Owing to the high steam requirements of this sorbent, zinc titanate was selected and tested in a fixed bed reactor and was found unacceptable due to loss of strength on cyclic absorption/regeneration operation. Another sorbent evaluated was Z-Sorb{reg_sign}, a proprietary sorbent developed by Phillips Petroleum Company, was found to have excellent sulfur capacity, structural strength and regenerability. Steam was found unsuitable as fixed bed regenerator diluent, this results in a requirement for a large amount of inert gas, whereas a transport regenerator requires no diluent. The final Sierra design features transport reactors for both desulfurization and regeneration steps using neat air. 3 refs., 3 figs., 2 tabs.

  19. Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

    SciTech Connect

    Ballinger, Ronald G.; Wang, Chun Yun; Kadak, Andrew; Todreas, Neil; Mirick, Bradley; Demetri, Eli; Koronowski, Martin

    2004-08-30

    The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a Generation IV nuclear system. The availability of controllable helium turbomachinery and compact heat exchangers are thus the critical enabling technology for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has been accomplished with the overriding constraint that this design could be built with existing technology and complies with all current codes and standards. Using the initial reference design, limiting features were identified. Finally, an optimized reference design was developed by identifying key advances in the technology that could reasonably be expected to be achieved with limited R&D. This final reference design is an indirect, intercooled and recuperated cycle consisting of a three-shaft arrangement for the turbomachinery system. A critical part of the design process involved the interaction between individual component design and overall plant performance. The helium cycle overall efficiency is significantly influenced by performance of individual components. Changes in the design of one component, a turbine for example, often required changes in other components. To allow for the optimization of the overall design with these interdependencies, a detailed steady state and transient control model was developed. The use of the steady state and transient models as a part of an iterative design process represents a key contribution of this work. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion system simultaneously. Physical parameters such as the heat exchangers; weights and practical performance maps such as the turbine characteristics and compressor characteristics are incorporated into the model. The individual component models as well as the fully integrated model of the

  20. A gas-cooled cermet reactor system for planetary base power

    SciTech Connect

    Jahshan, S.N.; Borkowski, J.A.

    1992-01-01

    Fission nuclear power is foreseen as the source for electricity in colonization exploration. A gas-cooled, cermet-fueled reactor is proposed that can meet many of the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers a Brayton cycle that compares well with the SP-100-based Brayton cycle. The power cycle can be upgraded further under certain siting-related conditions by the addition of a low temperature Rankine cycle.

  1. A gas-cooled cermet reactor system for planetary base power

    SciTech Connect

    Jahshan, S.N.; Borkowski, J.A.

    1992-08-01

    Fission nuclear power is foreseen as the source for electricity in colonization exploration. A gas-cooled, cermet-fueled reactor is proposed that can meet many of the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers a Brayton cycle that compares well with the SP-100-based Brayton cycle. The power cycle can be upgraded further under certain siting-related conditions by the addition of a low temperature Rankine cycle.

  2. Process Model of the Gas Recovery System in an IFE reactor

    NASA Astrophysics Data System (ADS)

    Gentile, Charles; Aristova, Maria

    2007-11-01

    It is necessary to develop a detailed representative model for the fuel recovery system (FRS) in the prospective direct drive inertial fusion energy (IFE) reactor. In order to observe the interaction of all components, a chemical process model is developed as part of the conceptual design phase of the project. Initially, the reactants, system structure, and processes are defined using the known contents of the vacuum vessel exhaust. The output, which will include physical properties and chemical content of the products, is analyzed to determine the most efficient and productive system parameters. The results of the modeling will be presented in this paper. This modeling exercise will be instrumental in optimizing and closing the fusion fuel cycle in the IFE power reactor.

  3. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    SciTech Connect

    Hassan, Yassin; Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  4. Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.

    2007-01-01

    In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is

  5. Reactor System Transient Code.

    1999-07-14

    RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.

  6. Rapid starting methanol reactor system

    DOEpatents

    Chludzinski, Paul J.; Dantowitz, Philip; McElroy, James F.

    1984-01-01

    The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

  7. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  8. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

  9. Gas-cooled nuclear reactor

    DOEpatents

    Peinado, Charles O.; Koutz, Stanley L.

    1985-01-01

    A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure. If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel. A metal liner covers the entire inside concrete surfaces of the concrete pressure vessel, and cooling tubes are welded to the exterior or concrete side of the metal liner. The gas coolant is in direct contact with the interior surface of the metal liner and transfers its heat through the metal liner to the liquid coolant flowing through the cooling tubes. The cooler gas is more dense and creates a downward convection flow in the region between the core and the sidewall until it reaches the bottom of the concrete pressure vessel when it flows radially inward and up into the core for another pass. Water is forced to flow through the cooling tubes to absorb heat from the core at a sufficient rate to remove enough of the decay heat created in the core to prevent overheating of the core or the vessel.

  10. Gas Turbine Energy Conversion Systems for Nuclear Power Plants Applicable to LiFTR Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.

    2014-01-01

    This panel plans to cover thermal energy and electric power production issues facing our nation and the world over the next decades, with relevant technologies ranging from near term to mid-and far term.Although the main focus will be on ground based plants to provide baseload electric power, energy conversion systems (ECS) for space are also included, with solar- or nuclear energy sources for output power levels ranging tens of Watts to kilo-Watts for unmanned spacecraft, and eventual mega-Watts for lunar outposts and planetary surface colonies. Implications of these technologies on future terrestrial energy systems, combined with advanced fracking, are touched upon.Thorium based reactors, and nuclear fusion along with suitable gas turbine energy conversion systems (ECS) will also be considered by the panelists. The characteristics of the above mentioned ECS will be described, both in terms of their overall energy utilization effectiveness and also with regard to climactic effects due to exhaust emissions.

  11. Gas Reactor Plant Analyzer and Simulator for Hydrogen Production

    2004-01-01

    This software is used to study and analyze various configurations of plant equipment for gas cooled nuclear reactor applications. The user of this software would likely be interested in optimizing the economic, safety, and operating performance of this type of reactor. The code provides the capability for the user through his input to configure networks of nuclear reactor components. The components available include turbine, compressor, heat exchanger, reactor core, coolers, bypass valves, and control systems.

  12. Reactor system for upgrading light olefins in staged reactors

    SciTech Connect

    Harandi, M.N.; Owen, H.

    1991-05-28

    This patent describes a semicontinuous multistage catalytic system for conversion of light olefinic gas feedstock to distillate range hydrocarbons. It comprises: a first reactor means; means for passing hot feedstock vapor upwardly; fluid handling means; means for recovering from primary stage effluent stream; second reactor means; and fluid handling means for periodically interrupting flow.

  13. Gas tagging and cover gas combination for nuclear reactor

    DOEpatents

    Gross, Kenny C.; Laug, Matthew T.

    1985-01-01

    The invention discloses the use of stable isotopes of neon and argon, that are grouped in preselected different ratios one to the other and are then sealed as tags in different cladded nuclear fuel elements to be used in a liquid metal fast breeder reactor. Failure of the cladding of any fuel element allows fission gases generated in the reaction and these tag isotopes to escape and to combine with the cover gas held in the reactor over the fuel elements. The isotopes specifically are Ne.sup.20, Ne.sup.21 and Ne.sup.22 of neon and Ar.sup.36, Ar.sup.38 and Ar.sup.40 of argon, and the cover gas is helium. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between approximately 0.degree. and -25.degree. C. operable to remove the fission gases from the cover gas and tags and the second or tag recovery system bed is held between approximately -170.degree. and -185.degree. C. operable to isolate the tags from the cover gas. Spectrometric analysis further is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be specifically determined.

  14. Reactor water cleanup system

    DOEpatents

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  15. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  16. HORIZONTAL BOILING REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  17. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Daniels, F.

    1957-10-15

    Gas-cooled solid-moderator type reactors wherein the fissionable fuel and moderator materials are each in the form of solid pebbles, or discrete particles, and are substantially homogeneously mixed in the proper proportion and placed within the core of the reactor are described. The shape of these discrete particles must be such that voids are present between them when mixed together. Helium enters the bottom of the core and passes through the voids between the fuel and moderator particles to absorb the heat generated by the chain reaction. The hot helium gas is drawn off the top of the core and may be passed through a heat exchanger to produce steam.

  18. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Goett, J.J.

    1961-01-24

    A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

  19. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    SciTech Connect

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

  20. Improved gas tagging and cover gas combination for nuclear reactor

    DOEpatents

    Gross, K.C.; Laug, M.T.

    1983-09-26

    The invention discloses the use of stable isotopes of neon and argon, sealed as tags in different cladding nuclear fuel elements to be used in a liquid metal fast breeder reactor. Cladding failure allows fission gases and these tag isotopes to escape and to combine with the cover gas. The isotopes are Ne/sup 20/, Ne/sup 21/ and Ne/sup 22/ and Ar/sup 36/, Ar/sup 38/ and Ar/sup 40/, and the cover gas is He. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between 0 and -25/sup 0/C to remove the fission gases from the cover gas and tags, and the second or tag recovery system bed between -170 and -185/sup 0/C to isolate the tags from the cover gas. Spectrometric analysis is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be determined.

  1. ECO LOGIC INTERNATIONAL GAS-PHASE CHEMICAL REDUCTION PROCESS - THE REACTOR SYSTEM - APPLICATIONS ANALYSIS REPORT

    EPA Science Inventory

    The ELI Eco Logic International Inc. (Eco Logic) process thermally separates organics, then chemically reduces them in a hydrogen atmosphere, converting them to a reformed gas that consists of light hydrocarbons and water. A scrubber treats the reformed gas to remove hydrogen chl...

  2. Dynamic neutronic and stability analysis of a burst mode, single cavity gas core reactor Brayton cycle space power system

    NASA Astrophysics Data System (ADS)

    Dugan, Edward T.; Kutikkad, Kiratadas

    The conceptual, burst-mode gaseous-core reactor (GCR) space nuclear power system presently subjected to reactor-dynamics and system stability studies operates on a closed Brayton cycle, via disk MHD generator for energy conversion. While the gaseous fuel density power coefficient of reactivity is found to be capable of rapidly stabilizing the GCR system, the power of this feedback renders standard external reactivity insertions inadequate for significant power-level changes during normal operation.

  3. Testing of Gas Reactor Fuel and Materials in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2006-10-01

    The recent growth in interest for high temperature gas reactors has resulted in an increased need for materials and fuel testing for this type of reactor. The Advanced Test Reactor (ATR), located at the US Department of Energy’s Idaho National Laboratory, has long been involved in testing gas reactor fuel and materials, and has facilities and capabilities to provide the right environment for gas reactor irradiation experiments. These capabilities include both passive sealed capsule experiments, and instrumented/actively controlled experiments. The instrumented/actively controlled experiments typically contain thermocouples and control the irradiation temperature, but on-line measurements and controls for pressure and gas environment have also been performed in past irradiations. The ATR has an existing automated gas temperature control system that can maintain temperature in an irradiation experiment within very tight bounds, and has developed an on-line fission product monitoring system that is especially well suited for testing gas reactor particle fuel. The ATR’s control system, which consists primarily of vertical cylinders used to rotate neutron poisons/reflectors toward or away from the reactor core, provides a constant vertical flux profile over the duration of each operating cycle. This constant chopped cosine shaped axial flux profile, with a relatively flat peak at the vertical centre of the core, is more desirable for experiments than a constantly moving axial flux peak resulting from a control system of axially positioned control components which are vertically withdrawn from the core.

  4. REACTOR CONTROL SYSTEM

    DOEpatents

    MacNeill, J.H.; Estabrook, J.Y.

    1960-05-10

    A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

  5. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1959-02-10

    A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

  6. Fuel leak detection apparatus for gas cooled nuclear reactors

    DOEpatents

    Burnette, Richard D.

    1977-01-01

    Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

  7. Fuel development for gas-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Meyer, M. K.; Fielding, R.; Gan, J.

    2007-09-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  8. Reactor models for a series of continuous stirred tank reactors with a gas-liquid-solid leaching system: Part I. Surface reaction control

    NASA Astrophysics Data System (ADS)

    Papangelakis, V. G.; Demopoulos, G. P.

    1992-12-01

    In this three-part series of articles, comprehensive three-phase steady-state hydrometallurgical reactor models of the continuous stirred tank reactor (CSTR) type are developed and applied to a commercial (pressure oxidation) process. The key features of the developed models are the coupling of both mass and heat balance equations, the description of the nonisothermal performance (autothermal) of a multistage continuous reactor, and the treatment of multimineral feed materials. The model considers only the oxidation reactions, because they mainly affect the thermal balance of the reactor. The stoichiometries and intrinsic kinetics of the heterogeneous leaching reactions, which are established via independent experiments, are the foundation of the developed model. A three-phase (g-l-s) reaction process might be controlled by either surface reaction control, i.e., the rate(s) of the heterogeneous leaching reaction(s), or by gas transfer control, i.e., the rate of transfer of the gaseous reactant into the liquid phase. In the present article (Part I), the case of surface reaction control is treated. The article addresses, in particular, the following topics: (1) it outlines the basic mass and heat balance equations which describe the performance of a multistage leaching reactor; (2) it presents a continuous function to describe the particle size distribution of the feed; and (3) it develops, on the basis of probability theory, number- and mass-particle size density functions which give the size distribution of particle populations reacting according to the surface reaction control-shrinking core model.

  9. Engineering Development of Ceramic Membrane Reactor System for Converting Natural Gas to Hydrogen and Synthesis Gas for Liquid Transportation Fuels

    SciTech Connect

    Air Products and Chemicals

    2008-09-30

    An Air Products-led team successfully developed ITM Syngas technology from the concept stage to a stage where a small-scale engineering prototype was about to be built. This technology produces syngas, a gas containing carbon monoxide and hydrogen, by reacting feed gas, primarily methane and steam, with oxygen that is supplied through an ion transport membrane. An ion transport membrane operates at high temperature and oxygen ions are transported through the dense membrane's crystal lattice when an oxygen partial pressure driving force is applied. This development effort solved many significant technical challenges and successfully scaled-up key aspects of the technology to prototype scale. Throughout the project life, the technology showed significant economic benefits over conventional technologies. While there are still on-going technical challenges to overcome, the progress made under the DOE-funded development project proved that the technology was viable and continued development post the DOE agreement would be warranted.

  10. Pin-Type Gas Cooled Reactor for Nuclear Electric Propulsion

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Lipinski, Ronald J.

    2003-01-01

    This paper describes a point design for a pin-type Gas-Cooled Reactor concept that uses a fuel pin design similar to the SP100 fuel pin. The Gas-Cooled Reactor is designed to operate at 100 kWe for 7 years plus have a reduced power mode of 20% power for a duration of 5 years. The power system uses a gas-cooled, UN-fueled, pin-type reactor to heat He/Xe gas that flows directly into a recuperated Brayton system to produce electricity. Heat is rejected to space via a thermal radiator that unfolds in space. The reactor contains approximately 154 kg of 93.15 % enriched UN in 313 fuel pins. The fuel is clad with rhenium-lined Nb-1Zr. The pressures vessel and ducting are cooled by the 900 K He/Xe gas inlet flow or by thermal radiation. This permits all pressure boundaries to be made of superalloy metals rather than refractory metals, which greatly reduces the cost and development schedule required by the project. The reactor contains sufficient rhenium (a neutron poison) to make the reactor subcritical under water immersion accidents without the use of internal shutdown rods. The mass of the reactor and reflectors is about 750 kg.

  11. Method of producing silicon. [gas phase reactor multiple injector liquid feed system

    NASA Technical Reports Server (NTRS)

    Wolf, C. B.; Meyer, T. N. (Inventor)

    1980-01-01

    A liquid reactant injector assembly suited for the injection of liquid reactant into a high temperature metal reductant vapor and carrier gas stream for the production of metal is presented. The assembly is especially adapted for the continuous production of high purity silicon by the reduction of SiCl4 with sodium. The assembly includes a refractory-lined, hollow metal shell having ten equally-spaced, concentric, radially directed ports provided in the shell and wall. A hydraulic, atomizing type spray nozzle is mounted in each of the ports recessed from the inner wall surface.

  12. Design and part-load performance of a hybrid system based on a solid oxide fuel cell reactor and a micro gas turbine

    NASA Astrophysics Data System (ADS)

    Costamagna, P.; Magistri, L.; Massardo, A. F.

    This paper addresses the design and off-design analysis of a hybrid system (HS) based on the coupling of a recuperated micro gas turbine (MGT) with a high temperature solid oxide fuel cell (SOFC) reactor. The SOFC reactor model is presented and discussed, taking into account the influence of the reactor lay-out, the current density, the air utilisation factor, the cell operating temperature, etc. The SOFC design and off-design performance is presented and discussed; the design and off-design models of a recuperated micro-gas turbine are also presented. The operating line, the influence of the micro gas turbine "variable speed" control, and the efficiency behaviour at part load are analysed in depth. Finally, the model of the hybrid system obtained by coupling the MGT and the SOFC reactor, considering the compatibility (technological constraints) of the two systems, is presented. The model allows the evaluation of the design and off-design behaviour of the hybrid system, particularly when the MGT variable speed control system is considered. The thermal efficiency of the hybrid system, taking into account its size (250/300 kW e), is noteworthy: higher than 60% at design point, and also very high at part load conditions. Such a result is mainly due to the simultaneous positive influence of SOFC off-design behaviour and MGT variable speed control. Moreover, it is possible to recover the waste heat from the gas at the MGT recuperator outlet ( Tgas is about 250°C) for cogeneration purposes.

  13. NOVEL REACTOR FOR THE PRODUCTION OF SYNTHESIS GAS

    SciTech Connect

    Vasilis Papavassiliou; Leo Bonnell; Dion Vlachos

    2004-12-01

    Praxair investigated an advanced technology for producing synthesis gas from natural gas and oxygen This production process combined the use of a short-reaction time catalyst with Praxair's gas mixing technology to provide a novel reactor system. The program achieved all of the milestones contained in the development plan for Phase I. We were able to develop a reactor configuration that was able to operate at high pressures (up to 19atm). This new reactor technology was used as the basis for a new process for the conversion of natural gas to liquid products (Gas to Liquids or GTL). Economic analysis indicated that the new process could provide a 8-10% cost advantage over conventional technology. The economic prediction although favorable was not encouraging enough for a high risk program like this. Praxair decided to terminate development.

  14. Scaling approach and thermal-hydraulic analysis in the reactor cavity cooling system of a high temperature gas -cooled reactor and thermal-jet mixing in a sodium fast reactor

    NASA Astrophysics Data System (ADS)

    Omotowa, Olumuyiwa A.

    This dissertation develops and demonstrates the application of the top-down and bottom-up scaling methodologies to thermal-hydraulic flows in the reactor cavity cooling system (RCCS) of the high temperature gas reactor (HTGR) and upper plenum of the sodium fast reactor (SFR), respectively. The need to integrate scaled separate effects and integral tests was identified. Experimental studies and computational tools (CFD) have been integrated to guide the engineering design, analysis and assessment of this scaling methods under single and two-phase flow conditions. To test this methods, two applicable case studies are considered, and original contributions are noted. Case 1: "Experimental Study of RCCS for the HTGR". Contributions include validation of scaling analysis using the top-down approach as guide to a ¼-scale integral test facility. System code, RELAP5, was developed based on the derived scaling parameters. Tests performed included system sensitivity to decay heat load and heat sink inventory variations. System behavior under steady-state and transient scenarios were predicted. Results show that the system has the capacity to protect the cavity walls from over-heating during normal operations and provide a means for decay heat removal under accident scenarios. A full width half maximum statistical method was devised to characterize the thermal-hydraulics of the non-linear two-phase oscillatory behavior. This facilitated understanding of the thermal hydraulic coupling of the loop segments of the RCCS, the heat transfer, and the two-phase flashing flow phenomena; thus the impact of scaling overall. Case 2: "Computational Studies of Thermal Jet Mixing in SFR". In the pool-type SFR, susceptible regions to thermal striping are the upper instrumentation structure and the intermediate heat exchanger (IHX). We investigated the thermal mixing above the core to UIS and the potential impact due to poor mixing. The thermal mixing of dual-jet flows at different

  15. An atmospheric pressure high-temperature laminar flow reactor for investigation of combustion and related gas phase reaction systems.

    PubMed

    Oßwald, Patrick; Köhler, Markus

    2015-10-01

    A new high-temperature flow reactor experiment utilizing the powerful molecular beam mass spectrometry (MBMS) technique for detailed observation of gas phase kinetics in reacting flows is presented. The reactor design provides a consequent extension of the experimental portfolio of validation experiments for combustion reaction kinetics. Temperatures up to 1800 K are applicable by three individually controlled temperature zones with this atmospheric pressure flow reactor. Detailed speciation data are obtained using the sensitive MBMS technique, providing in situ access to almost all chemical species involved in the combustion process, including highly reactive species such as radicals. Strategies for quantifying the experimental data are presented alongside a careful analysis of the characterization of the experimental boundary conditions to enable precise numeric reproduction of the experimental results. The general capabilities of this new analytical tool for the investigation of reacting flows are demonstrated for a selected range of conditions, fuels, and applications. A detailed dataset for the well-known gaseous fuels, methane and ethylene, is provided and used to verify the experimental approach. Furthermore, application for liquid fuels and fuel components important for technical combustors like gas turbines and engines is demonstrated. Besides the detailed investigation of novel fuels and fuel components, the wide range of operation conditions gives access to extended combustion topics, such as super rich conditions at high temperature important for gasification processes, or the peroxy chemistry governing the low temperature oxidation regime. These demonstrations are accompanied by a first kinetic modeling approach, examining the opportunities for model validation purposes.

  16. An atmospheric pressure high-temperature laminar flow reactor for investigation of combustion and related gas phase reaction systems.

    PubMed

    Oßwald, Patrick; Köhler, Markus

    2015-10-01

    A new high-temperature flow reactor experiment utilizing the powerful molecular beam mass spectrometry (MBMS) technique for detailed observation of gas phase kinetics in reacting flows is presented. The reactor design provides a consequent extension of the experimental portfolio of validation experiments for combustion reaction kinetics. Temperatures up to 1800 K are applicable by three individually controlled temperature zones with this atmospheric pressure flow reactor. Detailed speciation data are obtained using the sensitive MBMS technique, providing in situ access to almost all chemical species involved in the combustion process, including highly reactive species such as radicals. Strategies for quantifying the experimental data are presented alongside a careful analysis of the characterization of the experimental boundary conditions to enable precise numeric reproduction of the experimental results. The general capabilities of this new analytical tool for the investigation of reacting flows are demonstrated for a selected range of conditions, fuels, and applications. A detailed dataset for the well-known gaseous fuels, methane and ethylene, is provided and used to verify the experimental approach. Furthermore, application for liquid fuels and fuel components important for technical combustors like gas turbines and engines is demonstrated. Besides the detailed investigation of novel fuels and fuel components, the wide range of operation conditions gives access to extended combustion topics, such as super rich conditions at high temperature important for gasification processes, or the peroxy chemistry governing the low temperature oxidation regime. These demonstrations are accompanied by a first kinetic modeling approach, examining the opportunities for model validation purposes. PMID:26520986

  17. An atmospheric pressure high-temperature laminar flow reactor for investigation of combustion and related gas phase reaction systems

    SciTech Connect

    Oßwald, Patrick; Köhler, Markus

    2015-10-15

    A new high-temperature flow reactor experiment utilizing the powerful molecular beam mass spectrometry (MBMS) technique for detailed observation of gas phase kinetics in reacting flows is presented. The reactor design provides a consequent extension of the experimental portfolio of validation experiments for combustion reaction kinetics. Temperatures up to 1800 K are applicable by three individually controlled temperature zones with this atmospheric pressure flow reactor. Detailed speciation data are obtained using the sensitive MBMS technique, providing in situ access to almost all chemical species involved in the combustion process, including highly reactive species such as radicals. Strategies for quantifying the experimental data are presented alongside a careful analysis of the characterization of the experimental boundary conditions to enable precise numeric reproduction of the experimental results. The general capabilities of this new analytical tool for the investigation of reacting flows are demonstrated for a selected range of conditions, fuels, and applications. A detailed dataset for the well-known gaseous fuels, methane and ethylene, is provided and used to verify the experimental approach. Furthermore, application for liquid fuels and fuel components important for technical combustors like gas turbines and engines is demonstrated. Besides the detailed investigation of novel fuels and fuel components, the wide range of operation conditions gives access to extended combustion topics, such as super rich conditions at high temperature important for gasification processes, or the peroxy chemistry governing the low temperature oxidation regime. These demonstrations are accompanied by a first kinetic modeling approach, examining the opportunities for model validation purposes.

  18. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    SciTech Connect

    Not Available

    1980-09-01

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases.

  19. Versatile in situ gas analysis apparatus for nanomaterials reactors.

    PubMed

    Meysami, Seyyed Shayan; Snoek, Lavina C; Grobert, Nicole

    2014-09-01

    We report a newly developed technique for the in situ real-time gas analysis of reactors commonly used for the production of nanomaterials, by showing case-study results obtained using a dedicated apparatus for measuring the gas composition in reactors operating at high temperature (<1000 °C). The in situ gas-cooled sampling probe mapped the chemistry inside the high-temperature reactor, while suppressing the thermal decomposition of the analytes. It thus allows a more accurate study of the mechanism of progressive thermocatalytic cracking of precursors compared to previously reported conventional residual gas analyses of the reactor exhaust gas and hence paves the way for the controlled production of novel nanomaterials with tailored properties. Our studies demonstrate that the composition of the precursors dynamically changes as they travel inside of the reactor, causing a nonuniform growth of nanomaterials. Moreover, mapping of the nanomaterials reactor using quantitative gas analysis revealed the actual contribution of thermocatalytic cracking and a quantification of individual precursor fragments. This information is particularly important for quality control of the produced nanomaterials and for the recycling of exhaust residues, ultimately leading toward a more cost-effective continuous production of nanomaterials in large quantities. Our case study of multiwall carbon nanotube synthesis was conducted using the probe in conjunction with chemical vapor deposition (CVD) techniques. Given the similarities of this particular CVD setup to other CVD reactors and high-temperature setups generally used for nanomaterials synthesis, the concept and methodology of in situ gas analysis presented here does also apply to other systems, making it a versatile and widely applicable method across a wide range of materials/manufacturing methods, catalysis, as well as reactor design and engineering.

  20. Versatile in situ gas analysis apparatus for nanomaterials reactors.

    PubMed

    Meysami, Seyyed Shayan; Snoek, Lavina C; Grobert, Nicole

    2014-09-01

    We report a newly developed technique for the in situ real-time gas analysis of reactors commonly used for the production of nanomaterials, by showing case-study results obtained using a dedicated apparatus for measuring the gas composition in reactors operating at high temperature (<1000 °C). The in situ gas-cooled sampling probe mapped the chemistry inside the high-temperature reactor, while suppressing the thermal decomposition of the analytes. It thus allows a more accurate study of the mechanism of progressive thermocatalytic cracking of precursors compared to previously reported conventional residual gas analyses of the reactor exhaust gas and hence paves the way for the controlled production of novel nanomaterials with tailored properties. Our studies demonstrate that the composition of the precursors dynamically changes as they travel inside of the reactor, causing a nonuniform growth of nanomaterials. Moreover, mapping of the nanomaterials reactor using quantitative gas analysis revealed the actual contribution of thermocatalytic cracking and a quantification of individual precursor fragments. This information is particularly important for quality control of the produced nanomaterials and for the recycling of exhaust residues, ultimately leading toward a more cost-effective continuous production of nanomaterials in large quantities. Our case study of multiwall carbon nanotube synthesis was conducted using the probe in conjunction with chemical vapor deposition (CVD) techniques. Given the similarities of this particular CVD setup to other CVD reactors and high-temperature setups generally used for nanomaterials synthesis, the concept and methodology of in situ gas analysis presented here does also apply to other systems, making it a versatile and widely applicable method across a wide range of materials/manufacturing methods, catalysis, as well as reactor design and engineering. PMID:25090251

  1. DESIGN OF AN ON-LINE, MULTI-SPECTROMETER FISSION PRODUCT MONITORING SYSTEM (FPMS) TO SUPPORT ADVANCED GAS REACTOR (AGR) FUEL TESTING AND QUALIFICATION IN THE ADVANCED TEST REACTOR

    SciTech Connect

    J. K. Hartwell; D. M. Scates; M. W. Drigert

    2005-11-01

    The US Department of Energy (DOE) is embarking on a series of tests of coated-particle reactor fuel for the Advanced Gas Reactor (AGR). As one part of this fuel development program a series of eight (8) fuel irradiation tests are planned for the Idaho National Laboratory’s (INL’s) Advanced Test Reactor (ATR). The first test in this series (AGR-1) will incorporate six separate “capsules” irradiated simultaneously, each containing about 51,000 TRISO-coated fuel particles supported in a graphite matrix and continuously swept with inert gas during irradiation. The effluent gas from each of the six capsules must be independently monitored in near real time and the activity of various fission gas nuclides determined and reported. A set of seven heavily-shielded high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based total radiation detectors have been designed, and are being configured and tested for use during the AGR-1 experiment. The AGR-1 test specification requires that the AGR-1 fission product measurement system (FPMS) have sufficient sensitivity to detect the failure of a single coated fuel particle and sufficient range to allow it to “count” multiple (up to 250) successive particle failures. This paper describes the design and expected performance of the AGR-1 FPMS.

  2. Installation and Final Testing of an On-Line, Multi-Spectrometer Fission Product Monitoring System (FPMS) to Support Advanced Gas Reactor (AGR) Fuel Testing and Qualification in the Advanced Test Reactor

    SciTech Connect

    J. K. Hartwell; D. M. Scates; M. W. Drigert; J. B. Walter

    2006-10-01

    The US Department of Energy (DOE) is initiating tests of reactor fuel for use in an Advanced Gas Reactor (AGR). The AGR will use helium coolant, a low-power-density ceramic core, and coated-particle fuel. A series of eight (8) fuel irradiation tests are planned for the Idaho National Laboratory’s (INL’s) Advanced Test Reactor (ATR). One important measure of fuel performance in these tests is quantification of the fission gas releases over the nominal 2-year duration of each irradiation experiment. This test objective will be met using the AGR Fission Product Monitoring System (FPMS) which includes seven (7) on-line detection stations viewing each of the six test capsule effluent lines (plus one spare). Each station incorporates both a heavily-shielded high-purity germanium (HPGe) gamma-ray spectrometer for quantification of the isotopic releases, and a NaI(Tl) scintillation detector to monitor the total count rate and identify the timing of the releases. The AGR-1 experiment will begin irradiation after October 1, 2006. To support this experiment, the FPMS has been completely assembled, tested, and calibrated in a laboratory at the INL, and then reassembled and tested in its final location in the ATR reactor basement. This paper presents the details of the equipment performance, the control and acquisition software, the test plan for the irradiation monitoring, and the installation in the ATR basement. Preliminary on-line data may be available by the Conference date.

  3. Integral reactor system and method for fuel cells

    DOEpatents

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  4. Rodded shutdown system for a nuclear reactor

    DOEpatents

    Golden, Martin P.; Govi, Aldo R.

    1978-01-01

    A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core.

  5. Nuclear reactor shutdown system

    DOEpatents

    Bhate, Suresh K.; Cooper, Martin H.; Riffe, Delmar R.; Kinney, Calvin L.

    1981-01-01

    An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

  6. Gas core reactors for coal gasification

    NASA Technical Reports Server (NTRS)

    Weinstein, H.

    1976-01-01

    The concept of using a gas core reactor to produce hydrogen directly from coal and water is presented. It is shown that the chemical equilibrium of the process is strongly in favor of the production of H2 and CO in the reactor cavity, indicating a 98% conversion of water and coal at only 1500 K. At lower temperatures in the moderator-reflector cooling channels the equilibrium strongly favors the conversion of CO and additional H2O to CO2 and H2. Furthermore, it is shown the H2 obtained per pound of carbon has 23% greater heating value than the carbon so that some nuclear energy is also fixed. Finally, a gas core reactor plant floating in the ocean is conceptualized which produces H2, fresh water and sea salts from coal.

  7. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  8. The DOE advanced gas reactor fuel development and qualification program

    NASA Astrophysics Data System (ADS)

    Petti, David; Maki, John; Hunn, John; Pappano, Pete; Barnes, Charles; Saurwein, John; Nagley, Scott; Kendall, Jim; Hobbins, Richard

    2010-09-01

    The high outlet temperatures and high thermal-energy conversion efficiency of modular high-temperature gas-cooled reactors (HTGRs) enable an efficient and cost-effective integration of the reactor system with non-electricity-generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300°C and 900°C. The U.S. Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. An overview of the program and recent progress is presented.

  9. Gas core reactor concepts and technology - Issues and baseline strategy

    NASA Technical Reports Server (NTRS)

    Diaz, Nils J.; Dugan, Edward T.; Kahook, Samer; Maya, Isaac

    1991-01-01

    Results of a research program including phenomenological studies, conceptual design, and systems analysis of a series of gaseous/vapor fissile fuel driven engines for space power platforms and for thermal and electric propulsion are reviewed. It is noted that gas and vapor phase reactors provide the path for minimum mass in orbit and trip times, with a specific impulse from 1020 sec at the lowest technololgical risk to 5200 sec at the highest technological risk. The discussion covers various configurations of gas core reactors and critical technologies and the nuclear vapor thermal rocket engine.

  10. Energetic closed-cycle gas core reactors for orbit raising

    NASA Technical Reports Server (NTRS)

    Rosa, R. J.; Myrabo, L. N.

    1983-01-01

    Closed-cycle gas core reactor power plants can be of two types. In the 'mixed flow' type, the gaseous nuclear fuel is intimately mixed with the working gas in the cavity. In the 'light bulb' type the fissioning plasma is enclosed in a transparent tube, and energy transfer to the separate working gas occurs by thermal radiation. The potentials of high temperature gas core reactors in terrestrial electric power generator applications have been considered, and a number of civilian power-beaming applications for gaseous fuel nuclear-MHD power plants in space have been suggested. Major conclusions of investigations related to the design of space power systems are discussed. Attention is given to options for conversion cycles, the power system specific mass, and research and technology issues.

  11. Attrition reactor system

    SciTech Connect

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  12. Attrition reactor system

    SciTech Connect

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  13. Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

    SciTech Connect

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma; Al Rashdan, Ahmad; Tsvetkov, Pavel Valeryevich; Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

    2011-05-01

    This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

  14. The DOE Advanced Gas Reactor Fuel Development and Qualification Program

    SciTech Connect

    David Petti

    2010-09-01

    The high outlet temperatures and high thermal-energy conversion efficiency of modular High Temperature Gas-cooled Reactors (HTGRs) enable an efficient and cost effective integration of the reactor system with non-electricity generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300°C and 900°C. The Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission-product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete, fundamental understanding of the relationship between the fuel fabrication process and key fuel properties, the irradiation and accident safety performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. An overview of the program and recent progress is presented.

  15. Multi-channel gas-delivery system

    DOEpatents

    Rozenzon, Yan; Trujillo, Robert T.; Beese, Steven C.

    2016-09-13

    One embodiment of the present invention provides a gas-delivery system for delivering reaction gas to a reactor chamber. The gas-delivery system includes a main gas-inlet port for receiving reaction gases and a gas-delivery plate that includes a plurality of gas channels. A gas channel includes a plurality of gas holes for allowing the reaction gases to enter the reactor chamber from the gas channel. The gas-delivery system further includes a plurality of sub-gas lines coupling together the main gas-inlet port and the gas-delivery plate, and a respective sub-gas line is configured to deliver a portion of the received reaction gases to a corresponding gas channel.

  16. High pressure synthesis gas conversion. Task 1, System construction

    SciTech Connect

    Not Available

    1992-01-01

    A high pressure gas phase fermentation system has been constructed for the biological production of ethanol from coal synthesis gas. The reactors in the system consist of a 650 mL continuous stirred tank reactor and a 1 L continuous column reactor. The reactors are designed for individual or dual operation in series or parallel, with continuous gas and liquid feed. The system is housed in a constant temperature, explosion-proof room, equipped with gas leak detectors.

  17. FLOW SYSTEM FOR REACTOR

    DOEpatents

    Zinn, W.H.

    1963-06-11

    A reactor is designed with means for terminating the reaction when returning coolant is below a predetermined temperature. Coolant flowing from the reactor passes through a heat exchanger to a lower reservoir, and then circulates between the lower reservoir and an upper reservoir before being returned to the reactor. Means responsive to the temperature of the coolant in the return conduit terminate the chain reaction when the temperature reaches a predetermined minimum value. (AEC)

  18. Pressurized reactor system and a method of operating the same

    DOEpatents

    Isaksson, Juhani M.

    1996-01-01

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Superatmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gassification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor.

  19. Pressurized reactor system and a method of operating the same

    DOEpatents

    Isaksson, J.M.

    1996-06-18

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Super-atmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gasification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor. 2 figs.

  20. 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.

  1. NUCLEAR REACTOR FUEL SYSTEMS

    DOEpatents

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  2. Cooling molten salt reactors using "gas-lift"

    NASA Astrophysics Data System (ADS)

    Zitek, Pavel; Valenta, Vaclav; Klimko, Marek

    2014-08-01

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a "Two-phase flow demonstrator" (TFD) used for experimental study of the "gas-lift" system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for "gas-lift" (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

  3. Cooling molten salt reactors using “gas-lift”

    SciTech Connect

    Zitek, Pavel E-mail: klimko@kke.zcu.cz; Valenta, Vaclav E-mail: klimko@kke.zcu.cz; Klimko, Marek E-mail: klimko@kke.zcu.cz

    2014-08-06

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a “Two-phase flow demonstrator” (TFD) used for experimental study of the “gas-lift” system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for “gas-lift” (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

  4. Development of a system model for advanced small modular reactors.

    SciTech Connect

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  5. Gas-core reactor power transient analysis

    NASA Technical Reports Server (NTRS)

    Kascak, A. F.

    1972-01-01

    The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of this study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process.

  6. Advanced light water reactor requirements document: Chapter 4, Reactor systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the following for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR): reactor pressure vessel, nozzles and safe-ends, reactor internals, in-vessel portions of fluid systems (including reactor internal pumps (Emergency Core Cooling System (ECCS) piping and spargers), nuclear fuel, and the control rods and control rod drive system (including hydraulic supply and accumulators). Special tools required for reactor system maintenance, inspection and testing are also covered.

  7. Circulation in gas-slurry column reactors

    SciTech Connect

    Clark, N.; Kuhlman, J.; Celik, I.; Gross, R.; Nebiolo, E.; Wang, Yi-Zun.

    1990-08-15

    Circulation in bubble columns, such as those used in fischer-tropsch synthesis, detracts from their performance in that gas is carried on average more rapidly through the column, and the residence time distribution of the gas in the column is widened. Both of these factors influence mass-transfer operations in bubble columns. Circulation prediction and measurement has been undertaken using probes, one-dimensional models, laser Doppler velocimetry, and numerical modeling. Local void fraction was measured using resistance probes and a newly developed approach to determining air/water threshold voltage for the probe. A tall column of eight inch diameter was constructed of Plexiglas and the distributor plate was manufactured to distribute air evenly through the base of the column. Data were gathered throughout the volume at three different gas throughputs. Bubble velocities proved difficult to measure using twin probes with cross-correlation because of radial bubble movement. A series of three-dimensional mean and RMS bubble and liquid velocity measurements were also obtained for a turbulent flow in a laboratory model of a bubble column. These measurements have been made using a three-component laser Doppler velocimeter (LDV), to determine velocity distributions non-intrusively. Finally, the gas-liquid flow inside a vertically situated circular isothermal column reactor was simulated numerically. 74 refs., 170 figs., 5 tabs.

  8. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    NASA Astrophysics Data System (ADS)

    Harto, Andang Widi

    2012-06-01

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  9. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    SciTech Connect

    Harto, Andang Widi

    2012-06-06

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  10. SYSTEM FOR UNLOADING REACTORS

    DOEpatents

    Rand, A.C. Jr.

    1961-05-01

    An unloading device for individual vertical fuel channels in a nuclear reactor is shown. The channels are arranged in parallel rows and underneath each is a separate supporting block on which the fuel in the channel rests. The blocks are raounted in contiguous rows on an array of parallel pairs of tracks over the bottom of the reactor. Oblong hollows in the blocks form a continuous passageway through the middle of the row of blocks on each pair of tracks. At the end of each passageway is a horizontal grappling rod with a T- or L extension at the end next to the reactor of a length to permit it to pass through the oblong passageway in one position, but when rotated ninety degrees the head will strike one of the longer sides of the oblong hollow of one of the blocks. The grappling rod is actuated by a controllable reciprocating and rotating device which extends it beyond any individual block desired, rotates it and retracts it far enough to permit the fuel in the vertical channel above the block to fall into a handling tank below the reactor.

  11. Process Intensification with Integrated Water-Gas-Shift Membrane Reactor

    SciTech Connect

    2009-11-01

    This factsheet describes a research project whose objective is to develop hydrogen-selective membranes for an innovative gas-separation process based on a water-gas-shift membrane reactor (WGS-MR) for the production of hydrogen.

  12. Hot-Gas Filter Testing with a Transport Reactor Gasifier

    SciTech Connect

    Swanson, M.L.; Hajicek, D.R.

    2002-09-18

    Today, coal supplies over 55% of the electricity consumed in the United States and will continue to do so well into the next century. One of the technologies being developed for advanced electric power generation is an integrated gasification combined cycle (IGCC) system that converts coal to a combustible gas, cleans the gas of pollutants, and combusts the gas in a gas turbine to generate electricity. The hot exhaust from the gas turbine is used to produce steam to generate more electricity from a steam turbine cycle. The utilization of advanced hot-gas particulate and sulfur control technologies together with the combined power generation cycles make IGCC one of the cleanest and most efficient ways available to generate electric power from coal. One of the strategic objectives for U.S. Department of Energy (DOE) IGCC research and development program is to develop and demonstrate advanced gasifiers and second-generation IGCC systems. Another objective is to develop advanced hot-gas cleanup and trace contaminant control technologies. One of the more recent gasification concepts to be investigated is that of the transport reactor gasifier, which functions as a circulating fluid-bed gasifier while operating in the pneumatic transport regime of solid particle flow. This gasifier concept provides excellent solid-gas contacting of relatively small particles to promote high gasification rates and also provides the highest coal throughput per unit cross-sectional area of any other gasifier, thereby reducing capital cost of the gasification island.

  13. High Temperature Gas-Cooled Test Reactor Options Status Report

    SciTech Connect

    Sterbentz, James William; Bayless, Paul David

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  14. Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2004-10-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations.

  15. TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

    SciTech Connect

    Grover, S.B.

    2004-10-06

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations.

  16. Evaluation of the Gas Turbine Modular Helium Reactor

    SciTech Connect

    Not Available

    1994-02-01

    Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs.

  17. Staged membrane oxidation reactor system

    SciTech Connect

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2013-04-16

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  18. Staged membrane oxidation reactor system

    SciTech Connect

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2014-05-20

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  19. Staged membrane oxidation reactor system

    DOEpatents

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2012-09-11

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  20. Applications of plasma core reactors to terrestrial energy systems

    NASA Technical Reports Server (NTRS)

    Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

    1974-01-01

    Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

  1. Reactor refueling containment system

    DOEpatents

    Gillett, James E.; Meuschke, Robert E.

    1995-01-01

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  2. Reactor refueling containment system

    DOEpatents

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  3. Reactor core isolation cooling system

    DOEpatents

    Cooke, F.E.

    1992-12-08

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

  4. Reactor core isolation cooling system

    DOEpatents

    Cooke, Franklin E.

    1992-01-01

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

  5. Plasma reactor waste management systems

    NASA Technical Reports Server (NTRS)

    Ness, Robert O., Jr.; Rindt, John R.; Ness, Sumitra R.

    1992-01-01

    The University of North Dakota is developing a plasma reactor system for use in closed-loop processing that includes biological, materials, manufacturing, and waste processing. Direct-current, high-frequency, or microwave discharges will be used to produce plasmas for the treatment of materials. The plasma reactors offer several advantages over other systems, including low operating temperatures, low operating pressures, mechanical simplicity, and relatively safe operation. Human fecal material, sunflowers, oats, soybeans, and plastic were oxidized in a batch plasma reactor. Over 98 percent of the organic material was converted to gaseous products. The solids were then analyzed and a large amount of water and acid-soluble materials were detected. These materials could possibly be used as nutrients for biological systems.

  6. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

    Howard, D.F.; Motta, E.E.

    1961-06-27

    A method for controlling the excess reactivity in a nuclear reactor throughout the core life while maintaining the neutron flux distribution at the desired level is described. The control unit embodies a container having two electrodes of different surface area immersed in an electrolytic solution of a good neutron sbsorbing metal ion such as boron, gadolinium, or cadmium. Initially, the neutron absorber is plated on the larger electrode to control the greater neutron flux of a freshly refueled core. As the fuel burns up, the excess reactivity decreases and the neutron absorber is then plated onto the smaller electrode so that the number of neutrons absorbed also decreases. The excess reactivity in the core may thus be maintained without the introduction of serious perturbations in the neutron flux distributibn.

  7. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, R.M.

    1983-11-08

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

  8. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, Robert M.

    1983-01-01

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

  9. Reactor control rod timing system

    DOEpatents

    Wu, Peter T. K.

    1982-01-01

    A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  10. Reactor control rod timing system

    SciTech Connect

    Wu, P.T.

    1982-02-09

    A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (Above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  11. Fuel Development For Gas-Cooled Fast Reactors

    SciTech Connect

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  12. Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    NASA Technical Reports Server (NTRS)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

    2010-01-01

    Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

  13. Overview of environmental control aspects for the gas-cooled fast reactor

    SciTech Connect

    Nolan, A.M.

    1981-05-01

    Environmental control aspects relating to release of radionuclides have been analyzed for the Gas-Cooled Fast Reactor (GCFR). Information on environmental control systems was obtained for the most recent GCFR designs, and was used to evaluate the adequacy of these systems. The GCFR has been designed by the General Atomic Company as an alternative to other fast breeder reactor designs, such as the Liquid Metal Fast Breeder Reactor (LMFBR). The GCFR design includes mixed oxide fuel and helium coolant. The environmental impact of expected radionuclide releases from normal operation of the GCFR was evaluated using estimated collective dose equivalent commitments resulting from 1 year of plant operation. The results were compared to equivalent estimates for the Light Water Reactor (LWR) and High-Temperature Gas-Cooled Reactor (HTGR). A discussion of uncertainties in system performances, tritium production rates, and radiation quality factors for tritium is included.

  14. High temperature gas-cooled reactor: gas turbine application study

    SciTech Connect

    Not Available

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  15. TRITIUM PERMEATION AND TRANSPORT IN THE GASOLINE PRODUCTION SYSTEM COUPLED WITH HIGH TEMPERATURE GAS-COOLED REACTORS (HTGRS)

    SciTech Connect

    Chang H. Oh; Eung S. Kim; Mike Patterson

    2011-05-01

    This paper describes scoping analyses on tritium behaviors in the HTGR-integrated gasoline production system, which is based on a methanol-to-gasoline (MTG) plant. In this system, the HTGR transfers heat and electricity to the MTG system. This system was analyzed using the TPAC code, which was recently developed by Idaho National Laboratory. The global sensitivity analyses were performed to understand and characterize tritium behaviors in the coupled HTGR/MTG system. This Monte Carlo based random sampling method was used to evaluate maximum 17,408 numbers of samples with different input values. According to the analyses, the average tritium concentration in the product gasoline is about 3.05×10-3 Bq/cm3, and 62 % cases are within the tritium effluent limit (= 3.7x10-3 Bq/cm3[STP]). About 0.19% of released tritium is finally transported from the core to the gasoline product through permeations. This study also identified that the following four parameters are important concerning tritium behaviors in the HTGR/MTG system: (1) tritium source, (2) wall thickness of process heat exchanger, (3) operating temperature, and (4) tritium permeation coefficient of process heat exchanger. These four parameters contribute about 95 % of the total output uncertainties. This study strongly recommends focusing our future research on these four parameters to improve modeling accuracy and to mitigate tritium permeation into the gasol ine product. If the permeation barrier is included in the future study, the tritium concentration will be significantly reduced.

  16. Computer study of emergency shutdowns of a 60-kilowatt reactor Brayton space power system

    NASA Technical Reports Server (NTRS)

    Tew, R. C.; Jefferies, K. S.

    1974-01-01

    A digital computer study of emergency shutdowns of a 60-kWe reactor Brayton power system was conducted. Malfunctions considered were (1) loss of reactor coolant flow, (2) loss of Brayton system gas flow, (3)turbine overspeed, and (4) a reactivity insertion error. Loss of reactor coolant flow was the most serious malfunction for the reactor. Methods for moderating the reactor transients due to this malfunction are considered.

  17. System and method for air temperature control in an oxygen transport membrane based reactor

    DOEpatents

    Kelly, Sean M

    2016-09-27

    A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  18. Nuclear reactor insulation and preheat system

    DOEpatents

    Wampole, Nevin C.

    1978-01-01

    An insulation and preheat system for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the compartment. An external surface of the compartment or enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair.

  19. Proposal of Space Reactor for Nuclear Electric Propulsion System

    NASA Astrophysics Data System (ADS)

    Nagata, Hidetaka; Nishiyama, Takaaki; Nakashima, Hideki

    Currently, the solar battery, the chemical cell, and the RI battery are used for the energy source in space. However, it is difficult for them to satisfy requirements for deep space explorations. Therefore, other electric power sources which can stably produce high electric energy output, regardless of distance from the sun, are necessary to execute such missions. Then, we here propose small nuclear reactors as power sources for deep space exploration, and consider a conceptual design of a small nuclear reactor for Nuclear Electric Propulsion System. It is found from nuclear analyses that the Gas-Cooled reactor could not meet the design requirement imposed on the core mass. On the other hand, a light water reactor is found to be a promising alternative to the Gas-Cooled reactor.

  20. Gas release driven dynamics in research reactors piping

    SciTech Connect

    Kolev, Nikolay Ivanov; Roloff-Bock, Iris; Schlicht, Gerhard

    2006-07-01

    Analysis of the physical and chemical processes of radiolysis gas production, air absorption, diffusion controlled gas release and transport in the coolant cleaning system of the research reactor FRM II, which is now being in routine power operation in Munich, Germany, lead to the following conclusions: 1) The steady state pressure distribution in the siphon pipe allows that the horizontal part of the siphon pipe is filled with air. The air is isolated by about 1 m water column from the main pipe of the coolant cleaning system (CCS). This is a stable steady state. It has two positive impacts on the normal operation of the CCS: (a) there is effectively no bypass flow; (b) The air can not be transported through the pipe and therefore no deterioration of the pump performance is expected from the function of the siphon pipe. 2) Radiolysis gas production for coolant, that initially does not contain dissolved air, does not lead to any problem for the system. The gases are dissolved in the coolant at 2.2 bar and are not released for pressures reduction to about 1 bar, which is the minimum pressure in the CCS. 3) Assuming hypothetically a radiolysis gas production for coolant, which initially does contain dissolved air close to its saturation, leads to gas slug formation and its transport up to the pump. This could reduce the pump head and could lead to distortion of the normal operation. Systematic measurement of the hydrogen in the primary system at 100% power indicated, that this state is not realized in the system. The observed H{sub 2} concentration was between 0.016 e-6 and 0.380 e-6 which is of no concern at all. (authors)

  1. Reactor vessel annealing system

    DOEpatents

    Miller, Phillip E.; Katz, Leonoard R.; Nath, Raymond J.; Blaushild, Ronald M.; Tatch, Michael D.; Kordalski, Frank J.; Wykstra, Donald T.; Kavalkovich, William M.

    1991-01-01

    A system for annealing a vessel (14) in situ by heating the vessel (14) to a defined temperature, composed of: an electrically operated heater assembly (10) insertable into the vessel (14) for heating the vessel (14) to the defined temperature; temperature monitoring components positioned relative to the heater assembly (10) for monitoring the temperature of the vessel (14); a controllable electric power supply unit (32-60) for supplying electric power required by the heater assembly (10); a control unit (80-86) for controlling the power supplied by the power supply unit (32-60); a first vehicle (2) containing the power supply unit (32-60); a second vehicle (4) containing the control unit (80-86); power conductors (18,22) connectable between the power supply unit (32-60) and the heater unit (10) for delivering the power supplied by the power supply unit (32-60) to the heater assembly (10); signal conductors (20,24) connectable between the temperature monitoring components and the control unit (80-86) for delivering temperature indicating signals from the temperature monitoring components to the control unit (80-86); and control conductors (8) connectable between the control unit (80-86) and the power supply unit (32-60) for delivering to the power supply unit (32-60) control signals for controlling the level of power supplied by the power supply unit (32-60) to the heater assembly (10).

  2. POWER GENERATING NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Vernon, H.C.

    1958-03-01

    This patent relates to reactor systems of the type wherein the cooiing medium is a liquid which is converted by the heat of the reaction to steam which is conveyed directly to a pnime mover such as a steam turbine driving a generatore after which it is condensed and returred to the coolant circuit. In this design, the reactor core is disposed within a tank for containing either a slurry type fuel or an aggregation of solid fuel elements such as elongated rods submerged in a liquid moderator such as heavy water. The top of the tank is provided with a nozzle which extends into an expansion chamber connected with the upper end of the tank, the coolant being maintained in the expansion chamber at a level above the nozzle and the steam being formed in the expansion chamber.

  3. Fast breeder reactor protection system

    DOEpatents

    van Erp, J.B.

    1973-10-01

    Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

  4. Liquid metal cooled nuclear reactor plant system

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  5. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    SciTech Connect

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  6. Thermal reactor. [liquid silicon production from silane gas

    NASA Technical Reports Server (NTRS)

    Levin, H.; Ford, L. B. (Inventor)

    1982-01-01

    A thermal reactor apparatus and method of pyrolyticaly decomposing silane gas into liquid silicon product and hydrogen by-product gas is disclosed. The thermal reactor has a reaction chamber which is heated well above the decomposition temperature of silane. An injector probe introduces the silane gas tangentially into the reaction chamber to form a first, outer, forwardly moving vortex containing the liquid silicon product and a second, inner, rewardly moving vortex containing the by-product hydrogen gas. The liquid silicon in the first outer vortex deposits onto the interior walls of the reaction chamber to form an equilibrium skull layer which flows to the forward or bottom end of the reaction chamber where it is removed. The by-product hydrogen gas in the second inner vortex is removed from the top or rear of the reaction chamber by a vortex finder. The injector probe which introduces the silane gas into the reaction chamber is continually cooled by a cooling jacket.

  7. Solar coal gasification reactor with pyrolysis gas recycle

    DOEpatents

    Aiman, William R.; Gregg, David W.

    1983-01-01

    Coal (or other carbonaceous matter, such as biomass) is converted into a duct gas that is substantially free from hydrocarbons. The coal is fed into a solar reactor (10), and solar energy (20) is directed into the reactor onto coal char, creating a gasification front (16) and a pyrolysis front (12). A gasification zone (32) is produced well above the coal level within the reactor. A pyrolysis zone (34) is produced immediately above the coal level. Steam (18), injected into the reactor adjacent to the gasification zone (32), reacts with char to generate product gases. Solar energy supplies the energy for the endothermic steam-char reaction. The hot product gases (38) flow from the gasification zone (32) to the pyrolysis zone (34) to generate hot char. Gases (38) are withdrawn from the pyrolysis zone (34) and reinjected into the region of the reactor adjacent the gasification zone (32). This eliminates hydrocarbons in the gas by steam reformation on the hot char. The product gas (14) is withdrawn from a region of the reactor between the gasification zone (32) and the pyrolysis zone (34). The product gas will be free of tar and other hydrocarbons, and thus be suitable for use in many processes.

  8. A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

    NASA Astrophysics Data System (ADS)

    Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Peters, Curtis D.; Brown, Nicholas; Williamson, Joshua; Jablonski, Jennifer

    2005-02-01

    A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

  9. A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

    SciTech Connect

    Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Williamson, Joshua; Peters, Curtis D.; Brown, Nicholas; Jablonski, Jennifer

    2005-02-06

    A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

  10. Kinetics of hot-gas desulfurization sorbents for transport reactors

    SciTech Connect

    K.C. Kwon

    2000-01-01

    Hot-gas desulfurization for the integrated gasification combined cycle (IGCC) process has been investigated by many researchers to remove effectively hydrogen sulfide with various metal oxide sorbents at elevated temperatures. Various metal oxide sorbents are formulated with metal oxides such as Fe, Co, Zn, and Ti. Initial reaction kinetics of formulated sorbents with hydrogen sulfide is studied in the presence of various amounts of moisture and hydrogen at various reaction temperatures. The objectives of this research are to study initial reaction kinetics for a sorbent-hydrogen sulfide heterogeneous reaction system, to investigate effects of concentrations of hydrogen sulfide, hydrogen, and moisture on dynamic absorption of H{sub 2}S into sorbents, to understand effects of space time of reaction gas mixtures on initial reaction kinetics of the sorbent-hydrogen sulfide system, and to evaluate effects of temperature and sorbent amounts on dynamic absorption of H{sub 2}S into sorbents. Experimental data on initial reaction kinetics of hydrogen sulfide with metal oxide sorbents were obtained with a 0.83-cm{sup 3} differential reactor. The reactivity of MCRH-67 sorbent and AHI-1 was examined. These sorbents were obtained from the Research Triangle Institute (RTI). The sorbents in the form of 70 {micro}m particles are reacted with 1,000--4,000 ppm hydrogen sulfide at 450--600 C. The range of space time of reaction gas mixtures is 0.03--0.09 s. The range of reaction duration is 4--14,400 s.

  11. An improved external recycle reactor for determining gas-solid reaction kinetics

    NASA Technical Reports Server (NTRS)

    Miller, Irvin M.; Hoyt, Ronald F.

    1987-01-01

    These improvements in the recycle system effectively eliminate initial concentration variation by two modifications: (1) a vacuum line connection to the recycle loop which permits this loop to be evacuated and then filled with the test gas mixture to slightly above atmospheric pressure; and (2) a bypass line across the reactor which permits the reactor to be held under vacuum while the rest of the recycle loop is filled with test gas. A three-step procedure for bringing the feed gas mixture into contact with the catalyst at time zero is described.

  12. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  13. Multiphysics methods development for high temperature gas reactor analysis

    NASA Astrophysics Data System (ADS)

    Seker, Volkan

    Multiphysics computational methods were developed to perform design and safety analysis of the next generation Pebble Bed High Temperature Gas Cooled Reactors. A suite of code modules was developed to solve the coupled thermal-hydraulics and neutronics field equations. The thermal-hydraulics module is based on the three dimensional solution of the mass, momentum and energy equations in cylindrical coordinates within the framework of the porous media method. The neutronics module is a part of the PARCS (Purdue Advanced Reactor Core Simulator) code and provides a fine mesh finite difference solution of the neutron diffusion equation in three dimensional cylindrical coordinates. Coupling of the two modules was performed by mapping the solution variables from one module to the other. Mapping is performed automatically in the code system by the use of a common material mesh in both modules. The standalone validation of the thermal-hydraulics module was performed with several cases of the SANA experiment and the standalone thermal-hydraulics exercise of the PBMR-400 benchmark problem. The standalone neutronics module was validated by performing the relevant exercises of the PBMR-268 and PBMR-400 benchmark problems. Additionally, the validation of the coupled code system was performed by analyzing several steady state and transient cases of the OECD/NEA PBMR-400 benchmark problem.

  14. CONTROL SYSTEM FOR NEUTRONIC REACTORS

    DOEpatents

    Crever, F.E.

    1962-05-01

    BS>A slow-acting shim rod for control of major variations in reactor neutron flux and a fast-acting control rod to correct minor flux variations are employed to provide a sensitive, accurate control system. The fast-acting rod is responsive to an error signal which is produced by changes in the neutron flux from a predetermined optimum level. When the fast rod is thus actuated in a given direction, means is provided to actuate the slow-moving rod in that direction to return the fast rod to a position near the midpoint of its control range. (AEC)

  15. A thermodynamic approach for advanced fuels of gas-cooled reactors

    NASA Astrophysics Data System (ADS)

    Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

    2005-09-01

    For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

  16. Shutdown system for a nuclear reactor

    DOEpatents

    Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

    1984-06-05

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

  17. Shutdown system for a nuclear reactor

    DOEpatents

    Groh, Edward F.; Olson, Arne P.; Wade, David C.; Robinson, Bryan W.

    1984-01-01

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

  18. Designing a Gas Test Loop for the Advanced Test Reactor

    SciTech Connect

    James R. Parry

    2005-11-01

    The Generation IV Reactor Program and the Advanced Fuel Cycle Initiative are investigating some new reactor concepts which require extensive materials and fuels testing in a fast neutron spectrum. The capability to test materials and fuels in a fast neutron flux in the United States is very limited to non-existent. It has been proposed to install a gas test loop (GTL) in one of the lobes of the Advanced Test Reactor (ATR) at the Idaho National Laboratory and harden the spectrum to provide some fast neutron flux testing capabilities in the United States. This paper describes the neutronics investigation into the design of the GTL for the ATR.

  19. Method and apparatus for enhancing reactor air-cooling system performance

    DOEpatents

    Hunsbedt, Anstein

    1996-01-01

    An enhanced decay heat removal system for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer.

  20. Method and apparatus for enhancing reactor air-cooling system performance

    DOEpatents

    Hunsbedt, A.

    1996-03-12

    An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

  1. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2008-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  2. A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative)

    NASA Astrophysics Data System (ADS)

    Marshall, A. C.

    1989-08-01

    A review was completed of multimegawatt gas cooled reactor concepts proposed for SDI applications. The study concluded that the principal reason for considering gas cooled reactors for burst mode operation was the potential for significant system mass savings over closed cycle systems if open cycle gas cooled operation (effluent exhausted to space) is acceptable. The principal reason for considering gas cooled reactors for steady state operation is that they may represent a lower technology risk than other approaches. In the review, nine gas cooled reactor concepts were compared to identify the most promising. For burst mode operation, the NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor concept emerged as a strong first choice since its performance exceeds the anticipated operational requirements and the technology was demonstrated and is retrievable. Although the NERVA derivative concepts were determined to be the lead candidates for the Multimegawatt Steady State (MMWSS) mode as well, their lead over the other candidates is not as great as for the burst mode.

  3. A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative) applications

    SciTech Connect

    Marshall, A.C.

    1989-08-01

    We have completed a review of multimegawatt gas-cooled reactor concepts proposed for SDI applications. Our study concluded that the principal reason for considering gas-cooled reactors for burst-mode operation was the potential for significant system mass savings over closed-cycle systems if open-cycle gas-cooled operation (effluent exhausted to space) is acceptable. The principal reason for considering gas-cooled reactors for steady-state operation is that they may represent a lower technology risk than other approaches. In the review, nine gas-cooled reactor concepts were compared to identify the most promising. For burst-mode operation, the NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor concept emerged as a strong first choice since its performance exceeds the anticipated operational requirements and the technology has been demonstrated and is retrievable. Although the NERVA derivative concepts were determined to be the lead candidates for the Multimegawatt Steady-State (MMWSS) mode as well, their lead over the other candidates is not as great as for the burst mode. 90 refs., 2 figs., 10 tabs.

  4. Inertial confinement fusion reactor systems

    SciTech Connect

    Frank, T.G.; Bohachevsky, I.O.; Pendergrass, J.H.

    1980-01-01

    A variety of reactor cavity concepts, drivers, and energy conversion mechanisms are being considered to realize commercial applications of ICF. Presented in this paper are: (1) a review of reactor concepts with estimates of practically achievable pulse repetition rates; (2) a survey of drivers with estimates of the requirements on reactor conditions imposed by beam propagation characteristics; and (3) an assessment of compatible driver-reactor combinations.

  5. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    SciTech Connect

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  6. A 50-100 kWe gas-cooled reactor for use on Mars.

    SciTech Connect

    Peters, Curtis D.

    2006-04-01

    In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor's thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor's safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered.

  7. Tandem Mirror Reactor Systems Code (Version I)

    SciTech Connect

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.

  8. Optimization of gas hydrate reactors with slug flow

    SciTech Connect

    Elperin, T.; Fominykh, A.

    1997-11-01

    A model of heat transfer during gas hydrate formation at a gas-liquid interface in gas-liquid slug flow with liquid plugs containing small bubbles is suggested. Under the assumption of perfect mixing of liquid in liquid plugs, recurrent relations for temperature in the n-th liquid plug and heat and mass fluxes from the n-th unit cell in a gas-liquid slug flow are derived. The ratio of the total mass flux during gas hydrate formation in a cluster with N unit cells to the mass flux in a cluster with an infinite number of unit cells is determined. The number of unit cells that yield 95% of the total amount of gas hydrates in an infinite cluster of unit cells is calculated and formula for an optimal length of a gas hydrate slug flow reactor is derived.

  9. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    SciTech Connect

    Shropshire, D.E.; Herring, J.S.

    2004-10-03

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  10. Lithium Ceramic Blankets for Russian Fusion Reactors and Influence of Breeding Operation Mode on Parameters of Reactor Tritium Systems

    SciTech Connect

    Kapyshev, Victor K.; Chernetsov, Mikhail Yu.; Zhevotov, Sergej I.; Kersnovskij, Alexandr Yu.; Kolbasov, Boris N.; Kovalenko, Victor G.; Paltusov, Nikolaj P.; Sernyaev, Georgeij A.; Sterebkov, Juri S.; Zyryanov, Alexej P.

    2005-07-15

    Russian controlled fusion program supposes development of a DEMO reactor design and participation in ITER Project. A solid breeder blanket of DEMO contains a ceramic lithium orthosilicate breeder and a beryllium multiplier. Test modules of the blanket are developed within the scope of ITER activities. Experimental models of module tritium breeding zones (TBZ), materials and fabrication technology of the TBZ, tritium reactor systems to analyse and process gas released from lithium ceramics are being developed. Two models of tritium breeding and neutron multiplying elements of the TBZ have been designed, manufactured and tested in IVV-2M nuclear reactor. Initial results of the in-pile experiments and outcome of lithium ceramics irradiation in a water-graphite nuclear reactor are considered to be a data base for development of the test modules and initial requirements for DEMO tritium system design. Influence of the tritium release parameters and hydrogen concentration in a purge gas on parameters of reactor system are discussed.

  11. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    SciTech Connect

    Not Available

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  12. High-Temperature Water-Gas Shift Membrane Reactor Study

    SciTech Connect

    Ciocco, M.V.; Iyoha, O.; Enick, R.M.; Killmeyer, R.P.

    2007-06-01

    NETL’s Office of Research and Development is exploring the integration of membrane reactors into coal gasification plants as a way of increasing efficiency and reducing costs. Water-Gas Shift Reaction experiments were conducted in membrane reactors at conditions similar to those encountered at the outlet of a coal gasifier. The changes in reactant conversion and product selectivity due to the removal of hydrogen via the membrane reactor were quantified. Research was conducted to determine the influence of residence time and H2S on CO conversion in both Pd and Pd80wt%Cu membrane reactors. Effects of the hydrogen sulfide-to-hydrogen ratio on palladium and a palladium-copper alloy at high-temperature were also investigated. These results were compared to thermodynamic calculations for the stability of palladium sulfides.

  13. Preparation of high temperature gas-cooled reactor fuel element

    DOEpatents

    Bradley, Ronnie A.; Sease, John D.

    1976-01-01

    This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

  14. Thermally Simulated Testing of a Direct-Drive Gas-Cooled Nuclear Reactor

    NASA Technical Reports Server (NTRS)

    Godfroy, Thomas; Bragg-Sitton, Shannon; VanDyke, Melissa

    2003-01-01

    This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrical thermal simulation of reactor components and concepts.

  15. Thermal Reactor Code System for Reactor Design and Analysis.

    2003-04-21

    Version: 00 SRAC95 is a general purpose neutronics code system applicable to core analyses of various types of reactors, including cell calculation with burn up, core calculation for any type of thermal reactor; where core burn up calculation and fuel management were done by an auxiliary code. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications were made for nuclear data libraries and programs. In this version,more » many new functions and data are implemented to support nuclear design studies of advanced reactors. SRAC95 can be used for burnup credit analysis within the ORIGEN2 and SWAT (CCC-714) code system.« less

  16. Development of Novel Water-Gas Shift Membrane Reactor

    SciTech Connect

    Ho, W. S. Winston

    2004-12-29

    This report summarizes the objectives, technical barrier, approach, and accomplishments for the development of a novel water-gas-shift (WGS) membrane reactor for hydrogen enhancement and CO reduction. We have synthesized novel CO{sub 2}-selective membranes with high CO{sub 2} permeabilities and high CO{sub 2}/H{sub 2} and CO{sub 2}/CO selectivities by incorporating amino groups in polymer networks. We have also developed a one-dimensional non-isothermal model for the countercurrent WGS membrane reactor. The modeling results have shown that H{sub 2} enhancement (>99.6% H{sub 2} for the steam reforming of methane and >54% H{sub 2} for the autothermal reforming of gasoline with air on a dry basis) via CO{sub 2} removal and CO reduction to 10 ppm or lower are achievable for synthesis gases. With this model, we have elucidated the effects of system parameters, including CO{sub 2}/H{sub 2} selectivity, CO{sub 2} permeability, sweep/feed flow rate ratio, feed temperature, sweep temperature, feed pressure, catalyst activity, and feed CO concentration, on the membrane reactor performance. Based on the modeling study using the membrane data obtained, we showed the feasibility of achieving H{sub 2} enhancement via CO{sub 2} removal, CO reduction to {le} 10 ppm, and high H{sub 2} recovery. Using the membrane synthesized, we have obtained <10 ppm CO in the H{sub 2} product in WGS membrane reactor experiments. From the experiments, we verified the model developed. In addition, we removed CO{sub 2} from a syngas containing 17% CO{sub 2} to about 30 ppm. The CO{sub 2} removal data agreed well with the model developed. The syngas with about 0.1% CO{sub 2} and 1% CO was processed to convert the carbon oxides to methane via methanation to obtain <5 ppm CO in the H{sub 2} product.

  17. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    SciTech Connect

    Lee O. Nelson

    2011-04-01

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540°C and the helium coolant was delivered at 7 MPa at 625–925°C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the hightemperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

  18. Nuclear electric propulsion reactor control systems status

    NASA Technical Reports Server (NTRS)

    Ferg, D. A.

    1973-01-01

    The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

  19. Fluid sampling system for a nuclear reactor

    DOEpatents

    Lau, L.K.; Alper, N.I.

    1994-11-22

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump. 1 fig.

  20. Fluid sampling system for a nuclear reactor

    DOEpatents

    Lau, Louis K.; Alper, Naum I.

    1994-01-01

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump.

  1. Development of Tritium Permeation Analysis Code and Tritium Transport in a High Temperature Gas-Cooled Reactor Coupled with Hydrogen Production System

    SciTech Connect

    Chang H. Oh; Eung S. Kim; Mike Patterson

    2010-06-01

    Abstract – A tritium permeation analyses code (TPAC) was developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in very high temperature reactor (VHTR) systems, including integrated hydrogen production systems. A MATLAB SIMULINK software package was used in developing the code. The TPAC is based on the mass balance equations of tritium-containing species and various forms of hydrogen coupled with a variety of tritium sources, sinks, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, and 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of tritium and H2 through pipes, vessels, and heat exchangers were considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems, including high temperature electrolysis and sulfur-iodine processes.

  2. Advanced light water reactor requirements document: Chapter 3, Reactor coolant system and reactor non-safety auxiliary systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Coolant System and the Reactor Non-safety Auxiliary Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the reactor coolant system and reactor non-safety auxiliary systems for Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Non-safety auxiliaries include systems which are required for normal operation of the plant but are not required to operate for accident mitigation or to bring the plant to a safe shutdown condition. For PWRs, the reactor coolant system, steam generator system, chemical and volume control system and boron recycle system are included. For BWRs, the reactor coolant system and reactor water cleanup system are included. The chapter also includes requirements for the above systems which are common to BWRs and PWRs and requirements for process sampling for BWRs and PWRs.

  3. Feasibility study of full-reactor gas core demonstration test

    NASA Technical Reports Server (NTRS)

    Kunze, J. F.; Lofthouse, J. H.; Shaffer, C. J.; Macbeth, P. J.

    1973-01-01

    Separate studies of nuclear criticality, flow patterns, and thermodynamics for the gas core reactor concept have all given positive indications of its feasibility. However, before serious design for a full scale gas core application can be made, feasibility must be shown for operation with full interaction of the nuclear, thermal, and hydraulic effects. A minimum sized, and hence minimum expense, test arrangement is considered for a full gas core configuration. It is shown that the hydrogen coolant scattering effects dominate the nuclear considerations at elevated temperatures. A cavity diameter of somewhat larger than 4 ft (122 cm) will be needed if temperatures high enough to vaporize uranium are to be achieved.

  4. Fission control system for nuclear reactor

    DOEpatents

    Conley, G.H.; Estes, G.P.

    Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

  5. The 5-kwe reactor thermoelectric system summary

    NASA Technical Reports Server (NTRS)

    Vanosdol, J. H. (Editor)

    1973-01-01

    Design of the 5-kwe reactor thermoelectric system was initiated in February 1972 and extended through the conceptual design phase into the preliminary design phase. Design effort was terminated in January, 1973. This report documents the system and component requirements, design approaches, and performance and design characteristics for the 5-kwe system. Included is summary information on the reactor, radiation shields, power conversion systems, thermoelectric pump, radiator/structure, liquid metal components, and the control system.

  6. Comments on the feasibility of developing gas core nuclear reactors. [for manned interplanetary spacecraft propulsion

    NASA Technical Reports Server (NTRS)

    Rom, F. E.

    1969-01-01

    Recent developments in the fields of gas core hydrodynamics, heat transfer, and neutronics indicate that gas core nuclear rockets may be feasible from the point of view of basic principles. Based on performance predictions using these results, mission analyses indicate that gas core nuclear rockets may have the potential for reducing the initial weight in orbit of manned interplanetary vehicles by a factor of 5 when compared to the best chemical rocket systems. In addition, there is a potential for reducing total trip times from 450 to 500 days for chemical systems to 250 to 300 days for gas core systems. The possibility of demonstrating the feasibility of gas core nuclear rocket engines by means of a logical series of experiments of increasing difficulty that ends with ground tests of full scale gas core reactors is considered.

  7. Brayton Cycle for High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Oh, Chang H.; Moore, Richard L.

    2005-03-15

    This paper describes research on improving the Brayton cycle efficiency for a high-temperature gas-cooled reactor (HTGR). In this study, we are investigating the efficiency of an indirect helium Brayton cycle for the power conversion side of an HTGR power plant. A reference case based on a 250-MW(thermal) pebble bed HTGR was developed using helium gas as a working fluid in both the primary and power conversion sides. The commercial computer code HYSYS was used for process optimization. A numerical model using the Visual-Basic (V-B) computer language was also developed to assist in the evaluation of the Brayton cycle efficiency. Results from both the HYSYS simulation and the V-B model were compared with Japanese calculations based on the 300-MW(electric) Gas Turbine High-Temperature Reactor (GTHTR) that was developed by the Japan Atomic Energy Research Institute. After benchmarking our models, parametric investigations were performed to see the effect of important parameters on the cycle efficiency. We also investigated single-shaft versus multiple-shaft arrangements for the turbomachinery. The results from this study are applicable to other reactor concepts such as fast gas-cooled reactors, supercritical water reactors, and others.The ultimate goal of this study is to use other fluids such as supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency over that of the helium Brayton cycle. This study is in progress, and the results will be published in a subsequent paper.

  8. Brayton Cycle for High Temperature Gas-Cooled Reactors

    SciTech Connect

    Chang Oh

    2005-03-01

    This paper describes research on improving the Brayton cycle efficiency for a high-temperature gas-cooled reactor (HTGR). In this study, we are investigating the efficiency of an indirect helium Brayton cycle for the power conversion side of an HTGR power plant. A reference case based on a 250-MW(thermal) pebble bed HTGR was developed using helium gas as a working fluid in both the primary and power conversion sides. The commercial computer code HYSYS was used for process optimization. A numerical model using the Visual-Basic (V-B) computer language was also developed to assist in the evaluation of the Brayton cycle efficiency. Results from both the HYSYS simulation and the V-B model were compared with Japanese calculations based on the 300-MW(electric) Gas Turbine High-Temperature Reactor (GTHTR) that was developed by the Japan Atomic Energy Research Institute. After benchmarking our models, parametric investigations were performed to see the effect of important parameters on the cycle efficiency. We also investigated single-shaft versus multiple-shaft arrangements for the turbomachinery. The results from this study are applicable to other reactor concepts such as fast gas-cooled reactors, supercritical water reactors, and others. The ultimate goal of this study is to use other fluids such as supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency over that of the helium Brayton cycle. This study is in progress, and the results will be published in a subsequent paper.

  9. Gas venting system

    DOEpatents

    Khan, Amjad; Dreier, Ken Wayne; Moulthrop, Lawrence Clinton; White, Erik James

    2010-06-29

    A system to vent a moist gas stream is disclosed. The system includes an enclosure and an electrochemical cell disposed within the enclosure, the electrochemical cell productive of the moist gas stream. A first vent is in fluid communication with the electrochemical cell for venting the moist gas stream to an exterior of the enclosure, and a second vent is in fluid communication with an interior of the enclosure and in thermal communication with the first vent for discharging heated air to the exterior of the enclosure. At least a portion of the discharging heated air is for preventing freezing of the moist gas stream within the first vent.

  10. A combined gas cooled nuclear reactor and fuel cell cycle

    NASA Astrophysics Data System (ADS)

    Palmer, David J.

    Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

  11. Exhaust gas recirculation system

    SciTech Connect

    Minoura, M.; Yorioka, K.

    1980-11-18

    An exhaust gas recirculation system for cleaning exhaust gas from an internal combustion engine is provided in which a variable constriction is provided between an intake pipe and a pressure control valve in operative connection to a throttle valve in the carburetor and the pressure differential across said variable constriction is maintained constant to keep off any influence of the exhaust gas pressure while the ratio of the exhaust gas flow rate to the air intake into the engine is varied in correspondence to the intake pipe negative pressure. This exhaust gas recirculation system can be adapted to a fuel injection type intake system as well as other intake systems provided with an air valve for regulating air intake or having no venturi constriction such as employed in an su type carburetor.

  12. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    SciTech Connect

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  13. TREAT Reactor Control and Protection System

    SciTech Connect

    Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

    1985-01-01

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.

  14. Self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

    1988-01-01

    A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

  15. Improvement of Analysis Technology for High Temperature Gas-Cooled Reactor by Using Data Obtained in High Temperature Engineering Test Reactor

    NASA Astrophysics Data System (ADS)

    Nakagawa, Shigeaki; Tochio, Daisuke; Takamatsu, Kuniyoshi; Goto, Minoru; Takeda, Tetsuaki

    The Very High Temperature Reactor (VHTR) system, which is one of generation IV reactors, is the high temperature gas-cooled reactor (HTGR) with capabilities of hydrogen production and high efficiency electricity generation. The High Temperature Engineering Test Reactor (HTTR) is the first HTGR in Japan. The HTTR achieved full power of 30MW at a reactor outlet coolant temperature of about 950°C in April, 2004 during the “rise-to-power tests” confirming the reactor performance. The safety demonstration tests by using the HTTR started from 2002 and are under going to demonstrate inherent safety features of HTGRs. The experimental data obtained in these tests are inevitable to design the VHTR with high cost performance. The analytical models validated through these tests in the HTTR are applicable to precise simulation of an HTGR performance and can contribute to the research and development of the VHTR.

  16. Safety philosophy of gas turbine high temperature reactor (GTHTR300)

    SciTech Connect

    Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

    2002-07-01

    Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Major features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)

  17. An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

    NASA Technical Reports Server (NTRS)

    Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

    1988-01-01

    A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

  18. High Temperature Materials - The Challenge for Future Advanced Gas Cooled Reactors

    SciTech Connect

    Hoffelner, Wolfgang

    2004-07-01

    Advanced gas cooled reactor systems for future combined cycle applications (electricity and process heat) are planned to operate at temperatures above 1000 deg. C. The reliable and safe operation of such plants requires materials that are able to carry loads at these temperatures in impure helium and under neutron irradiation. The most exposed components are the pressure vessel, reactor internals, gas turbine, pipes and valves. Considering the envisaged long operating time (6 years for replaceable components) life time assessments and extrapolation methods are necessary for the determination of damage evolution and long term behaviour of the reactor components. This paper gives a summary of candidate materials and possible approaches to life-time assessment. (author)

  19. Advanced Water-Gas Shift Membrane Reactor

    SciTech Connect

    Sean Emerson; Thomas Vanderspurt; Susanne Opalka; Rakesh Radhakrishnan; Rhonda Willigan

    2009-01-07

    The overall objectives for this project were: (1) to identify a suitable PdCu tri-metallic alloy membrane with high stability and commercially relevant hydrogen permeation in the presence of trace amounts of carbon monoxide and sulfur; and (2) to identify and synthesize a water gas shift catalyst with a high operating life that is sulfur and chlorine tolerant at low concentrations of these impurities. This work successfully achieved the first project objective to identify a suitable PdCu tri-metallic alloy membrane composition, Pd{sub 0.47}Cu{sub 0.52}G5{sub 0.01}, that was selected based on atomistic and thermodynamic modeling alone. The second objective was partially successful in that catalysts were identified and evaluated that can withstand sulfur in high concentrations and at high pressures, but a long operating life was not achieved at the end of the project. From the limited durability testing it appears that the best catalyst, Pt-Re/Ce{sub 0.333}Zr{sub 0.333}E4{sub 0.333}O{sub 2}, is unable to maintain a long operating life at space velocities of 200,000 h{sup -1}. The reasons for the low durability do not appear to be related to the high concentrations of H{sub 2}S, but rather due to the high operating pressure and the influence the pressure has on the WGS reaction at this space velocity.

  20. SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

    DOEpatents

    Dickinson, R.W.

    1963-03-01

    This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

  1. Reactor system for olefin conversion and etherification

    SciTech Connect

    Harandi, M.N.

    1990-09-18

    This patent describes a reactor system. It is for producing liquid fuel mixtures from olefin-containing feedstock and lower alcohols by multistage etherification, olefin interconversion and oligomerization reactions.

  2. Completion of the first NGNP Advanced Gas Reactor Fuel Irradiation Experiment, AGR-1, in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover; John Maki; David Petti

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The design of AGR-1 test train and support systems used to monitor and control the experiment during

  3. Scanning tunneling microscope assembly, reactor, and system

    SciTech Connect

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  4. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  5. REACTOR CONTROL ROD OPERATING SYSTEM

    DOEpatents

    Miller, G.

    1961-12-12

    A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

  6. Test Results From a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    NASA Technical Reports Server (NTRS)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.

    2009-01-01

    The Brayton Power Conversion Unit (BPCU) located at NASA Glenn Research Center (GRC) in Cleveland, OH is a closed cycle system incorporating a turboaltemator, recuperator, and gas cooler connected by gas ducts to an external gas heater. For this series of tests, the BPCU was modified by replacing the gas heater with the Direct Drive Gas heater or DOG. The DOG uses electric resistance heaters to simulate a fast spectrum nuclear reactor similar to those proposed for space power applications. The combined system thermal transient behavior was the focus of these tests. The BPCU was operated at various steady state points. At each point it was subjected to transient changes involving shaft rotational speed or DOG electrical input. This paper outlines the changes made to the test unit and describes the testing that took place along with the test results.

  7. ORIGEN-ARP Cross-Section Libraries for Magnox, Advanced Gas-Cooled, and VVER Reactor Designs

    SciTech Connect

    Murphy, BD

    2004-03-10

    Cross-section libraries for the ORIGEN-ARP system were extended to include four non-U.S. reactor types: the Magnox reactor, the Advanced Gas-Cooled Reactor, the VVER-440, and the VVER-1000. Typical design and operational parameters for these four reactor types were determined by an examination of a variety of published information sources. Burnup simulation models of the reactors were then developed using the SAS2H sequence from the Oak Ridge National Laboratory SCALE code system. In turn, these models were used to prepare the burnup-dependent cross-section libraries suitable for use with ORIGEN-ARP. The reactor designs together with the development of the SAS2H models are described, and a small number of validation results using spent-fuel assay data are reported.

  8. 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.

  9. Closed Brayton cycle power conversion systems for nuclear reactors :

    SciTech Connect

    Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.; Sanchez, Travis

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

  10. Control system for a small fission reactor

    DOEpatents

    Burelbach, James P.; Kann, William J.; Saiveau, James G.

    1986-01-01

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired. In another embodiment, a plurality of flexible hollow tubes each containing a neutron absorber are positioned adjacent to one another in spaced relation around the periphery of the reactor vessel and inside the outer neutron reflector with reactivity controlled by the extension and compression of all or some of the coiled hollow tubes. Yet another embodiment of the invention envisions the neutron reflector in the form of an expandable coil spring positioned in an annular space between the reactor vessel and an outer neutron absorbing structure for controlling the neutron flux reflected back into the reactor vessel.

  11. Power Conversion Study for High Temperature Gas-Cooled Reactors

    SciTech Connect

    Chang Oh; Richard Moore; Robert Barner

    2005-05-01

    The Idaho National Laboratory (INL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. There are some technical issues to be resolved before the selection of the final design of the high temperature gascooled reactor, called as a Next Generation Nuclear Plant (NGNP), which is supposed to be built at the INEEL by year 2017. The technical issues are the selection of the working fluid, direct vs. indirect cycle, power cycle type, the optimized design in terms of a number of intercoolers, and others. In this paper, we investigated a number of working fluids for the power conversion loop, direct versus indirect cycle, the effect of intercoolers, and other thermal hydraulics issues. However, in this paper, we present part of the results we have obtained. HYSYS computer code was used along with a computer model developed using Visual Basic computer language.

  12. Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

    SciTech Connect

    Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y.

    2012-05-22

    There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

  13. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    SciTech Connect

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  14. CONUP: A code that calculates tag gas concentrations for reactor components

    SciTech Connect

    Hammervold, D.J.; Siciliano, E.R.; Schmittroth, F.; Schenter, R.E.

    1991-11-01

    The CONUP code calculates the current tag isotopic concentrations for the Tag Gas Analysis Code, SMTAG. The combined codes, CONUP and SMTAG, represent the Tag Gas System. CONUP produces tag concentrations that are decayed and transmuted over specific reactor core cycles. The calculated concentrations are used, together with measured concentrations, as input for the SMTAG code, which identifies the failed reactor components that have released tag gas. The CONUP code has two modes for calculating isotopic concentrations: absolute and incremental. In the absolute mode, the CONUP code calculates concentrations from the beginning of the reactor startup through the current cycle. In the incremental mode, the CONUP code processes concentrations from the last reactor component cycle for each component. The incremental mode saves significant processing time because the concentrations are updated only for the current cycle. A description of the underlying physical model and method of solution are presented. A description of the code and a user`s guide are also given, along with example input and corresponding concentration output.

  15. Summary of space nuclear reactor power systems, 1983 - 1992

    NASA Astrophysics Data System (ADS)

    Buden, D.

    1993-08-01

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987-88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  16. Control system for a small fission reactor

    DOEpatents

    Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

    1985-02-08

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

  17. 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.

  18. Gas dynamics modeling of the HYLIFE-II reactor

    SciTech Connect

    Jantzen, C.

    1995-08-01

    Gas dynamics in the IFE reactor, HYLIFE-II is modeled using the code, TSUNAMI. This code is a 2-D shock-solver that uses the Godunov method with operator splitting. Results from a cylindrically symmetric simulation indicate an initial, low density, burst of high energy particles enters the final focus transport lens within 40 microseconds after the blast, much faster than the proposed 1 millisecond shutter closing time. After approximately 100 microseconds the chamber debris flux levels off to one eighth its peak value and maintains this intensity until the shutter closes. Although initial protective jet ablation is considered, neither secondary radiation nor condensation are modeled. Therefore results are conservative.

  19. Statistical Methods Handbook for Advanced Gas Reactor Fuel Materials

    SciTech Connect

    J. J. Einerson

    2005-05-01

    Fuel materials such as kernels, coated particles, and compacts are being manufactured for experiments simulating service in the next generation of high temperature gas reactors. These must meet predefined acceptance specifications. Many tests are performed for quality assurance, and many of these correspond to criteria that must be met with specified confidence, based on random samples. This report describes the statistical methods to be used. The properties of the tests are discussed, including the risk of false acceptance, the risk of false rejection, and the assumption of normality. Methods for calculating sample sizes are also described.

  20. Hot gas cleanup for molten carbonate fuel cells. A zinc oxide reactor model, Final report

    SciTech Connect

    Steinfeld, G.

    1980-09-16

    Utilization of coal gasifiers to power MCFC requires a cleanup system to remove sulfur and particulates. Of the two near term options available for desulfurization of gasifier effluent, namely low temperature cleanup utilizing absorber/stripper technology, and hot gas cleanup utilizing metal oxides, there is a clear advantage to using hot gas cleanup. Since the MCFC will operate at 1200/sup 0/F, and the gasifier effluent could be between 1200 to 1900/sup 0/F, a hot gas cleanup system will require little or no change in process gas temperature, thereby contributing to a high overall system efficiency. A hot gas cleanup system will consist of FeO for bulk H/sub 2/S removal and ZnO for reduction of H/sub 2/S to sub ppM levels. Hot gas cleanup systems at present are not available commercially, and therefore it is the objective of this project to model the components of the system in order to help bring this technology closer to commercialization, by providing simulated operating characteristics to aid in system design, and system simulations of gasifier/MCFC systems. The modeling of the ZnO reactor is presented.

  1. Code System to Calculate Reactor Coolant System Leak Rate.

    1999-10-19

    Version 00 RCSLK9 was developed to analyze the leak tightness of the primary coolant system for any pressurized water reactor (PWR). From given system conditions, water levels in tanks, and certain system design parameters, RCSLK9 calculates the loss of water from the reactor coolant system (RCS) and the increase of water in the leakage collection system during an arbitrary time interval. The program determines the system leak rates and displays or prints a report ofmore » the results. During the initial application to a specific reactor, RCSLK9 creates a file of system parameters and saves it for future use.« less

  2. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  3. Overall plant design specification Modular High Temperature Gas-cooled Reactor. Revision 9

    SciTech Connect

    1990-05-01

    Revision 9 of the ``Overall Plant Design Specification Modular High Temperature Gas-Cooled Reactor,`` DOE-HTGR-86004 (OPDS) has been completed and is hereby distributed for use by the HTGR Program team members. This document, Revision 9 of the ``Overall Plant Design Specification`` (OPDS) reflects those changes in the MHTGR design requirements and configuration resulting form approved Design Change Proposals DCP BNI-003 and DCP BNI-004, involving the Nuclear Island Cooling and Spent Fuel Cooling Systems respectively.

  4. Novel online monitoring and alert system for anaerobic digestion reactors.

    PubMed

    Dong, Fang; Zhao, Quan-Bao; Li, Wen-Wei; Sheng, Guo-Ping; Zhao, Jin-Bao; Tang, Yong; Yu, Han-Qing; Kubota, Kengo; Li, Yu-You; Harada, Hideki

    2011-10-15

    Effective monitoring and diagnosis of anaerobic digestion processes is a great challenge for anaerobic digestion reactors, which limits their stable operation. In this work, an online monitoring and alert system for upflow anaerobic sludge blanket (UASB) reactors is developed on the basis of a set of novel evaluating indexes. The two indexes, i.e., stability index S and auxiliary index a, which incorporate both gas- and liquid-phase parameters for UASB, enable a quantitative and comprehensive evaluation of reactor status. A series of shock tests is conducted to evaluate the response of the monitoring and alert system to organic overloading, hydraulic, temperature, and toxicant shocks. The results show that this system enables an accurate and rapid monitoring and diagnosis of the reactor status, and offers reliable early warnings on the potential risks. As the core of this system, the evaluating indexes are demonstrated to be of high accuracy and sensitivity in process evaluation and good adaptability to the artificial intelligence and automated control apparatus. This online monitoring and alert system presents a valuable effort to promote the automated monitoring and control of anaerobic digestion process, and holds a high promise for application.

  5. SCALE Code Validation for Prismatic High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Ilas, Dan

    2013-01-01

    Using experimental data published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments for the fresh cold core of the High Temperature Engineering Test Reactor, a comprehensive validation study has been carried out to assess the performance of the SCALE code system for analysis of high-temperature gas-cooled reactor configurations. This paper describes part of the results of that effort. The studies performed included criticality evaluations for the full core and for the annular cores realized during the fuel loading, as well as calculations and comparisons for excess reactivity, shutdown margin, control rod worths, temperature coefficient of reactivity, and reaction rate distributions. Comparisons of the SCALE results with both experimental values and MCNP-calculated values are presented. The comparisons show that the SCALE calculated results, obtained with both multigroup and continuous energy cross sections, are in reasonable agreement with the experimental data and the MCNP predictions.

  6. Gas core reactors for actinide transmutation. [uranium hexafluoride

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.; Wan, P. T.; Chow, S.

    1979-01-01

    The preliminary design of a uranium hexafluoride actinide transmutation reactor to convert long-lived actinide wastes to shorter-lived fission product wastes was analyzed. It is shown that externally moderated gas core reactors are ideal radiators. They provide an abundant supply of thermal neutrons and are insensitive to composition changes in the blanket. For the present reactor, an initial load of 6 metric tons of actinides is loaded. This is equivalent to the quantity produced by 300 LWR-years of operation. At the beginning, the core produces 2000 MWt while the blanket generates only 239 MWt. After four years of irradiation, the actinide mass is reduced to 3.9 metric tonnes. During this time, the blanket is becoming more fissile and its power rapidly approaches 1600 MWt. At the end of four years, continuous refueling of actinides is carried out and the actinide mass is held constant. Equilibrium is essentially achieved at the end of eight years. At equilibrium, the core is producing 1400 MWt and the blanket 1600 MWt. At this power level, the actinide destruction rate is equal to the production rate from 32 LWRs.

  7. Gas core reactors for actinide transmutation and breeder applications

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.

    1978-01-01

    This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

  8. Ongoing Development of a Series Bosch Reactor System

    NASA Technical Reports Server (NTRS)

    Abney, Morgan; Mansell, Matt; DuMez, Sam; Thomas, John; Cooper, Charlie; Long, David

    2013-01-01

    Future manned missions to deep space or planetary surfaces will undoubtedly require highly robust, efficient, and regenerable life support systems that require minimal consumables. To meet this requirement, NASA continues to explore a Bosch-based carbon dioxide reduction system to recover oxygen from CO2. In order to improve the equivalent system mass of Bosch systems, we seek to design and test a "Series Bosch" system in which two reactors in series are optimized for the two steps of the reaction, as well as to explore the use of in situ materials as carbon deposition catalysts. Here we report recent developments in this effort including assembly and initial testing of a Reverse Water-Gas Shift reactor (RWGSr) and initial testing of two gas separation membranes. The RWGSr was sized to reduce CO2 produced by a crew of four to carbon monoxide as the first stage in a Series Bosch system. The gas separation membranes, necessary to recycle unreacted hydrogen and CO2, were similarly sized. Additionally, we report results of preliminary experiments designed to determine the catalytic properties of Martian and Lunar regolith simulant for the carbon deposition step.

  9. Ongoing Development of a Series Bosch Reactor System

    NASA Technical Reports Server (NTRS)

    Abney, Morgan B; Mansell, J. Matthew; Stanley, Christine; Edmunson, Jennifer; DuMez, Samuel J.; Chen, Kevin

    2013-01-01

    Future manned missions to deep space or planetary surfaces will undoubtedly incorporate highly robust, efficient, and regenerable life support systems that require minimal consumables. To meet this requirement, NASA continues to explore a Bosch-based carbon dioxide reduction system to recover oxygen from CO2. In order to improve the equivalent system mass of Bosch systems, we seek to design and test a "Series Bosch" system in which two reactors in series are optimized for the two steps of the reaction, as well as to explore the use of in situ materials as carbon deposition catalysts. Here we report recent developments in this effort including assembly and initial testing of a Reverse Water-Gas Shift reactor (RWGSr) and initial testing of two gas separation membranes. The RWGSr was sized to reduce CO2 produced by a crew of four to carbon monoxide as the first stage in a Series Bosch system. The gas separation membranes, necessary to recycle unreacted hydrogen and CO2, were similarly sized. Additionally, we report results of preliminary experiments designed to determine the catalytic properties of Martian regolith simulant for the carbon formation step.

  10. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1979-September 30, 1979

    SciTech Connect

    Not Available

    1980-03-07

    The results of work performed from July 1, 1979 through September 30, 1979 on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

  11. System and method for producing substitute natural gas from coal

    DOEpatents

    Hobbs, Raymond

    2012-08-07

    The present invention provides a system and method for producing substitute natural gas and electricity, while mitigating production of any greenhouse gasses. The system includes a hydrogasification reactor, to form a gas stream including natural gas and a char stream, and an oxygen burner to combust the char material to form carbon oxides. The system also includes an algae farm to convert the carbon oxides to hydrocarbon material and oxygen.

  12. Flowing gas, non-nuclear experiments on the gas core reactor

    NASA Technical Reports Server (NTRS)

    Kunze, J. F.; Cooper, C. G.; Macbeth, P. J.

    1973-01-01

    Variations in cavity wall and injection configurations of the gas core reactor were aimed at establishing flow patterns that give a maximum of the nuclear criticality eigenvalue. Correlation with the nuclear effect was made using multigroup diffusion theory normalized by previous benchmark critical experiments. Air was used to simulate the hydrogen propellant in the flow tests, and smoked air, argon, or Freon to simulate the central nuclear fuel gas. Tests were run both in the down-firing and upfiring directions. Results showed that acceptable flow patterns with volume fraction for the simulated nuclear fuel gas and high flow rate ratios of propellant to fuel can be obtained. Using a point injector for the fuel, good flow patterns are obtained by directing the outer gas at high velocity long the cavity wall, using louvered injection schemes. Recirculation patterns were needed to stabilize the heavy central gas when different gases are used.

  13. Evaluation of an Integrated Gas-Cooled Reactor Simulator and Brayton Turbine-Generator

    NASA Technical Reports Server (NTRS)

    Hissam, David Andy; Stewart, Eric T.

    2006-01-01

    A closed-loop brayton cycle, powered by a fission reactor, offers an attractive option for generating both planetary and in-space electric power. Non-nuclear testing of this type of system provides the opportunity to safely work out integration and system control challenges for a modest investment. Recognizing this potential, a team at Marshall Space Flight Center has evaluated the viability of integrating and testing an existing gas-cooled reactor simulator and a modified commercially available, off-the-shelf, brayton turbine-generator. Since these two systems were developed independently of one another, this evaluation had to determine if they could operate together at acceptable power levels, temperatures, and pressures. Thermal, fluid, and structural analyses show that this combined system can operate at acceptable power levels and temperatures. In addition, pressure drops across the reactor simulator, although higher than desired, are also viewed as acceptable. Three potential working fluids for the system were evaluated: N2, He/Ar, and He/Xe. Other potential issues, such as electrical breakdown in the generator and the operation of the brayton foil bearings using various gas mixtures, were also investigated.

  14. Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system

    NASA Technical Reports Server (NTRS)

    Jefferies, K. S.; Tew, R. C.

    1974-01-01

    A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

  15. Automated Gas Distribution System

    NASA Astrophysics Data System (ADS)

    Starke, Allen; Clark, Henry

    2012-10-01

    The cyclotron of Texas A&M University is one of the few and prized cyclotrons in the country. Behind the scenes of the cyclotron is a confusing, and dangerous setup of the ion sources that supplies the cyclotron with particles for acceleration. To use this machine there is a time consuming, and even wasteful step by step process of switching gases, purging, and other important features that must be done manually to keep the system functioning properly, while also trying to maintain the safety of the working environment. Developing a new gas distribution system to the ion source prevents many of the problems generated by the older manually setup process. This developed system can be controlled manually in an easier fashion than before, but like most of the technology and machines in the cyclotron now, is mainly operated based on software programming developed through graphical coding environment Labview. The automated gas distribution system provides multi-ports for a selection of different gases to decrease the amount of gas wasted through switching gases, and a port for the vacuum to decrease the amount of time spent purging the manifold. The Labview software makes the operation of the cyclotron and ion sources easier, and safer for anyone to use.

  16. Space-reactor system and subsystem investigations: cost and schedule estimates for reactor and shield subsystems technology development. SP-100 Program

    SciTech Connect

    Determan, W.R.; Harty, R.B.; Hylin, C.

    1983-06-30

    This report presents cost and schedule estimates of the technology development for reactor and shielding subsystems of a 100-kWe class space reactor electric system. The subsystems technology development (which includes reactor and shield subsystems ground testing) is supported by materials and processes development and component development. For the purpose of the cost estimate, seven generic types of reactor subsystems were used: uranium-zirconium hydride, NaK-cooled thermal reactor; lithium-cooled, refractory-clad fast reactor; Na- or K-cooled fast reactor; in-core thermionic reactor; inert gas-cooled particle fuel reactor; inert gas-cooled metal-clad fast reactor; and heat pipe-cooled fast reactor. Also three levels of technology were included for each of the generic types of reactor subsystem: current, improved, and advanced. The data in this report encompass all these technology levels. The shielding subsystem uses both gamma (heavy-metal) and neutron (hydrogenous material) shields. The shields considered in this report would be used in conjunction with unmanned payloads.

  17. Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

    NASA Astrophysics Data System (ADS)

    Godfroy, Thomas J.; Kapernick, Richard J.; Bragg-Sitton, Shannon M.

    2004-02-01

    One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

  18. Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

    SciTech Connect

    Godfroy, Thomas J.; Bragg-Sitton, Shannon M.; Kapernick, Richard J.

    2004-02-04

    One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

  19. Utilization of the Recycle Reactor in Determining Kinetics of Gas-Solid Catalytic Reactions.

    ERIC Educational Resources Information Center

    Paspek, Stephen C.; And Others

    1980-01-01

    Describes a laboratory scale reactor that determines the kinetics of a gas-solid catalytic reaction. The external recycle reactor construction is detailed with accompanying diagrams. Experimental details, application of the reactor to CO oxidation kinetics, interphase gradients, and intraphase gradients are discussed. (CS)

  20. 157. ARAIII Reactor building (ARA608) Main gas loop mechanical flow ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    157. ARA-III Reactor building (ARA-608) Main gas loop mechanical flow sheet. This drawing was selected as a typical example of mechanical arrangements within reactor building. Aerojet-general 880-area/GCRE-0608-50-013-102634. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  1. Air Ingress Analyses on a High Temperature Gas-Cooled Reactor

    SciTech Connect

    Oh, Chang H; Moore, Richard Leroy; Merrill, Brad Johnson; Petti, David Andrew

    2001-11-01

    A primary-pipe break accident is one of the design-basis accidents of a high-temperature gas-cooled reactor (HTGR). When this accident occurs, air is anticipated to enter the reactor core from the break and oxidize the in-core graphite structure in the modular pebble bed reactor (MPBR). This paper presents the results of the graphite oxidation model developed as part of the Idaho National Engineering and Environmental Laboratory's Direct Research and Development effort. Although gas reactors have been tried in the past with limited success, the innovations of modularity and integrated state-ofart control systems coupled with improved fuel design and a pebble bed core make this design potentially very attractive from an economic and technical perspective. A schematic diagram on a reference design of the MPBR has been established on a major component level (INEEL & MIT, 1999). Steady-state and transient thermal hydraulics models will be produced with key parameters established for these conditions at all major components. Development of an integrated plant model to allow for transient analysis on a more sophisticated level is now being developed. In this paper, preliminary results of the hypothetical air ingress are presented. A graphite oxidation model was developed to determine temperature and the control mechanism in the spherical graphite geometry.

  2. The Secure, Transportable, Autonomous Reactor System

    SciTech Connect

    Brown, N.W.; Hassberger, J.A.; Smith, C.; Carelli, M.; Greenspan, E.; Peddicord, K.L.; Stroh, K.; Wade, D.C.; Hill, R.N.

    1999-05-27

    The Secure, Transportable, Autonomous Reactor (STAR) system is a development architecture for implementing a small nuclear power system, specifically aimed at meeting the growing energy needs of much of the developing world. It simultaneously provides very high standards for safety, proliferation resistance, ease and economy of installation, operation, and ultimate disposition. The STAR system accomplishes these objectives through a combination of modular design, factory manufacture, long lifetime without refueling, autonomous control, and high reliability.

  3. Dynamic Impregnator Reactor System (Poster)

    SciTech Connect

    Not Available

    2012-09-01

    IBRF poster developed for the IBRF showcase. Describes the multifarious system designed for complex feedstock impregnation and processing. IBRF feedstock system has several unit operations combined into one robust system that provides for flexible and staged process configurations, such as spraying, soaking, low-severity pretreatment, enzymatic hydrolysis, fermentation, concentration/evaporation, and distillation.

  4. Gas turbine premixing systems

    DOEpatents

    Kraemer, Gilbert Otto; Varatharajan, Balachandar; Evulet, Andrei Tristan; Yilmaz, Ertan; Lacy, Benjamin Paul

    2013-12-31

    Methods and systems are provided for premixing combustion fuel and air within gas turbines. In one embodiment, a combustor includes an upstream mixing panel configured to direct compressed air and combustion fuel through premixing zone to form a fuel-air mixture. The combustor includes a downstream mixing panel configured to mix additional combustion fuel with the fule-air mixture to form a combustion mixture.

  5. NON-CORROSIVE REACTOR FUEL SYSTEM

    DOEpatents

    Herrick, C.C.

    1962-08-14

    A non-corrosive nuclear reactor fuel system was developed utilizing a molten plutonium-- iron alloy fuel having about 2 at.% carbon and contained in a tantalum vessel. This carbon reacts with the interior surface of the tantalum vessel to form a plutonium resistant self-healing tantalum carbide film. (AEC)

  6. Hybrid Molten Salt Reactor (HMSR) System Study

    SciTech Connect

    Woolley, Robert D; Miller, Laurence F

    2014-04-01

    Can the hybrid system combination of (1) a critical fission Molten Salt Reactor (MSR) having a thermal spectrum and a high Conversion Ratio (CR) with (2) an external source of high energy neutrons provide an attractive solution to the world's expanding demand for energy? The present study indicates the answer is an emphatic yes.

  7. The CANDU Reactor System: An Appropriate Technology.

    PubMed

    Robertson, J A

    1978-02-10

    CANDU power reactors are characterized by the combination of heavy water as moderator and pressure tubes to contain the fuel and coolant. Their excellent neutron economy provides the simplicity and low costs of once-through natural-uranium fueling. Future benefits include the prospect of a near-breeder thorium fuel cycle to provide security of fuel supply without the need to develop a new reactor such as the fast breeder. These and other features make the CANDU system an appropriate technology for countries, like Canada, of intermediate economic and industrial capacity. PMID:17788102

  8. Reactor control rod timing system. [LMFBR

    DOEpatents

    Wu, P.T.K.

    1980-03-18

    A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  9. Control system studies for thermionic reactors

    NASA Technical Reports Server (NTRS)

    Hermsen, R. J.; Gronroos, H. G.

    1978-01-01

    In core thermionic reactor concepts are of interest for space missions that require electrical power in the range of a few tens of kilowatts up to several megawatts. The physical principle involved--thermionic direct conversion of heat to electricity at net efficiencies up to 15 percent--offers potential advantages when compared to other nuclear powerplant concepts. However, the integration of the thermionic diode electrode structure with high-temperature nuclear fuel materials presents new design problems and new reactor physical constraints. Among the topics that must be investigated are those associated with the control system. The results of analytical and simulation studies of thermionic reactor control performed at the Jet Propulsion Laboratory are discussed.

  10. Gaseous fuel reactor systems for aerospace applications

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schwenk, F. C.

    1977-01-01

    Research on the gaseous fuel nuclear rocket concept continues under the programs of the U.S. National Aeronautics and Space Administration (NASA) Office for Aeronautics and Space Technology and now includes work related to power applications in space and on earth. In a cavity reactor test series, initial experiments confirmed the low critical mass determined from reactor physics calculations. Recent work with flowing UF6 fuel indicates stable operation at increased power levels. Preliminary design and experimental verification of test hardware for high-temperature experiments have been accomplished. Research on energy extraction from fissioning gases has resulted in lasers energized by fission fragments. Combined experimental results and studies indicate that gaseous-fuel reactor systems have significant potential for providing nuclear fission power in space and on earth.

  11. Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis

    SciTech Connect

    Ilas, Germina; Ilas, Dan; Kelly, Ryan P; Sunny, Eva E

    2012-08-01

    This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.

  12. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems...

  13. Reactor power system deployment and startup

    NASA Technical Reports Server (NTRS)

    Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

    1985-01-01

    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

  14. Fault-tolerant reactor protection system

    DOEpatents

    Gaubatz, Donald C.

    1997-01-01

    A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service.

  15. Fault-tolerant reactor protection system

    DOEpatents

    Gaubatz, D.C.

    1997-04-15

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service. 16 figs.

  16. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980

    SciTech Connect

    Not Available

    1980-06-25

    Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Included are the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described, including screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, and 950/sup 0/C.

  17. Development of a differential volume reactor system for soil biodegradation studies

    SciTech Connect

    Webb, O.F.; Bienkowski, P.R.; Reed, G.D.

    1991-12-31

    A bench scale experimental system was developed for the analysis of polycyclic aromatic hydrocarbon (PAH) degradation by mixed microbial cultures in PAH contaminated Manufactured Gas Plant (MGP) soils and on sand. The reactor system was chosen in order to provide a fundamental protocol capable for evaluating the performance of specific mixed microbial cultures on specific soil systems by elucidating the important system variables and their interactions. The reactor design and peripherals are described. A plug flow differential volume reactor (DVR) was used in order to remove performance effects related to reactor type, as opposed to system structure. This reactor system could be well represented mathematically. Methods were developed for on-line quantitative determination of PAH liquid phase concentrations. The mathematical models and experimental data are presented for the biodegradation of naphthalene on artificial and MGP soils.

  18. Development of a differential volume reactor system for soil biodegradation studies.

    PubMed

    Webb, O F; Phelps, T J; Bienkowski, P R; DiGrazia, P M; Reed, G D; Applegate, B; White, D C; Sayler, G S

    1991-01-01

    A bench scale experimental system was developed for the analysis of polycyclic aromatic hydrocarbon (PAH) degradation by mixed microbial cultures in PAH contaminated Manufactured Gas Plant (MGP) soils and on sand. The reactor system was chosen in order to provide a fundamental protocol capable for evaluating the performance of specific mixed microbial cultures on specific soil systems by elucidating the important system variables and their interactions. The reactor design and peripherals are described. A plug flow differential volume reactor (DVR) was used in order to remove performance effects related to reactor type, as opposed to system structure. This reactor system could be well represented mathematically. Methods were developed for on-line quantitative determination of PAH liquid phase concentrations. The mathematical models and experimental data are presented for the biodegradation of naphthalene on artificial and MGP soils.

  19. Parallel reactor systems for bioprocess development.

    PubMed

    Weuster-Botz, Dirk

    2005-01-01

    Controlled parallel bioreactor systems allow fed-batch operation at early stages of process development. The characteristics of shaken bioreactors operated in parallel (shake flask, microtiter plate), sparged bioreactors (small-scale bubble column) and stirred bioreactors (stirred-tank, stirred column) are briefly summarized. Parallel fed-batch operation is achieved with an intermittent feeding and pH-control system for up to 16 bioreactors operated in parallel on a scale of 100 ml. Examples of the scale-up and scale-down of pH-controlled microbial fed-batch processes demonstrate that controlled parallel reactor systems can result in more effective bioprocess development. Future developments are also outlined, including units of 48 parallel stirred-tank reactors with individual pH- and pO2-controls and automation as well as liquid handling system, operated on a scale of ml.

  20. Utilization/Disposition of Reactor-Grade Plutonium in High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Bonin, Bernard; Greneche, Dominique; Carre, Frank; Damian, Frederic; Doriath, Jean-Yves

    2004-03-15

    High-temperature gas-cooled reactors (HTRs) are able to accommodate a wide variety of mixtures of fissile and fertile materials without any significant modification of the core design. This flexibility is due to an uncoupling between the parameters of cooling geometry and the parameters that characterize neutronic optimization (moderation ratio or heavy nuclide concentration and distribution).Among other advantageous features, an HTR core has a better neutron economy than a light water reactor (LWR) because there is much less parasitic capture in the moderator (capture cross section of graphite is 100 times less than the one of water), in internal structures, and in fission products (because of a harder spectrum).Moreover, thanks to the high strength of the coated particles, HTR fuels are able to reach very high burnups, far beyond the possibilities offered by other fuels (except the special case of molten salt reactors).These features make HTRs potentially interesting for closing the nuclear fuel cycle and stabilizing the plutonium inventory.A large number of fuel cycle studies are already available today on three main categories of fuel cycles involving HTRs: (a) high-enriched uranium cycle, based on thorium utilization as a fertile material and high-enriched uranium as a fissile material; (b) low-enriched uranium cycle (LEU), where only LEU is used (from 5 to 15%); (c) plutonium cycle based on the utilization of plutonium only as a fissile material, with (or without) fertile materials.Plutonium consumption at high burnups in HTRs has already been tested with encouraging results under the DRAGON project and at Peach Bottom. To maximize plutonium consumption, recent core studies have also been performed on plutonium HTR cores, with special emphasis on weapon-grade plutonium consumption. We complete the picture by a core study for an HTR burning reactor-grade plutonium. Limits in burnup due to core neutronics are investigated for this type of core.With these limits

  1. Direct conversion nuclear reactor space power systems

    SciTech Connect

    Britt, E.J.; Fitzpatrick, G.O.

    1982-08-01

    This paper presents the results of a study of space nuclear reactor power systems using either thermoelectric or thermionic energy converters. An in-core reactor design and two heat pipe cooled out-of-core reactor designs were considered. One of the out-of-core cases utilized, long heat pipes (LHP) directly coupled to the energy converter. The second utilized a larger number of smaller heat pipes (mini-pipe) radiatively coupled to the energy converter. In all cases the entire system, including power conditioning, was constrained to be launched in a single shuttle flight. Assuming presently available performance, both the LHP thermoelectric system and minipipe thermionic system, designed to produce 100 kWe for seven years, would have a specific mass near 22kg/kWe. The specific mass of the thermionic minipipe system designed for a one year mission is 165 kg/kWe due to less fuel swelling. Shuttle imposed growth limits are near 300 kWe and 1.2 MWe for the thermoelectric and thermionic systems, respectively. Converter performance improvements could double this potential, and over 10 MWe may be possible for very short missions.

  2. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, Philippe

    1994-01-01

    A system for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary.

  3. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, P.

    1994-07-05

    A system is described for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary. 2 figures.

  4. Emergency Decay Heat Removal in a GEN-IV Gas-Cooled Fast Reactor

    SciTech Connect

    Cheng, Lap Y.; Ludewig, Hans; Jo, Jae

    2006-07-01

    A series of transient analyses using the system code RELAP5-3d has been performed to confirm the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400 MWt GEN-IV gas-cooled fast reactor. The accident sequence of interest is a station blackout simultaneous with a small break (10 sq.inch/0.645 m{sup 2}) in the reactor vessel. The analyses cover the three phases of decay heat removal in a depressurization accident: (1) forced flow cooling by the power conversion unit (PCU) coast down, (2) active forced flow cooling by a battery powered blower, and (3) passive cooling by natural circulation. The blower is part of an emergency cooling system (ECS) that by design is to sustain passive decay heat removal via natural circulation cooling 24 hours after shutdown. The RELAP5 model includes the helium-cooled reactor, the ECS (primary and secondary side), the PCU with all the rotating machinery (turbine and compressors) and the heat transfer components (recuperator, pre-cooler and inter-cooler), and the guard containment that surrounds the reactor and the PCU. The transient analysis has demonstrated the effectiveness of passive decay heat removal by natural circulation cooling when the guard containment pressure is maintained at or above 800 kPa. (authors)

  5. Monitoring system for a liquid-cooled nuclear fission reactor

    DOEpatents

    DeVolpi, Alexander

    1987-01-01

    A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.

  6. Static conversion systems. [for space power reactors

    NASA Technical Reports Server (NTRS)

    Ewell, R.; Mondt, J.

    1985-01-01

    Historically, all space power systems that have actually flown in space have relied on static energy conversion technology. Thus, static conversion is being considered for space nuclear power systems as well. There are four potential static conversion technologies which should be considered. These include: the alkali metal thermoelectric converter (AMTEC), the thermionic converter, the thermoelectric converter, and the thermophotovoltaic converter (TPV). These four conversion technologies will be described in brief detail along with their current status and development needs. In addition, the systems implications of using each of these conversion technologies with a space nuclear reactor power system will be evaluated and some comparisons made.

  7. Microchannel Reactor System for Catalytic Hydrogenation

    SciTech Connect

    Adeniyi Lawal; Woo Lee; Ron Besser; Donald Kientzler; Luke Achenie

    2010-12-22

    We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstrated on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.

  8. Reactor User Interface Technology Development Roadmaps for a High Temperature Gas-Cooled Reactor Outlet Temperature of 750 degrees C

    SciTech Connect

    Ian Mckirdy

    2010-12-01

    This report evaluates the technology readiness of the interface components that are required to transfer high-temperature heat from a High Temperature Gas-Cooled Reactor (HTGR) to selected industrial applications. This report assumes that the HTGR operates at a reactor outlet temperature of 750°C and provides electricity and/or process heat at 700°C to conventional process applications, including the production of hydrogen.

  9. Neutral-beam-injection systems for reactors

    SciTech Connect

    Pyle, R.V.

    1983-06-01

    Increasing effort is being put into engineering designs of reactors and reactor-like magnetic confinement experiments. A central question concerns the methods of heating, fueling, and maintaining the plasmas, functions that primarily are now performed by neutral beams. Planning in the USA does not include the use of neutral beams on tokamaks in the 1990's and beyond. Tandem mirrors, however, will use energetic beams (sloshing ion beams) in the end plugs to produce electrostatic potentials that will confine plasma ions. These systems will be based on the production, acceleration, transport, and neutralization of negative hydrogen-ion (D/sup -/), multiampere beams with energies of 200-to 500-keV. In addition, lower-energy D and T beams may be used. These systems must operate steady state, with high reliability, and be compatible with radiation from a D-T burning plasma.

  10. Plasma generators, reactor systems and related methods

    DOEpatents

    Kong, Peter C.; Pink, Robert J.; Lee, James E.

    2007-06-19

    A plasma generator, reactor and associated systems and methods are provided in accordance with the present invention. A plasma reactor may include multiple sections or modules which are removably coupled together to form a chamber. Associated with each section is an electrode set including three electrodes with each electrode being coupled to a single phase of a three-phase alternating current (AC) power supply. The electrodes are disposed about a longitudinal centerline of the chamber and are arranged to provide and extended arc and generate an extended body of plasma. The electrodes are displaceable relative to the longitudinal centerline of the chamber. A control system may be utilized so as to automatically displace the electrodes and define an electrode gap responsive to measure voltage or current levels of the associated power supply.

  11. Summary of space nuclear reactor power systems, 1983--1992

    SciTech Connect

    Buden, D.

    1993-08-11

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  12. Power conditioning for space nuclear reactor systems

    NASA Technical Reports Server (NTRS)

    Berman, Baruch

    1987-01-01

    This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

  13. Development of ceramic membrane reactors for high temperature gas cleanup

    SciTech Connect

    Roberts, D.L.; Way, J.D.

    1992-01-01

    The more difficult part of developing a membrane reactor was making a membrane with suitable permselectivity. Since both the ammonia and H[sub 2]S decomposition reactions produce H[sub 2], we want a membrane that selectively permeates hydrogen. We have taken the approach of coating a substrate that has fine pores (a microfilter or ultrafilter) with a completely dense layer and then (except for palladium coatings) malting super fine pores in the dense layer by leaching or pyrolysis. We used four different substrates (Norton 0.2 alumina monolith, Refractron alumina microfilters, Vycor glass, Alcoa ultrafilter) and five different coating materials (Poly N-methyl silazane, Aremco 617 alumina-based glaze, Polycyclohydridomethyl silazane, Aluminum phosphorus oxides, palladium). Only the palladium films on an alumina ultrafilter could be made regularly with a negligible defect population. The other approaches were plagued with cracks and poor reproducibility. We made palladium films on a tubular alumina ultrafilter (US Filter, Warrendale, PA) by electroless plating from a platinum amine complex solution in hydrazine (Rhoda, 1959). The best results were achieved with an ultrafilter pore size of 100 angstroms and with pretreatment of the surface in a stannic/stannous chloride solution. The films were about 5 microns thick and appeared defect free under electron microscopy. Because defect-free palladium films are infinitely selective to hydrogen with respect to the other components of gasifier product gas, we expect that the selectivity of our palladium films will exceed 1000 even with the inevitable defects that accompany real membranes. We have analyzed the performance and economics of membrane reactors for H[sub 2]S and NH[sub 3] decomposition using the reaction rates determined experimentally in our catalyst development work and using the permeation behavior of palladium films reported in the literature.

  14. Development of ceramic membrane reactors for high temperature gas cleanup

    SciTech Connect

    Roberts, D.L.; Way, J.D.

    1992-11-01

    The more difficult part of developing a membrane reactor was making a membrane with suitable permselectivity. Since both the ammonia and H{sub 2}S decomposition reactions produce H{sub 2}, we want a membrane that selectively permeates hydrogen. We have taken the approach of coating a substrate that has fine pores (a microfilter or ultrafilter) with a completely dense layer and then (except for palladium coatings) malting super fine pores in the dense layer by leaching or pyrolysis. We used four different substrates (Norton 0.2 alumina monolith, Refractron alumina microfilters, Vycor glass, Alcoa ultrafilter) and five different coating materials (Poly N-methyl silazane, Aremco 617 alumina-based glaze, Polycyclohydridomethyl silazane, Aluminum phosphorus oxides, palladium). Only the palladium films on an alumina ultrafilter could be made regularly with a negligible defect population. The other approaches were plagued with cracks and poor reproducibility. We made palladium films on a tubular alumina ultrafilter (US Filter, Warrendale, PA) by electroless plating from a platinum amine complex solution in hydrazine (Rhoda, 1959). The best results were achieved with an ultrafilter pore size of 100 angstroms and with pretreatment of the surface in a stannic/stannous chloride solution. The films were about 5 microns thick and appeared defect free under electron microscopy. Because defect-free palladium films are infinitely selective to hydrogen with respect to the other components of gasifier product gas, we expect that the selectivity of our palladium films will exceed 1000 even with the inevitable defects that accompany real membranes. We have analyzed the performance and economics of membrane reactors for H{sub 2}S and NH{sub 3} decomposition using the reaction rates determined experimentally in our catalyst development work and using the permeation behavior of palladium films reported in the literature.

  15. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Asiah, Nur; Shafii, M. Ali

    2010-12-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio.

  16. Reduction of Switching Surge by Controlled Shunt Reactor Switching of Gas Circuit Breaker

    NASA Astrophysics Data System (ADS)

    Kohyama, Haruhiko; Ito, Hiroki; Asai, Jun; Hidaka, Mikio; Yonezawa, Takashi

    Controlled switching system (CSS) with a 204kV gas circuit breaker (GCB) demonstrates to reduce an inrush current and to eliminate re-ignitions in shunt reactor switching. Target closing and opening instants for controlled switching are determined by the electrical and mechanical characteristics of GCB. Idle time dependence of the operating time, which gives a sufficient impact on controlling accuracy, is evaluated and successfully compensated by synchronous switching controller (SSC). Field operation of CSS for a year shows an effective suppression of inrush current on closing and prevention of re-ignitions on opening of shunt reactor. The operations also proved that the CSS has a satisfactory performance for electro-magnetic disturbance imposed in practical field.

  17. Nuclear reactor pressure vessel support system

    DOEpatents

    Sepelak, George R.

    1978-01-01

    A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

  18. The liquid annular reactor system (LARS) propulsion

    SciTech Connect

    Maise, G.; Lazareth, O.W.; Horn, F.; Powell, J.R.; Ludewig, H. ); Lenard, R.X. )

    1991-01-05

    A new concept for very high specific impulse ({gt}2000 seconds) direct nuclear propulsion is described. The concept, termed LARS (Liquid Annular Reactor System) uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures ({similar to}6000 K). Operating pressure is moderate ({similar to}10 atm), with the result that the outlet hydrogen is virtually 100% dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use 7 rotating fuel elements, are beryllium moderated and have critical radii of {similar to}100 cm (core L/D{approx}1.5).

  19. 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.

  20. 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.

  1. The Liquid Annular Reactor System (LARS) propulsion

    NASA Technical Reports Server (NTRS)

    Powell, James; Ludewig, Hans; Horn, Frederick; Lenard, Roger

    1990-01-01

    A concept for very high specific impulse (greater than 2000 seconds) direct nuclear propulsion is described. The concept, termed the liquid annular reactor system (LARS), uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (approximately 6000 K). Operating pressure is moderate (approximately 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use seven rotating fuel elements, are beryllium moderated, and have critical radii of approximately 100 cm (core L/D approximately equal to 1.5).

  2. AURORA: A FORTRAN program for modeling well stirred plasma and thermal reactors with gas and surface reactions

    SciTech Connect

    Meeks, E.; Grcar, J.F.; Kee, R.J.; Moffat, H.K.

    1996-02-01

    The AURORA Software is a FORTRAN computer program that predicts the steady-state or time-averaged properties of a well mixed or perfectly stirred reactor for plasma or thermal chemistry systems. The software was based on the previously released software, SURFACE PSR which was written for application to thermal CVD reactor systems. AURORA allows modeling of non-thermal, plasma reactors with the determination of ion and electron concentrations and the electron temperature, in addition to the neutral radical species concentrations. Well stirred reactors are characterized by a reactor volume, residence time or mass flow rate, heat loss or gas temperature, surface area, surface temperature, the incoming temperature and mixture composition, as well as the power deposited into the plasma for non-thermal systems. The model described here accounts for finite-rate elementary chemical reactions both in the gas phase and on the surface. The governing equations are a system of nonlinear algebraic relations. The program solves these equations using a hybrid Newton/time-integration method embodied by the software package TWOPNT. The program runs in conjunction with the new CHEMKIN-III and SURFACE CHEMKIN-III packages, which handle the chemical reaction mechanisms for thermal and non-thermal systems. CHEMKIN-III allows for specification of electron-impact reactions, excitation losses, and elastic-collision losses for electrons.

  3. Gas turbine cooling system

    SciTech Connect

    Bancalari, Eduardo E.

    2001-01-01

    A gas turbine engine (10) having a closed-loop cooling circuit (39) for transferring heat from the hot turbine section (16) to the compressed air (24) produced by the compressor section (12). The closed-loop cooling system (39) includes a heat exchanger (40) disposed in the flow path of the compressed air (24) between the outlet of the compressor section (12) and the inlet of the combustor (14). A cooling fluid (50) may be driven by a pump (52) located outside of the engine casing (53) or a pump (54) mounted on the rotor shaft (17). The cooling circuit (39) may include an orifice (60) for causing the cooling fluid (50) to change from a liquid state to a gaseous state, thereby increasing the heat transfer capacity of the cooling circuit (39).

  4. Heat exchanger design considerations for high temperature gas-cooled reactor (HTGR) plants

    SciTech Connect

    McDonald, C.F.; Vrable, D.L.; Van Hagan, T.H.; King, J.H.; Spring, A.H.

    1980-02-01

    Various aspects of the high-temperature heat exchanger conceptual designs for the gas turbine (HTGR-GT) and process heat (HTGR-PH) plants are discussed. Topics include technology background, heat exchanger types, surface geometry, thermal sizing, performance, material selection, mechanical design, fabrication, and the systems-related impact of installation and integration of the units in the prestressed concrete reactor vessel. The impact of future technology developments, such as the utilization of nonmetallic materials and advanced heat exchanger surface geometries and methods of construction, is also discussed.

  5. Numerical simulation of gas dynamics and heat exchange tasks in fuel assemblies of the nuclear reactors

    SciTech Connect

    Zhuchenko, S. V.

    2014-11-12

    This report presents a PC-based program for solution gas dynamics and heat exchange mathematical tasks in fuel assemblies of the fast-neutron nuclear reactors. A fuel assembly consisting of bulk heat-generating elements, which are integrated together by the system of supply and pressure manifolds, is examined. Spherical heat-generating microelements, which contain nuclear fuel, are pulled into the heat-generating elements. Gaseous coolant proceed from supply manifolds to heat-generating elements, where it withdraws the nuclear reaction heat and assembles in pressure manifolds.

  6. Reactor coolant pump monitoring and diagnostic system

    SciTech Connect

    Singer, R.M.; Gross, K.C.; Walsh, M. ); Humenik, K.E. )

    1990-01-01

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs.

  7. Colliding Beam Fusion Reactor Space Propulsion System

    NASA Astrophysics Data System (ADS)

    Cheung, A.; Binderbauer, M.; Liu, F.; Qerushi, A.; Rostoker, N.; Wessel, F. J.

    2004-02-01

    The Colliding Beam Fusion Reactor Space Propulsion System, CBFR-SPS, is an aneutronic, magnetic-field-reversed configuration, fueled by an energetic-ion mixture of hydrogen and boron11 (H-B11). Particle confinement and transport in the CBFR-SPS are classical, hence the system is scaleable. Fusion products are helium ions, α-particles, expelled axially out of the system. α-particles flowing in one direction are decelerated and their energy recovered to ``power'' the system; particles expelled in the opposite direction provide thrust. Since the fusion products are charged particles, the system does not require the use of a massive-radiation shield. This paper describes a 100 MW CBFR-SPS design, including estimates for the propulsion-system parameters and masses. Specific emphasis is placed on the design of a closed-cycle, Brayton-heat engine, consisting of heat-exchangers, turbo-alternator, compressor, and finned radiators.

  8. Work Breakdown Structure and Plant/Equipment Designation System Numbering Scheme for the High Temperature Gas- Cooled Reactor (HTGR) Component Test Capability (CTC)

    SciTech Connect

    Jeffrey D Bryan

    2009-09-01

    This white paper investigates the potential integration of the CTC work breakdown structure numbering scheme with a plant/equipment numbering system (PNS), or alternatively referred to in industry as a reference designation system (RDS). Ideally, the goal of such integration would be a single, common referencing system for the life cycle of the CTC that supports all the various processes (e.g., information, execution, and control) that necessitate plant and equipment numbers be assigned. This white paper focuses on discovering the full scope of Idaho National Laboratory (INL) processes to which this goal might be applied as well as the factors likely to affect decisions about implementation. Later, a procedure for assigning these numbers will be developed using this white paper as a starting point and that reflects the resolved scope and outcome of associated decisions.

  9. Practical reactor systems for yeast cell immobilization using biomass support particles

    SciTech Connect

    Black, G.M.; Webb, C.; Mattews, T.M.; Atkinson, B.

    1984-01-01

    The technique of cell immobilization using porous support particles (biomass support particles) has been successfully applied to yeast cells. Two reactor configurations exploiting the use of these particles have been developed and assessed for use in aseptic yeast fermentations. A liquid-fluidized bed fermenter has been devised for use with particles denser than the fermentation liquor whilst a gas-stirred circulating bed fermenter proved suitable for particles of essentially neutral buoyancy. Both systems have been operated successfully for extended periods of continuous operation. The utilization of biomass support particle technology in such reactors provides a practical and robust system for immobilized cell reactors. This technology offers significant opportunities for further development.

  10. Weld monitor and failure detector for nuclear reactor system

    DOEpatents

    Sutton, Jr., Harry G.

    1987-01-01

    Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

  11. The DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification Program

    SciTech Connect

    David Petti; Hans Gougar; Gary Bell

    2005-05-01

    The Department of Energy has established the Advanced Gas Reactor Fuel Development and Qualification Program to address the following overall goals: Provide a baseline fuel qualification data set in support of the licensing and operation of the Next Generation Nuclear Plant (NGNP). Gas-reactor fuel performance demonstration and qualification comprise the longest duration research and development (R&D) task for the NGNP feasibility. The baseline fuel form is to be demonstrated and qualified for a peak fuel centerline temperature of 1250°C. Support near-term deployment of an NGNP by reducing market entry risks posed by technical uncertainties associated with fuel production and qualification. Utilize international collaboration mechanisms to extend the value of DOE resources. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, postirradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete fundamental understanding of the relationship between the fuel fabrication process, key fuel properties, the irradiation performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. Fuel performance modeling and analysis of the fission product behavior in the primary circuit are important aspects of this work. The performance models are considered essential for several reasons, including guidance for the plant designer in establishing the core design and operating limits, and demonstration to the licensing authority that the applicant has a thorough understanding of the in-service behavior of the fuel system. The fission product behavior task will also provide primary source term data needed for licensing. An overview of the program and recent progress will be presented.

  12. A multi-phase, micro-dispersion reactor for the continuous production of methane gas hydrate

    SciTech Connect

    Taboada Serrano, Patricia L; Ulrich, Shannon M; Szymcek, Phillip; McCallum, Scott; Phelps, Tommy Joe; Palumbo, Anthony Vito; Tsouris, Costas

    2009-01-01

    A continuous-jet hydrate reactor originally developed to generate a CO2 hydrate stream has been modified to continuously produce CH4 hydrate. The reactor has been tested in the Seafloor Process Simulator (SPS), a 72-L pressure vessel available at Oak Ridge National Laboratory. During experiments, the reactor was submerged in water inside the SPS and received water from the surrounding through a submersible pump and CH4 externally through a gas booster pump. Thermodynamic conditions in the hydrate stability regime were employed in the experiments. The reactor produced a continuous stream of CH4 hydrate, and based on pressure values and amount of gas injected, the conversion of gas to hydrate was estimated. A conversion of up to 70% was achieved using this reactor.

  13. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  14. Indirect passive cooling system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1990-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  15. Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

    SciTech Connect

    Lin, Jerry Y.S.

    2013-01-29

    Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

  16. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    SciTech Connect

    Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville; Gougar, Hans David; Strydom, Gerhard

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  17. THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors

    SciTech Connect

    Kroeger, P.G.; Kennett, R.J.; Colman, J.; Ginsberg, T. )

    1991-10-01

    This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MW{sub th} reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs.

  18. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    NASA Technical Reports Server (NTRS)

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

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  19. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    DOE PAGES

    Cheng, Lap-Yan; Wei, Thomas Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow weremore » evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.« less

  20. Gas-Recovery System

    DOEpatents

    Heckman, R. A.

    1971-12-14

    Nuclear explosions have been proposed as a means for recovering gas from underground gas-bearing rock formations. In present practice, the nuclear device is positioned at the end of a long pipe which is subsequently filled with grout or concrete. After the device is exploded, the grout is drilled through to provide a flow path for the released gas to the ground surface. As settled grout is brittle, often the compressive shock of the explosion fractures the grout and deforms the pipe so that it may not be removed nor reused. In addition, the pipe is sometimes pinched off completely and the gas flow is totally obstructed. (2 claims)

  1. Gas-recovery system

    DOEpatents

    Heckman, R.A.

    1971-12-14

    Nuclear explosions have been proposed as a means for recovering gas from underground gas-bearing rock formations. In present practice, the nuclear device is positioned at the end of a long pipe which is subsequently filled with grout or concrete. After the device is exploded, the grout is drilled through to provide a flow path for the released gas to the ground surface. As settled grout is brittle, often the compressive shock of the explosion fractures the grout and deforms the pipe so that it may not be removed nor reused. In addition, the pipe is sometimes pinched off completely and the gas flow is totally obstructed. (2 claims)

  2. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere...

  3. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere...

  4. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere...

  5. Integrated systems analysis of the PIUS reactor

    SciTech Connect

    Fullwood, F.; Kroeger, P.; Higgins, J.

    1993-11-01

    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  6. Nuclear reactor fuel rod attachment system

    DOEpatents

    Not Available

    1980-09-17

    A reusable system is described for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member. The locking cap has two opposing fingers shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed.

  7. New approaches in the USA for high-temperature gas-cooled reactors. Gas-Cooled Reactor Programs

    SciTech Connect

    Kasten, P.R.; Neylan, A.J.; Penfield, S.R. Jr.

    1985-08-01

    Several concepts are being evaluated in the US HTR Program to explore designs which might improve the commercial viability of nuclear power. The general approach is to reduce the reactor power and increase the ability to use inherent features for removing heat following extreme accidents. The unit size and design of these concepts are constrained so that extreme accidents do not result in significant release of radioactivity from the reactor plant. Through the greater reliance on inherent safety features in small HTRs, it should be possible to minimize the amount of nuclear grade components required in the balance-of-plant, which could lead to an economic system. Four HTR concepts are presently being evaluated within the US Program, and these concepts are briefly summarized. A modular HTR using a steel pressure vessel, which is very similar to one of the four HTR concepts being evaluated within the US National program, is presented as an example of a specific concept to illustrate the features and performance of HTRs having a high degree of inherent safety.

  8. Fission gas detection system

    DOEpatents

    Colburn, Richard P.

    1985-01-01

    A device for collecting fission gas released by a failed fuel rod which device uses a filter to pass coolant but which filter blocks fission gas bubbles which cannot pass through the filter due to the surface tension of the bubble.

  9. Gas hydrate cool storage system

    DOEpatents

    Ternes, M.P.; Kedl, R.J.

    1984-09-12

    The invention presented relates to the development of a process utilizing a gas hydrate as a cool storage medium for alleviating electric load demands during peak usage periods. Several objectives of the invention are mentioned concerning the formation of the gas hydrate as storage material in a thermal energy storage system within a heat pump cycle system. The gas hydrate was formed using a refrigerant in water and an example with R-12 refrigerant is included. (BCS)

  10. Alaska Natural Gas Transportation System

    SciTech Connect

    Jones, V.T.

    1984-04-27

    The proven reserves of natural gas in Prudhoe Bay remain the single largest block of reserves under US control. The sponsors of the Alaska Natural Gas Transportation System, including The Williams Companies, remain convinced that Alaskan gas will be increasingly important to meet future needs here in the lower 48 states. Both Canada and the US will increasingly have to turn to more costly supplies of gas as the closer, traditional areas of gas supply are exhausted. A principal motivation for Canada's participation in the ANGTS was the prospect of a jointly sponsored pipeline through Canada which would facilitate bringing frontier gas to market - through the so-called Dempster lateral. The high cost of transportation systems in the Artic necessitates pipelines with large capacities in order to minimize the cost of transportation per unit of gas delivered. It is clear that Canada still strongly supports the ANGTS project as a means of opening up the frontier resources of both Alaska and Canada.

  11. Structural materials challenges for advanced reactor systems

    NASA Astrophysics Data System (ADS)

    Yvon, P.; Carré, F.

    2009-03-01

    Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials

  12. Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.

    1977-01-01

    Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

  13. ANALYSIS OF A HIGH TEMPERATURE GAS-COOLED REACTOR POWERED HIGH TEMPERATURE ELECTROLYSIS HYDROGEN PLANT

    SciTech Connect

    M. G. McKellar; E. A. Harvego; A. M. Gandrik

    2010-11-01

    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  14. Process gas solidification system

    DOEpatents

    Fort, William G. S.; Lee, Jr., William W.

    1978-01-01

    It has been the practice to (a) withdraw hot, liquid UF.sub.6 from various systems, (b) direct the UF.sub.6 into storage cylinders, and (c) transport the filled cylinders to another area where the UF.sub.6 is permitted to solidify by natural cooling. However, some hazard attends the movement of cylinders containing liquid UF.sub.6, which is dense, toxic, and corrosive. As illustrated in terms of one of its applications, the invention is directed to withdrawing hot liquid UF.sub.6 from a system including (a) a compressor for increasing the pressure and temperature of a stream of gaseous UF.sub.6 to above its triple point and (b) a condenser for liquefying the compressed gas. A network containing block valves and at least first and second portable storage cylinders is connected between the outlet of the condenser and the suction inlet of the compressor. After an increment of liquid UF.sub.6 from the condenser has been admitted to the first cylinder, the cylinder is connected to the suction of the compressor to flash off UF.sub.6 from the cylinder, thus gradually solidifying UF.sub.6 therein. While the first cylinder is being cooled in this manner, an increment of liquid UF.sub.6 from the condenser is transferred into the second cylinder. UF.sub.6 then is flashed from the second cylinder while another increment of liquid UF.sub.6 is being fed to the first. The operations are repeated until both cylinders are filled with solid UF.sub.6, after which they can be moved safely. As compared with the previous technique, this procedure is safer, faster, and more economical. The method also provides the additional advantage of removing volatile impurities from the UF.sub.6 while it is being cooled.

  15. Rapid pressure swing absorption cleanup of post-shift reactor synthesis gas

    SciTech Connect

    Sirkar, K.K.; Majumdar, S.; Bhaumik, S.

    1991-10-31

    This investigation is concerned with the separation of gas mixtures using a novel concept of rapid pressure swing absorption (RAPSAB) of gas in a stationary absorbent liquid through gas-liquid interfaces immobilized in the pore mouths of hydrophobic microporous membranes. The process is implemented in a module well-packed with hydrophobic microporous hollow fiber membranes. The specific objectives are (1) to fiber membranes. The specific objectives are (1) to develop a theoretical model which will provide guidelines for selecting an efficient RAPSAB process cycle which includes desorption; (2) to demonstrate the concept experimentally with a simple gas mixture (e.g., Co{sub 2}-N{sub 2}) and a simple absorbent liquid such as water, and (3) to extend the concept to reactive absorbent liquids for the separation of CO, Co{sub 2} from the post-shift reactor synthesis gas. A simplified theoretical description of the novel rapid pressure swing absorption process has been developed. The absorption part of the pressure swing absorption cycle has been predicted for CO{sub 2}-N{sub 2}-water system. Numerical simulation of the model is being carried out for different operating conditions for selecting an optimum pressure swing cycle.

  16. Validation of SCALE and the TRITON Depletion Sequence for Gas-Cooled Reactor Analysis

    SciTech Connect

    DeHart, Mark D; Pritchard, Megan L

    2008-01-01

    The very-high-temperature reactor (VHTR) is an advanced reactor concept that uses graphite-moderated fuel and helium gas as a coolant. At present there are two primary VHTR reactor designs under consideration for development: in the pebble-bed reactor, a core is loaded with 'pebbles' consisting of 6 cm diameter spheres, while in a high-temperature gas-cooled reactor, fuel rods are placed within prismatic graphite blocks. In both systems, fuel elements (spheres or rods) are comprised of tristructural-isotropic (TRISO) fuel particles. The TRISO particles are either dispersed in the matrix of a graphite pebble for the pebble-bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks for the prismatic concept. Two levels of heterogeneity exist in such fuel designs: (1) microspheres of TRISO particles dispersed in a graphite matrix of a cylindrical or spherical shape, and (2) neutron interactions at the rod-to-rod or sphere-to-sphere level. Such double heterogeneity (DH) provides a challenge to multigroup cross-section processing methods, which must treat each level of heterogeneity separately. A new capability to model doubly heterogeneous systems was added to the SCALE system in the release of Version 5.1. It was included in the control sequences CSAS and CSAS6, which use the Monte Carlo codes KENO V.a and KENO-VI, respectively, for three-dimensional neutron transport analyses and in the TRITON sequence, which uses the two-dimensional lattice physics code NEWT along with both versions of KENO for transport and depletion analyses. However, the SCALE 5.1 version of TRITON did not support the use of the DH approach for depletion. This deficiency has been addressed, and DH depletion will be available as an option in the upcoming release of SCALE 6. At present Oak Ridge National Laboratory (ORNL) staff are developing a set of calculations that may be used to validate SCALE for DH calculations. This paper discusses the results of

  17. THE INTEGRATION OF PROCESS HEAT APPLICATIONS TO HIGH TEMPERATURE GAS REACTORS

    SciTech Connect

    Michael G. McKellar

    2011-11-01

    A high temperature gas reactor, HTGR, can produce industrial process steam, high-temperature heat-transfer gases, and/or electricity. In conventional industrial processes, these products are generated by the combustion of fossil fuels such as coal and natural gas, resulting in significant emissions of greenhouse gases such as carbon dioxide. Heat or electricity produced in an HTGR could be used to supply process heat or electricity to conventional processes without generating any greenhouse gases. Process heat from a reactor needs to be transported by a gas to the industrial process. Two such gases were considered in this study: helium and steam. For this analysis, it was assumed that steam was delivered at 17 MPa and 540 C and helium was delivered at 7 MPa and at a variety of temperatures. The temperature of the gas returning from the industrial process and going to the HTGR must be within certain temperature ranges to maintain the correct reactor inlet temperature for a particular reactor outlet temperature. The returning gas may be below the reactor inlet temperature, ROT, but not above. The optimal return temperature produces the maximum process heat gas flow rate. For steam, the delivered pressure sets an optimal reactor outlet temperature based on the condensation temperature of the steam. ROTs greater than 769.7 C produce no additional advantage for the production of steam.

  18. Spacecraft cryogenic gas storage systems

    NASA Technical Reports Server (NTRS)

    Rysavy, G.

    1971-01-01

    Cryogenic gas storage systems were developed for the liquid storage of oxygen, hydrogen, nitrogen, and helium. Cryogenic storage is attractive because of the high liquid density and low storage pressure of cryogens. This situation results in smaller container sizes, reduced container-strength levels, and lower tankage weights. The Gemini and Apollo spacecraft used cryogenic gas storage systems as standard spacecraft equipment. In addition to the Gemini and Apollo cryogenic gas storage systems, other systems were developed and tested in the course of advancing the state of the art. All of the cryogenic storage systems used, developed, and tested to date for manned-spacecraft applications are described.

  19. Gas storage and recovery system

    NASA Technical Reports Server (NTRS)

    Cook, Joseph S., Jr. (Inventor)

    1994-01-01

    A system for recovering and recycling gases is disclosed. The system is comprised of inlet and outlet flow lines, controllers, an inflatable enclosure, and inflatable rib stiffeners which are inflatable by the gas to be stored. The system does not present gas at an undesirable back pressure to the gas source. A filtering relief valve is employed which prevents environmental airborne contamination from flowing back into the system when the relief valve is closing. The system is for storing and re-using helium.

  20. Dynamic safety systems in U.S. light water reactors

    SciTech Connect

    Miller, D.W.; Adams, G.; Hajek, B.K.

    1995-12-31

    The use of dynamic rather than static logic in reactor safety function systems provides significant benefits in achieving a fail-safe design. Dynamic safety system (DSS) are based on such an approach that can be realized in hardware- and/or software-based products. AEA Technology has implemented a dynamic architecture in a number of systems licensed and used on commercial gas-cooled reactors, including those in Refs. 1, 2, and 3, where software elements are operationally verified by hardwired components. The principal software-based components in DSS are the trip algorithm computers (TACs) and vote algorithm computers (VACs). The TACs provide trip thresholds or trip requirements for individual plant variables or channels, The VACs provide voter requirements for groups of channels or plant variables as specified to initiate a trip condition. Continuous dynamic testing of instrument loops occurs by a programmed pattern of simulated trip/nontrip conditions, which exercise both software and hardware in the safety channel. The pattern recognition logic (PRL) is a hardware wired component programmed to maintain nontrip output only when this excepted time-dependent pattern is not changed. If a change occurs, as will happen if there is a plant trip condition or safety system failure - either hardware or software - then the PRL will initiate a trip condition. In summary, DSS provides for continuous dynamic testing of safety-related components and fail-safe operation. Through scenario testing of a DSS emulator on a boiling water reactor (BWR) plant training simulator it has been shown that DSS can provide a cost- effective safety system in BWR power plants. Experimental research has been completed that indicates the feasibility of extending DSS to include the plant nuclear instrumentation in the DSS test domain. This extension has the potential to decrease operating and maintenance (O&M) costs and improve fault diagnosis.

  1. Compressed gas fuel storage system

    DOEpatents

    Wozniak, John J.; Tiller, Dale B.; Wienhold, Paul D.; Hildebrand, Richard J.

    2001-01-01

    A compressed gas vehicle fuel storage system comprised of a plurality of compressed gas pressure cells supported by shock-absorbing foam positioned within a shape-conforming container. The container is dimensioned relative to the compressed gas pressure cells whereby a radial air gap surrounds each compressed gas pressure cell. The radial air gap allows pressure-induced expansion of the pressure cells without resulting in the application of pressure to adjacent pressure cells or physical pressure to the container. The pressure cells are interconnected by a gas control assembly including a thermally activated pressure relief device, a manual safety shut-off valve, and means for connecting the fuel storage system to a vehicle power source and a refueling adapter. The gas control assembly is enclosed by a protective cover attached to the container. The system is attached to the vehicle with straps to enable the chassis to deform as intended in a high-speed collision.

  2. Development of tokamak reactor system analysis code NEW-TORSAC

    NASA Astrophysics Data System (ADS)

    Kasai, Masao; Ida, Toshio; Nishikawa, Masana; Kameari, Akihisa; Nishio, Satoshi; Tone, Tatsuzo

    1987-07-01

    A systems analysis code named NEW-TORSAC (TOkamak Reactor Systems Analysis Code) has been developed by modifying the TORSAC which had been already developed by us. The NEW-TORSAC is available for tokamak reactor designs and evaluations from experimental machines to commercial reactor plants. It has functions to design tokamaks automatically from plasma parameter setting to determining configurations of reactor equipments and calculating main characteristics parameters of auxiliary systems and the capital costs. In the case of analyzing tokamak reactor plants, the code can calculate busbar energy costs. In addition to numerical output, some output of this code such as a reactor configuration, plasma equilibrium, electro-magnetic forces, etc., are graphically displayed. The code has been successfully applied to the scoping studies of the next generation machines and commercial reactor plants.

  3. A novel reactor for the simulation of gas and ash interactions in volcanic eruption plumes

    NASA Astrophysics Data System (ADS)

    Ayris, Paul M.; Cimarelli, Corrado; Delmelle, Pierre; Dingwell, Donald B.

    2014-05-01

    The chemical interactions between volcanic ash and the atmosphere, hydrosphere, pedosphere, cryosphere and biosphere are initially the result of rapid mobilisation of soluble salts and aqueous acids from wetted particle surfaces. Such surface features are attributable to the scavenging of sulphur and halide species by ash during its transport through the eruption plume and volcanic cloud. It has been historically considered (e.g., Rose, 1977) that the primary mechanism driving scavenging of sulphur and halide species is via condensation of acid aerosols onto ash surfaces within the cold volcanic cloud. However, for large explosive eruptions, insights from new experimental highlight the potential for scavenging via adsorption onto ash within the high-temperature eruption plume. In previous investigations on simple SO2 (Ayris et al. 2013a) and HCl systems (Ayris et al. 2013b), we identified ash composition, and the duration and temperature of gas-ash interaction as key determinants of adsorption-mode scavenging. However, the first generation of gas-ash reactors could not fully investigate the interactions between ash and the hydrous volcanic atmosphere; we have therefore developed an Advanced Gas Ash Reactor (AGAR), which can be fluxed with varying proportions of H2O, CO2, SO2 and HCl. The AGAR consists of a longitudinally-rotating quartz glass reaction bulb contained within a horizontal, three-stage tube furnace operating at temperatures of 25-900° C. A sample mass of up to 100 g can traverse a thermal gradient via manual repositioning of the reaction bulb within the furnace. In combination with existing melt synthesis capabilities in our laboratories, this facility permits a detailed investigation of the effects of ash and gas composition, and temperature on in-plume scavenging of SO2 and HCl. Additionally, the longitudinal rotation enables particle-particle interaction under an 'in-plume' atmosphere, and may yield insight into the effects of gas-ash interaction

  4. Nuclear reactor cooling system decontamination reagent regeneration

    DOEpatents

    Anstine, Larry D.; James, Dean B.; Melaika, Edward A.; Peterson, Jr., John P.

    1985-01-01

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  5. Cavity temperature and flow characteristics in a gas-core test reactor

    NASA Technical Reports Server (NTRS)

    Putre, H. A.

    1973-01-01

    A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

  6. Principle and Performance of Gas Self-inducing Reactors and Applications to Biotechnology.

    PubMed

    Ye, Qin; Li, Zhimin; Wu, Hui

    2016-01-01

    Gas-liquid contacting is an important unit operation in chemical and biochemical processes, but the gas utilization efficiency is low in conventional gas-liquid contactors especially for sparingly soluble gases. The gas self-inducing impeller is able to recycle gas in the headspace of a reactor to the liquid without utilization of additional equipment such as a gas compressor, and thus, the gas utilization efficiency is significantly enhanced. Gas induction is caused by the low pressure or deep vortex at a sufficiently high impeller speed, and the speed at which gas induction starts is termed the critical speed. The critical impeller speed, gas-induction flow rate, power consumption, and gas-liquid mass transfer are determined by the impeller design and operation conditions. When the reactor is operated in a dead-end mode, all the introduced gas can be completely used, and this feature is especially favorable to flammable and/or toxic gases. In this article, the principles, designs, characteristics of self-inducing reactors, and applications to biotechnology are described.

  7. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John b. Walter

    2010-10-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  8. Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

    SciTech Connect

    Zhang Zuoyi; Dong Yujie; Scherer, Winfried

    2005-03-15

    Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was <0.3 kg/(m{sup 3}s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to {approx}3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged.

  9. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  10. Proceedings of a Symposium on Advanced Compact Reactor Systems

    NASA Technical Reports Server (NTRS)

    1983-01-01

    Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

  11. Systems aspects of a space nuclear reactor power system

    NASA Technical Reports Server (NTRS)

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Bloomfield, H.; Heller, J.

    1988-01-01

    Various system aspects of a 300-kW nuclear reactor power system for spacecraft have been investigated. Special attention is given to the cases of a reusable OTV and a space-based radar. It is demonstrated that the stowed length of the power system is important to mission design, and that orbital storage for months to years may be needed for missions involving orbital assembly.

  12. Process and system for removing impurities from a gas

    DOEpatents

    Henningsen, Gunnar; Knowlton, Teddy Merrill; Findlay, John George; Schlather, Jerry Neal; Turk, Brian S

    2014-04-15

    A fluidized reactor system for removing impurities from a gas and an associated process are provided. The system includes a fluidized absorber for contacting a feed gas with a sorbent stream to reduce the impurity content of the feed gas; a fluidized solids regenerator for contacting an impurity loaded sorbent stream with a regeneration gas to reduce the impurity content of the sorbent stream; a first non-mechanical gas seal forming solids transfer device adapted to receive an impurity loaded sorbent stream from the absorber and transport the impurity loaded sorbent stream to the regenerator at a controllable flow rate in response to an aeration gas; and a second non-mechanical gas seal forming solids transfer device adapted to receive a sorbent stream of reduced impurity content from the regenerator and transfer the sorbent stream of reduced impurity content to the absorber without changing the flow rate of the sorbent stream.

  13. System of treating flue gas

    DOEpatents

    Ziegler, D.L.

    1975-12-01

    A system is described for treating or cleaning incinerator flue gas containing acid gases and radioactive and fissionable contaminants. Flue gas and a quench solution are fed into a venturi and then tangentially into the lower portion of a receptacle for restricting volumetric content of the solution. The upper portion of the receptacle contains a scrub bed to further treat or clean the flue gas.

  14. Advanced High Temperature Reactor Systems and Economic Analysis

    SciTech Connect

    Holcomb, David Eugene; Peretz, Fred J; Qualls, A L

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience with

  15. Testing of an advanced thermochemical conversion reactor system

    SciTech Connect

    Not Available

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

  16. Adaptive polynomial chaos techniques for uncertainty quantification of a gas cooled fast reactor transient

    SciTech Connect

    Perko, Z.; Gilli, L.; Lathouwers, D.; Kloosterman, J. L.

    2013-07-01

    Uncertainty quantification plays an increasingly important role in the nuclear community, especially with the rise of Best Estimate Plus Uncertainty methodologies. Sensitivity analysis, surrogate models, Monte Carlo sampling and several other techniques can be used to propagate input uncertainties. In recent years however polynomial chaos expansion has become a popular alternative providing high accuracy at affordable computational cost. This paper presents such polynomial chaos (PC) methods using adaptive sparse grids and adaptive basis set construction, together with an application to a Gas Cooled Fast Reactor transient. Comparison is made between a new sparse grid algorithm and the traditionally used technique proposed by Gerstner. An adaptive basis construction method is also introduced and is proved to be advantageous both from an accuracy and a computational point of view. As a demonstration the uncertainty quantification of a 50% loss of flow transient in the GFR2400 Gas Cooled Fast Reactor design was performed using the CATHARE code system. The results are compared to direct Monte Carlo sampling and show the superior convergence and high accuracy of the polynomial chaos expansion. Since PC techniques are easy to implement, they can offer an attractive alternative to traditional techniques for the uncertainty quantification of large scale problems. (authors)

  17. Hydrogasification reactor and method of operating same

    DOEpatents

    Hobbs, Raymond; Karner, Donald; Sun, Xiaolei; Boyle, John; Noguchi, Fuyuki

    2013-09-10

    The present invention provides a system and method for evaluating effects of process parameters on hydrogasification processes. The system includes a hydrogasification reactor, a pressurized feed system, a hopper system, a hydrogen gas source, and a carrier gas source. Pressurized carbonaceous material, such as coal, is fed to the reactor using the carrier gas and reacted with hydrogen to produce natural gas.

  18. Combustion-gas recirculation system

    DOEpatents

    Baldwin, Darryl Dean

    2007-10-09

    A combustion-gas recirculation system has a mixing chamber with a mixing-chamber inlet and a mixing-chamber outlet. The combustion-gas recirculation system may further include a duct connected to the mixing-chamber inlet. Additionally, the combustion-gas recirculation system may include an open inlet channel with a solid outer wall. The open inlet channel may extend into the mixing chamber such that an end of the open inlet channel is disposed between the mixing-chamber inlet and the mixing-chamber outlet. Furthermore, air within the open inlet channel may be at a pressure near or below atmospheric pressure.

  19. High-temperature gas-cooled reactors: preliminary safety and environmental information document. Volume IV

    SciTech Connect

    Not Available

    1980-01-01

    Information is presented concerning medium-enriched uranium/thorium once-through fuel cycle; medium-enrichment uranium-233/thorium recycle fuel; high-enrichment uranium-235/thorium recycle (spiked) fuel cycle; high-enrichment uranium-233/thorium recycle (spiked) fuel cycle; and gas-turbine high-temperature gas-cooled reactor.

  20. APPARATUS FOR SHORT TIME MEASUREMENTS IN A FIXED-BED, GAS/SOLID REACTOR

    EPA Science Inventory

    An apparatus for exposure of a solid to reactive process gas is described which makes possible short time (≥ 0.3 to 15 s) exposures in a fixed-bed reactor. Operating conditions for differential reaction with respect to the gas concentration and rapid quench for arresting hi...

  1. Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    SciTech Connect

    Silady, F.A.; Millunzi, A.C.

    1989-08-01

    The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab.

  2. Nuclear reactor fuel rod attachment system

    DOEpatents

    Christiansen, David W.

    1982-01-01

    A reusable system for removably attaching a nuclear reactor fuel rod (12) to a support member (14). A locking cap (22) is secured to the fuel rod (12) and a locking strip (24) is fastened to the support member (14). The locking cap (22) has two opposing fingers (24a and 24b) shaped to form a socket having a body portion (26). The locking strip has an extension (36) shaped to rigidly attach to the socket's body portion (26). The locking cap's fingers are resiliently deflectable. For attachment, the locking cap (22) is longitudinally pushed onto the locking strip (24) causing the extension (36) to temporarily deflect open the fingers (24a and 24b) to engage the socket's body portion (26). For removal, the process is reversed.

  3. The effect of water vapor in the reactor cavity in a MHTGR (Modular High Temperature Gas Cooled Reactor) on the radiation heat transfer

    SciTech Connect

    Cappiello, M.W.

    1991-01-01

    Analyses have been completed to determine the effect of the presence of water vapor in the reactor cavity in a modular high temperature gas cooled reactor on the predicted radiation heat transfer from the vessel wall to the reactor cavity cooling system. The analysis involves the radiation heat transfer between two parallel plates with an absorbing and emitting medium present. Because the absorption in the water vapor is spectrally dependent, the solution is difficult even for simple geometries. A computer code was written to solve the problem using the Monte Carlo method. The code was validated against closed form solutions, and shows excellent agreement. In the analysis of the reactor problem, the results show that the reduction in heat transfer, and the consequent increase in the vessel wall temperature, can be significant. This effect can be cast in terms of a reduction in the wall surface emissivities from 0.8 to 0.59. Because of the insulating effect of the water vapor, increasing the gap distance between the vessel wall and the cooling system will cause the vessel wall temperature to increase further. Care should be taken in the design of the facility to minimize the gap distance and keep temperature increase within allowable limits. 3 refs., 6 figs., 4 tabs.

  4. Composite nuclear fuel fabrication methodology for gas fast reactors

    NASA Astrophysics Data System (ADS)

    Vasudevamurthy, Gokul

    An advanced fuel form for use in Gas Fast Reactors (GFR) was investigated. Criteria for the fuel includes operation at high temperature (˜1400°C) and high burnup (˜150 MWD/MTHM) with effective retention of fission products even during transient temperatures exceeding 1600°C. The GFR fuel is expected to contain up to 20% transuranics for a closed fuel cycle. Earlier evaluations of reference fuels for the GFR have included ceramic-ceramic (cercer) dispersion type composite fuels of mixed carbide or nitride microspheres coated with SiC in a SiC matrix. Studies have indicated that ZrC is a potential replacement for SiC on account of its higher melting point, increased fission product corrosion resistance and better chemical stability. The present work investigated natural uranium carbide microspheres in a ZrC matrix instead of SiC. Known issues of minor actinide volatility during traditional fabrication procedures necessitated the investigation of still high temperature but more rapid fabrication techniques to minimize these anticipated losses. In this regard, fabrication of ZrC matrix by combustion synthesis from zirconium and graphite powders was studied. Criteria were established to obtain sufficient matrix density with UC microsphere volume fractions up to 30%. Tests involving production of microspheres by spark erosion method (similar to electrodischarge machining) showed the inability of the method to produce UC microspheres in the desired range of 300 to 1200 mum. A rotating electrode device was developed using a minimum current of 80A and rotating at speeds up to 1500 rpm to fabricate microspheres between 355 and 1200 mum. Using the ZrC process knowledge, UC electrodes were fabricated and studied for use in the rotating electrode device to produce UC microspheres. Fabrication of the cercer composite form was studied using microsphere volume fractions of 10%, 20%, and 30%. The macrostructure of the composite and individual components at various stages were

  5. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    NASA Astrophysics Data System (ADS)

    Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

    2006-01-01

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  6. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    SciTech Connect

    Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

    2006-01-20

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  7. Reactor/Brayton power systems for nuclear electric spacecraft

    NASA Technical Reports Server (NTRS)

    Layton, J. P.

    1980-01-01

    Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. This vehicle would be capable of performing detailed exploration of the outer planets of the solar system during the remainder of this century. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system. The results have shown that the performance is very competitive and that a 400 kWe space power system is dimensionally compatible with a single Space Shuttle launch. Performance parameters of system mass and radiator area were determined for systems from 100 to 1000 kWe. A 400 kWe reference system received primary attention. The components of this system were defined and a conceptual layout was developed with encouraging results. The preliminary mass determination for the complete power system was very close to the desired goal of 20 kg/kWe. Use of more advanced technology (higher turbine inlet temperature) will substantially improve system performance characteristics.

  8. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    NASA Technical Reports Server (NTRS)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2010-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  9. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    NASA Technical Reports Server (NTRS)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2009-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  10. ATWS Transients for the 2400 MWt Gas-Cooled Fast Reactor

    SciTech Connect

    Cheng,L.Y.; Ludewig, H.

    2007-08-05

    Reactivity transients have been analyzed with an updated RELAPS-3D (ver. 2.4.2) system model of the pin core design for the 2400MWt gas-cooled fast reactor (GCFR). Additional reactivity parameters were incorporated in the RELAP5 point-kinetics model to account for reactivity feedbacks due to axial and radial expansion of the core, fuel temperature changes (Doppler effect), and pressure changes (helium density changes). Three reactivity transients without scram were analyzed and the incidents were initiated respectively by reactivity ramp, loss of load, and depressurization. During the course of the analysis the turbine bypass model for the power conversion unit (PCU) was revised to enable a better utilization of forced flow cooling after the PCU is tripped. The analysis of the reactivity transients demonstrates the significant impact of the PCU on system pressure and core flow. Results from the modified turbine bypass model suggest a success path for the GCFR to mitigate reactivity transients without scram.

  11. Unconventional shallow biogenic gas systems

    USGS Publications Warehouse

    Shurr, G.W.; Ridgley, J.L.

    2002-01-01

    Unconventional shallow biogenic gas falls into two distinct systems that have different attributes. Early-generation systems have blanketlike geometries, and gas generation begins soon after deposition of reservoir and source rocks. Late-generation systems have ringlike geometries, and long time intervals separate deposition of reservoir and source rocks from gas generation. For both types of systems, the gas is dominantly methane and is associated with source rocks that are not thermally mature. Early-generation biogenic gas systems are typified by production from low-permeability Cretaceous rocks in the northern Great Plains of Alberta, Saskatchewan, and Montana. The main area of production is on the southeastern margin of the Alberta basin and the northwestern margin of the Williston basin. The huge volume of Cretaceous rocks has a generalized regional pattern of thick, non-marine, coarse clastics to the west and thinner, finer grained marine lithologies to the east. Reservoir rocks in the lower part tend to be finer grained and have lower porosity and permeability than those in the upper part. Similarly, source beds in the units have higher values of total organic carbon. Patterns of erosion, deposition, deformation, and production in both the upper and lower units are related to the geometry of lineament-bounded basement blocks. Geochemical studies show that gas and coproduced water are in equilibrium and that the fluids are relatively old, namely, as much as 66 Ma. Other examples of early-generation systems include Cretaceous clastic reservoirs on the southwestern margin of Williston basin and chalks on the eastern margin of the Denver basin. Late-generation biogenic gas systems have as an archetype the Devonian Antrim Shale on the northern margin of the Michigan basin. Reservoir rocks are fractured, organic-rich black shales that also serve as source rocks. Although fractures are important for production, the relationships to specific geologic structures are

  12. Results of theoretical and experimental studies of hydrodynamics of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors

    NASA Astrophysics Data System (ADS)

    Ryabov, G. A.; Folomeev, O. M.; Sankin, D. A.; Melnikov, D. A.

    2015-02-01

    Problems of the calculation of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors (polygeneration systems for the production of electricity, heat, and useful products and chemical cycles of combustion and gasification of solid fuels)are considered. A method has been developed for the calculation of circulation loop of fuel particles with respect to boilers with circulating fluidized bed (CFB) and systems with interconnected reactors with fluidized bed (FB) and CFB. New dependences for the connection between the fluidizing agent flow (air, gas, and steam) and performance of reactors and for the whole system (solids flow rate, furnace and cyclone pressure drops, and bed level in the riser) are important elements of this method. Experimental studies of hydrodynamics of circulation loops on the aerodynamic unit have been conducted. Experimental values of pressure drop of the horizontal part of the L-valve, which satisfy the calculated dependence, have been obtained.

  13. REACTOR - a Concept for establishing a System-of-Systems

    NASA Astrophysics Data System (ADS)

    Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim

    2014-05-01

    REACTOR is a working title for activities implementing reliable, emergent, adaptive, and concurrent collaboration on the basis of transactional object repositories. It aims at establishing federations of autonomous yet interoperable systems (Systems-of-Systems), which are able to expose emergent behaviour. Following the principles of event-driven service-oriented architectures (SOA 2.0), REACTOR enables adaptive re-organisation by dynamic delegation of responsibilities and novel yet coherent monitoring strategies by combining information from different domains. Thus it allows collaborative decision-processes across system, discipline, and administrative boundaries. Interoperability is based on two approaches that implement interconnection and communication between existing heterogeneous infrastructures and information systems: Coordinated (orchestration-based) communication and publish/subscribe (choreography-based) communication. Choreography-based communication ensures the autonomy of the participating systems to the highest possible degree but requires the implementation of adapters, which provide functional access to information (publishing/consuming events) via a Message Oriented Middleware (MOM). Any interconnection of the systems (composition of service and message cascades) is established on the basis of global conversations that are enacted by choreographies specifying the expected behaviour of the participating systems with respect to agreed Service Level Agreements (SLA) required by e.g. national authorities. The specification of conversations, maintained in commonly available repositories also enables the utilisation of systems for purposes (evolving) other than initially intended. Orchestration-based communication additionally requires a central component that controls the information transfer via service requests or event processing and also takes responsibility of managing business processes. Commonly available transactional object repositories are

  14. Development and investigations of compact heat-transfer equipment for a nuclear power station equipped with a high-temperature gas-cooled reactor

    NASA Astrophysics Data System (ADS)

    Golovko, V. F.; Dmitrieva, I. V.; Kodochigov, N. G.; Bykh, O. A.

    2013-07-01

    The project of a nuclear power station the reactor coolant system of which includes a high-temperature gas-cooled reactor combined with a gas-turbine energy conversion unit supposes the use of high-efficient gas-cycle-based heat-transfer equipment. An analysis aimed at selecting the optimal heat-transfer surfaces is presented together with the results from their calculated and experimental investigation. The design features of recuperators arranged integrally with end and intermediate coolers and placed in a vertical sealed high-pressure vessel of limited sizes are considered.

  15. Liquid metal cooled nuclear reactors with passive cooling system

    DOEpatents

    Hunsbedt, Anstein; Fanning, Alan W.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

  16. Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

    NASA Technical Reports Server (NTRS)

    Bragg-Sitton, S. M.; Webster, K. L.

    2007-01-01

    Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

  17. Exhaust system with emissions storage device and plasma reactor

    DOEpatents

    Hoard, John W.

    1998-01-01

    An exhaust system for a combustion system, comprising a storage device for collecting NO.sub.x, hydrocarbon, or particulate emissions, or mixture of these emissions, and a plasma reactor for destroying the collected emissions is described. After the emission is collected in by the storage device for a period of time, the emission is then destroyed in a non-thermal plasma generated by the plasma reactor. With respect to the direction of flow of the exhaust stream, the storage device must be located before the terminus of the plasma reactor, and it may be located wholly before, overlap with, or be contained within the plasma reactor.

  18. Compact, high-power nuclear reactor systems based on small diameter particulate fuel

    NASA Astrophysics Data System (ADS)

    Powell, J. R.; Botts, T. E.

    Two compact, high-power nuclear reactor concepts are discussed. Both are gas-cooled cavity-type reactors which utilize particulate fuel of the type now used in HTGR reactors. Unshielded reactor volumes are on the order of one cubic meter. The Fixed Bed Reactor operating temperature is limited to 2500 K and the output power to 250 MW(e). In the Rotating Bed Reactor fuel is held within a rotating porous metal drum as a rotating fluidized bed. Rotating Bed Reactor outlet temperatures up to 3000 K and output power levels up to 1000 MW(e) are achievable. Both reactors can be brought up from stand by to full power in times on the order of a few seconds, due primarily to the short thermal time constant for the fuel particles. Turbine and MHD Brayton are the power conversion cycles of choice. Open cycle operation is generally favored for applications operating at less than 1000 sec of equivalent integrated full power. At power levels above 1 MW(e), the liquid droplet radiator is the favored means of heat rejection. Power system specific power levels of 10 kW(e)/kg (not including shield) appears to be quite feasible.

  19. An automated gas expansion system

    SciTech Connect

    Abercrombie, K.

    1993-01-01

    The Metrology Laboratory at the Rocky Flats Plant has constructed a new Vacuum Gauge Calibration System based on gas expansion. The system is used to calibrate vacuum pressure gauges between 1 mTorr and 1000 mTorr. The paper discusses an overview of the system including layout, software, testing and performance.

  20. An automated gas expansion system

    SciTech Connect

    Abercrombie, K.

    1993-05-01

    The Metrology Laboratory at the Rocky Flats Plant has constructed a new Vacuum Gauge Calibration System based on gas expansion. The system is used to calibrate vacuum pressure gauges between 1 mTorr and 1000 mTorr. The paper discusses an overview of the system including layout, software, testing and performance.

  1. Measurement of Turbulent Flow Phenomena for the Lower Plenum of a Prismatic Gas-Cooled Reactor

    SciTech Connect

    Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink; Keith G. Condie; Glenn E. McCreery

    2007-09-01

    Mean velocity field and turbulence data are presented for flow phenomena in a lower plenum of a typical prismatic gas-cooled reactor (GCR), such as in a Very High Temperature Reactor (VHTR) concept. In preparation for design, safety analyses and licensing, research has begun on readying the computational tools that will be needed to predict the thermal-hydraulics behavior of the reactor design. Fluid dynamics experiments have been designed and built to develop benchmark databases for the assessment of computational fluid dynamics (CFD) codes and their turbulence models for a typical VHTR plenum geometry in the limiting case of negligible buoyancy and constant fluid properties. This experiment has been proposed as a “Standard Problem” for assessing advanced reactor (CFD) analysis tools. Present results concentrate on the region of the plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum can locally be considered as multiple jets into a confined cross flow - with obstructions. A model of the lower plenum has been fabricated and scaled to the geometric dimensions of the Next Generation Nuclear Plant (NGNP) Point Design. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to induce flow features somewhat comparable to those expected from the staggered parallel rows of posts in the reactor design. Posts, side walls and end walls are fabricated from clear, fused quartz to match the refractive-index of the working fluid so that optical techniques may be employed for the measurements. The experiments were conducted in the Matched-Index-of-Refraction (MIR) Facility at the Idaho National Laboratory (INL). The benefit of the MIR technique is that it permits optical measurements to determine complex flow characteristics in passages and around objects to be obtained without locating a disturbing transducer in the flow field and without distortion of the optical paths. The

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

    SciTech Connect

    Lindemer, T.B.

    2002-09-05

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

  3. Gas House Autonomous System Monitoring

    NASA Technical Reports Server (NTRS)

    Miller, Luke; Edsall, Ashley

    2015-01-01

    Gas House Autonomous System Monitoring (GHASM) will employ Integrated System Health Monitoring (ISHM) of cryogenic fluids in the High Pressure Gas Facility at Stennis Space Center. The preliminary focus of development incorporates the passive monitoring and eventual commanding of the Nitrogen System. ISHM offers generic system awareness, adept at using concepts rather than specific error cases. As an enabler for autonomy, ISHM provides capabilities inclusive of anomaly detection, diagnosis, and abnormality prediction. Advancing ISHM and Autonomous Operation functional capabilities enhances quality of data, optimizes safety, improves cost effectiveness, and has direct benefits to a wide spectrum of aerospace applications.

  4. Thermionic switched self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M.; Shires, Charles D.; Brummond, William A.

    1989-01-01

    A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

  5. Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.

    SciTech Connect

    Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage

  6. METHODS AND RESULTS OF RECONSTRUCTION OF NOBLE GAS RELEASES FROM THE STACKS OF THE MAYAK PA GRAPHITE REACTORS OVER THE WHOLE PERIOD OF THEIR OPERATION

    SciTech Connect

    Glagolenko, Y. V.; Drozhko, Evgeniy G.; Mokrov, Y.; Pyatin, N. P.; Rovny, Sergey I.; Anspaugh, L. R.; Napier, Bruce A.

    2008-06-01

    Brief analysis of design features and operational modes of Mayak PA industrial graphite-uranium reactors (PUGRs) is given. The above mentioned Mayak PA PUGRs determined the rates of releases of radioactive noble gases (RNG) from activation (41Ar) and fission (isotopes of Krypton and Xenon) through the vent stack of the reactor. Information is given on methods and results of experimental determination of RNG atmospheric releases for the period starting from 1965 till PUGRs decommissioning in 1987-1990. A calculation method for reconstruction of radioactive noble gas releases is proposed and justified. The results of reconstruction are given. It is shown that maximum rates of RNG releases from PUGRs high stacks were observed in the 1950s, when ordinary atmospheric air was used as a cover gas for the reactor graphite stacks and gas purification systems (flow-type gas holders) had not been installed yet.

  7. Phonological studies of the new gas-induced agitated reactor using computational fluid dynamics.

    PubMed

    Yang, T C; Hsu, Y C; Wang, S F

    2001-06-01

    An ozone-induced agitated reactor has been found to be very effective in degrading industrial wastewater. However, the cost of the ozone generation as well as its short residence time in reactors has restricted its application in a commercial scale. An innovated gas-induced draft tube installed inside a conventional agitated reactor was proved to effectively retain the ozone in a reactor. The setup was demonstrated to significantly promote the ozone utilization rate up to 96% from the conventional rate of 60% above the onset speed. This work investigates the mixing mechanism of an innovated gas-induced reactor for the future scale-up design by using the technique of computational fluid dynamics. A three-dimensional flow model was proposed to compute the liquid-gas free surface as well as the flow patterns inside the reactor. The turbulent effects generated by two 45 degrees pitch-blade turbines were considered and the two phases mixing phenomena were also manipulated by the Eulerian-Eulerian techniques. The consistency of the free surface profiles and the fluid flow patterns proved a good agreement between computational results and the experimental observation. PMID:11482384

  8. The reactor-activator for gas-jet deposition of diamond structures

    NASA Astrophysics Data System (ADS)

    Rebrov, A. K.; Andreev, M. N.; B'yadovskiy, T. T.; Kubrak, K. V.; Yudin, I. B.

    2016-10-01

    This paper presents a new method of deposition of diamond and diamond like structures by using the original reactor for activation of carbonaceous gas mixtures. One of the important characteristics of the described reactor is the power efficiency owing to low heat losses. The new method of gas flow formation opens new possibilities to get important trade-offs (wide range of background conditions and others). The principal possibility of diamond deposition is proved by experiments with the flow of hydrogen and methane mixture in tungsten resistively heating cylindrical channels. Electron microphotographs and Raman spectrum of deposited structures prove the fact that deposited samples are diamond structures. Numerical study of the mixture flow through the reactor was carried for optimization of gas flow conducted to a substrate.

  9. Development of an Integrated Performance Model for TRISO-Coated Gas Reactor Particle Fuel

    SciTech Connect

    Petti, David Andrew; Miller, Gregory Kent; Martin, David George; Maki, John Thomas

    2005-05-01

    The success of gas reactors depends upon the safety and quality of the coated particle fuel. The understanding and evaluation of this fuel requires development of an integrated mechanistic fuel performance model that fully describes the mechanical and physico-chemical behavior of the fuel particle under irradiation. Such a model, called PARFUME (PARticle Fuel ModEl), is being developed at the Idaho National Engineering and Environmental Laboratory. PARFUME is based on multi-dimensional finite element modeling of TRISO-coated gas reactor fuel. The goal is to represent all potential failure mechanisms and to incorporate the statistical nature of the fuel. The model is currently focused on carbide, oxide nd oxycarbide uranium fuel kernels, while the coating layers are the classical IPyC/SiC/OPyC. This paper reviews the current status of the mechanical aspects of the model and presents results of calculations for irradiations from the New Production Modular High Temperature Gas Reactor program.

  10. Computer optimization of reactor-thermoelectric space power systems

    NASA Technical Reports Server (NTRS)

    Maag, W. L.; Finnegan, P. M.; Fishbach, L. H.

    1973-01-01

    A computer simulation and optimization code that has been developed for nuclear space power systems is described. The results of using this code to analyze two reactor-thermoelectric systems are presented.

  11. Backscatter absorption gas imaging system

    DOEpatents

    McRae, T.G. Jr.

    A video imaging system for detecting hazardous gas leaks. Visual displays of invisible gas clouds are produced by radiation augmentation of the field of view of an imaging device by radiation corresponding to an absorption line of the gas to be detected. The field of view of an imager is irradiated by a laser. The imager receives both backscattered laser light and background radiation. When a detectable gas is present, the backscattered laser light is highly attenuated, producing a region of contrast or shadow on the image. A flying spot imaging system is utilized to synchronously irradiate and scan the area to lower laser power requirements. The imager signal is processed to produce a video display.

  12. Backscatter absorption gas imaging system

    DOEpatents

    McRae, Jr., Thomas G.

    1985-01-01

    A video imaging system for detecting hazardous gas leaks. Visual displays of invisible gas clouds are produced by radiation augmentation of the field of view of an imaging device by radiation corresponding to an absorption line of the gas to be detected. The field of view of an imager is irradiated by a laser. The imager receives both backscattered laser light and background radiation. When a detectable gas is present, the backscattered laser light is highly attenuated, producing a region of contrast or shadow on the image. A flying spot imaging system is utilized to synchronously irradiate and scan the area to lower laser power requirements. The imager signal is processed to produce a video display.

  13. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high... 10 Energy 1 2012-01-01 2012-01-01 false Acceptance criteria for reactor coolant system venting systems. 50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION...

  14. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high... 10 Energy 1 2011-01-01 2011-01-01 false Acceptance criteria for reactor coolant system venting systems. 50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION...

  15. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high... 10 Energy 1 2014-01-01 2014-01-01 false Acceptance criteria for reactor coolant system venting systems. 50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION...

  16. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high... 10 Energy 1 2013-01-01 2013-01-01 false Acceptance criteria for reactor coolant system venting systems. 50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION...

  17. Optical fibre gas detections systems

    NASA Astrophysics Data System (ADS)

    Culshaw, Brian

    2016-05-01

    This tutorial review covers the principles of and prospects for fibre optic sensor technology in gas detection. Many of the potential benefits common to fibre sensor technology also apply in the context of gas sensing - notably long distance - many km - access to multiple remote measurement points; invariably intrinsic safety; access to numerous important gas species and often uniquely high levels of selectivity and/or sensitivity. Furthermore, the range of fibre sensor network architectures - single point, multiple point and distributed - enable unprecedented flexibility in system implementation. Additionally, competitive technologies and regulatory issues contribute to final application potential.

  18. Gaseous fission product management for molten salt reactors and vented fuel systems

    SciTech Connect

    Messenger, S. J.; Forsberg, C.; Massie, M.

    2012-07-01

    Fission gas disposal is one of the unresolved difficulties for Molten Salt Reactors (MSRs) and advanced reactors with vented fuel systems. As these systems operate, they produce many radioactive isotopes of xenon and krypton (e.g. {sup 135}Xe t{sub 1/2} = 9.14 hours and {sup 85}Kr t{sub 1/2}= 10.73 years). Removing these gases proves vital to the success of such reactor designs for two reasons. First, the gases act as large neutron sinks which decrease reactivity and must be counterbalanced by increasing fuel loading. Second, for MSRs, inert fission product gases naturally separate quickly from high temperature salts, thus creating high vapor pressure which poses safety concerns. For advanced reactors with solid vented fuel, the gases are allowed to escape into an off-gas system and thus must be managed. Because of time delays in transport of fission product gases in vented fuel systems, some of the shorter-lived radionuclides will decay away thereby reducing the fission gas source term relative to an MSR. To calculate the fission gas source term of a typical molten salt reactor, we modeled a 1000 MWe graphite moderated thorium MSR similar to that detailed in Mathieu et al. [1]. The fuel salt used in these calculations was LiF (78 mole percent) - (HN)F 4 (22 mole percent) with a heavy nuclide composition of 3.86% {sup 233}U and 96.14% {sup 232}Th by mass. Before we can remove the fission product gases produced by this reactor configuration, we must first develop an appropriate storage mechanism. The gases could be stored in pressurized containers but then one must be concerned about bottle failure. Methods to trap noble gases in matrices are expensive and complex. Alternatively, there are direct storage/disposal options: direct injection into the Earth or injecting a grout-based product into the Earth. Advances in drilling technologies, hydro fracture technologies, and methods for the sequestration of carbon dioxide from fossil fuel plants are creating new options

  19. Fission Product Monitoring and Release Data for the Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John B. Walter; Jason M. Harp; Mark W. Drigert; Edward L. Reber

    2010-10-01

    The AGR-1 experiment is a fueled multiple-capsule irradiation experiment that was irradiated in the Advanced Test Reactor (ATR) from December 26, 2006 until November 6, 2009 in support of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Fuel Development and Qualification program. An important measure of the fuel performance is the quantification of the fission product releases over the duration of the experiment. To provide this data for the inert fission gasses(Kr and Xe), a fission product monitoring system (FPMS) was developed and implemented to monitor the individual capsule effluents for the radioactive species. The FPMS continuously measured the concentrations of various krypton and xenon isotopes in the sweep gas from each AGR-1 capsule to provide an indicator of fuel irradiation performance. Spectrometer systems quantified the concentrations of Kr-85m, Kr-87, Kr-88, Kr-89, Kr-90, Xe-131m, Xe-133, Xe 135, Xe 135m, Xe-137, Xe-138, and Xe-139 accumulated over repeated eight hour counting intervals.-. To determine initial fuel quality and fuel performance, release activity for each isotope of interest was derived from FPMS measurements and paired with a calculation of the corresponding isotopic production or birthrate. The release activities and birthrates were combined to determine Release-to-Birth ratios for the selected nuclides. R/B values provide indicators of initial fuel quality and fuel performance during irradiation. This paper presents a brief summary of the FPMS, the release to birth ratio data for the AGR-1 experiment and preliminary comparisons of AGR-1 experimental fuels data to fission gas release models.

  20. Fuel cell gas management system

    DOEpatents

    DuBose, Ronald Arthur

    2000-01-11

    A fuel cell gas management system including a cathode humidification system for transferring latent and sensible heat from an exhaust stream to the cathode inlet stream of the fuel cell; an anode humidity retention system for maintaining the total enthalpy of the anode stream exiting the fuel cell equal to the total enthalpy of the anode inlet stream; and a cooling water management system having segregated deionized water and cooling water loops interconnected by means of a brazed plate heat exchanger.

  1. The Anaesthesia Gas Supply System

    PubMed Central

    Das, Sabyasachi; Chattopadhyay, Subhrajyoti; Bose, Payel

    2013-01-01

    The anaesthesia gas supply system is designed to provide a safe, cost-effective and convenient system for the delivery of medical gases at the point of-use. The doctrine of the anaesthesia gas supply system is based on four essential principles: Identity, continuity, adequacy and quality. Knowledge about gas supply system is an integral component of safe anaesthetic practice. Mishaps involving the malfunction or misuse of medical gas supply to operating theatres have cost many lives. The medical gases used in anaesthesia and intensive care are oxygen, nitrous oxide, medical air, entonox, carbon dioxide and heliox. Oxygen is one of the most widely used gases for life-support and respiratory therapy besides anaesthetic procedures. In this article, an effort is made to describe the production, storage and delivery of anaesthetic gases. The design of anaesthesia equipment must take into account the local conditions such as climate, demand and power supply. The operational policy of the gas supply system should have a backup plan to cater to the emergency need of the hospital, in the event of the loss of the primary source of supply. PMID:24249882

  2. Fast-quench reactor for hydrogen and elemental carbon production from natural gas and other hydrocarbons

    DOEpatents

    Detering, Brent A.; Kong, Peter C.

    2006-08-29

    A fast-quench reactor for production of diatomic hydrogen and unsaturated carbons is provided. During the fast quench in the downstream diverging section of the nozzle, such as in a free expansion chamber, the unsaturated hydrocarbons are further decomposed by reheating the reactor gases. More diatomic hydrogen is produced, along with elemental carbon. Other gas may be added at different stages in the process to form a desired end product and prevent back reactions. The product is a substantially clean-burning hydrogen fuel that leaves no greenhouse gas emissions, and elemental carbon that may be used in powder form as a commodity for several processes.

  3. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    SciTech Connect

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  4. Monitoring circuit for reactor safety systems

    DOEpatents

    Keefe, Donald J.

    1976-01-01

    The ratio between the output signals of a pair of reactor safety channels is monitored. When ratio falls outside of a predetermined range, it indicates that one or more of the safety channels has malfunctioned.

  5. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    NASA Astrophysics Data System (ADS)

    Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

  6. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    SciTech Connect

    Meriyanti; Su'ud, Zaki; Rijal, K.; Zuhair; Ferhat, A.; Sekimoto, H.

    2010-06-22

    In this study a feasibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850 deg. C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticality was obtained for this reactor.

  7. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    SciTech Connect

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  8. Water gas shift reaction kinetics and reactor modeling for fuel cell grade hydrogen

    NASA Astrophysics Data System (ADS)

    Choi, Yongtaek; Stenger, Harvey G.

    The kinetics of the water gas shift reaction was studied to evaluate existing reaction mechanisms, test various rate expressions and simulate the performance in a methanol fuel processor for fuel cell applications. The reaction was carried out in a micro reactor testing unit using a commercial Sud-Chemie Cu/ZnO/Al 2O 3 catalyst between 120 and 250 °C with a range of feed rates and compositions. Using non-linear least squares optimization, the parameters in five rate expressions were fit to the experimental data. Based on a review of published work on the WGS reaction mechanism, our study found that a rate expression derived from a regenerative mechanism and another rate expression derived from adsorptive mechanism fit the experimental data equally well. Numerical integration of a one-dimensional PFR model was used for this parameter fitting. An empirical rate expression, rCO= kPCOPH 2O (1- β) with activation energy of 47.4 kJ/mol was also obtained from the experimental data. Reactor performance was simulated to determine catalyst loadings required to achieve specific CO conversions as a function of temperature and water feed rate. These results are useful in studying the design trade offs available to reformer systems.

  9. Fluid modeling and design of gas channels of solar non-stoichiometric redox reactor

    NASA Astrophysics Data System (ADS)

    Kedlaya, Aditya

    The present numerical study in FLUENT analyzes the fluid flow field within a solar powered reactor designed for syngas production. The thermochemical reactor is based on continuous cycling of cerium oxide (ceria) in a non-stoichiometric reduction/oxidation cycle. The reactor uses a hollow cylinder of porous ceria which rotates through a high-temperature zone, by exposure to concentrated sunlight and partially reduced in an inert atmosphere due to flow of the sweep gas (N2), and then through a lower temperature zone where the reduced ceria is re-oxidized with a flow of CO2 and/or H2O, to produce CO and/or H2. In terms of fluid flow modeling, the issue of crossover of species (leakage) within the reactor is critical for proper functioning of the reactor. The first part of the work relates to the geometry and placement of the inlet/outlet gas channels for the reactor optimized to minimize crossover of the species. This is done by conducting a parametric study of geometric variables associated with the inlet/outlet geometry. A simplified 2D fluid flow reactor model which incorporates multi-species flow is used for this study. Further, 2D and 3D reactor models which capture the internal structure more accurately are used to refine the inlet/outlet design. The optimized reactor model is found to have an O2 crossover of 2%-6% and oxidizer crossover of 8%-21% at different flow rates of the sweep gas and the oxidizer studied. In the second part of the work, the reactor model is simulated under varying test conditions. Different working conditions include morphologies of the reactive material, rotational speed of the ceria ring and the recuperator, flow rates of sweep gas and the oxidizer, types of oxidizer (CO2, H2O). The 3D reactor model is also tested using one, two and three discrete inlet/outlet ports and compared with slot configuration.

  10. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981

    SciTech Connect

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

  11. SP-100 Program: space reactor system and subsystem investigations

    SciTech Connect

    Harty, R.B.

    1983-09-30

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

  12. 46 CFR 184.240 - Gas systems.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 false Gas systems. 184.240 Section 184.240 Shipping COAST... CONTROL AND MISCELLANEOUS SYSTEMS AND EQUIPMENT Cooking and Heating § 184.240 Gas systems. Cooking systems using liquefied petroleum gas (LPG) and compressed natural gas (CNG) must meet the...

  13. 46 CFR 121.240 - Gas systems.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 4 2013-10-01 2013-10-01 false Gas systems. 121.240 Section 121.240 Shipping COAST... SYSTEMS AND EQUIPMENT Cooking and Heating § 121.240 Gas systems. Cooking systems using liquefied petroleum gas (LPG) and compressed natural gas (CNG) must meet the following requirements: (a) The...

  14. 46 CFR 184.240 - Gas systems.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Gas systems. 184.240 Section 184.240 Shipping COAST... CONTROL AND MISCELLANEOUS SYSTEMS AND EQUIPMENT Cooking and Heating § 184.240 Gas systems. Cooking systems using liquefied petroleum gas (LPG) and compressed natural gas (CNG) must meet the...

  15. 46 CFR 184.240 - Gas systems.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 false Gas systems. 184.240 Section 184.240 Shipping COAST... CONTROL AND MISCELLANEOUS SYSTEMS AND EQUIPMENT Cooking and Heating § 184.240 Gas systems. Cooking systems using liquefied petroleum gas (LPG) and compressed natural gas (CNG) must meet the...

  16. 46 CFR 121.240 - Gas systems.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 4 2014-10-01 2014-10-01 false Gas systems. 121.240 Section 121.240 Shipping COAST... SYSTEMS AND EQUIPMENT Cooking and Heating § 121.240 Gas systems. Cooking systems using liquefied petroleum gas (LPG) and compressed natural gas (CNG) must meet the following requirements: (a) The...

  17. 46 CFR 121.240 - Gas systems.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Gas systems. 121.240 Section 121.240 Shipping COAST... SYSTEMS AND EQUIPMENT Cooking and Heating § 121.240 Gas systems. Cooking systems using liquefied petroleum gas (LPG) and compressed natural gas (CNG) must meet the following requirements: (a) The...

  18. 46 CFR 184.240 - Gas systems.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 7 2014-10-01 2014-10-01 false Gas systems. 184.240 Section 184.240 Shipping COAST... CONTROL AND MISCELLANEOUS SYSTEMS AND EQUIPMENT Cooking and Heating § 184.240 Gas systems. Cooking systems using liquefied petroleum gas (LPG) and compressed natural gas (CNG) must meet the...

  19. 46 CFR 121.240 - Gas systems.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 false Gas systems. 121.240 Section 121.240 Shipping COAST... SYSTEMS AND EQUIPMENT Cooking and Heating § 121.240 Gas systems. Cooking systems using liquefied petroleum gas (LPG) and compressed natural gas (CNG) must meet the following requirements: (a) The...

  20. 46 CFR 184.240 - Gas systems.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 false Gas systems. 184.240 Section 184.240 Shipping COAST... CONTROL AND MISCELLANEOUS SYSTEMS AND EQUIPMENT Cooking and Heating § 184.240 Gas systems. Cooking systems using liquefied petroleum gas (LPG) and compressed natural gas (CNG) must meet the...

  1. 46 CFR 121.240 - Gas systems.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 4 2012-10-01 2012-10-01 false Gas systems. 121.240 Section 121.240 Shipping COAST... SYSTEMS AND EQUIPMENT Cooking and Heating § 121.240 Gas systems. Cooking systems using liquefied petroleum gas (LPG) and compressed natural gas (CNG) must meet the following requirements: (a) The...

  2. A fast shutdown system for SRS (Savannah River Site) reactors

    SciTech Connect

    Baumann, N.P.

    1990-01-01

    Power has been sharply reduced at Savannah River Site (SRS) reactors in large part to ensure that no bulk boiling occurs during hypothesized loss of coolant accidents. A fast shutdown system is essential to regain much of this lost power. Computations and experiments indicate that a He-3 injection system will serve this function. Instrumented tests of a full system are planned for early 1991 for one of the SRS reactors. 4 refs., 7 figs., 1 tab.

  3. Destruction of chlorobenzene and carbon tetrachloride in a two-stage molten salt oxidation reactor system.

    PubMed

    Yang, Hee-Chul; Cho, Yong-Jun; Eun, Hee-Chul; Kim, Eung-Ho

    2008-08-01

    Molten salt oxidation (MSO) is one of the promising alternative destruction technologies for chlorinated organics, because it is capable of trapping chlorine during organic destruction. This study investigated the characteristics of a two-stage MSO reactor system for the destruction of CCl(4) and C(6)H(5)Cl. Investigated parameters were the MSO reactor temperature (from 1023 K to 1223 K) and the excess oxidizing air feed rate (50% and 100%). The destruction of chlorinated solvents is substantial in the Li(2)CO(3)-Na(2)CO(3) eutectic molten salt, irrespective of the tested condition. However, further oxidation of CO, which is found to be the major destruction product, is not substantial due to the limited temperature and gas residence time in the MSO reactor. Increases in the reactor temperature as well as those in the oxidizing air feed rate consistently lead to decreased emissions of carbon monoxide. No significant influence of the MSO reactor operating condition on the chlorine capturing efficiency was found. Over 99.95% and 99.997% of the chlorine was captured in the hot MSO reactors during the C(6)H(5)Cl and CCl(4) destructions, respectively. This result suggests a relatively low potential of the MSO system in the recombination of chlorinated organics, when compared to a conventional incineration system. PMID:18501405

  4. Destruction of chlorobenzene and carbon tetrachloride in a two-stage molten salt oxidation reactor system.

    PubMed

    Yang, Hee-Chul; Cho, Yong-Jun; Eun, Hee-Chul; Kim, Eung-Ho

    2008-08-01

    Molten salt oxidation (MSO) is one of the promising alternative destruction technologies for chlorinated organics, because it is capable of trapping chlorine during organic destruction. This study investigated the characteristics of a two-stage MSO reactor system for the destruction of CCl(4) and C(6)H(5)Cl. Investigated parameters were the MSO reactor temperature (from 1023 K to 1223 K) and the excess oxidizing air feed rate (50% and 100%). The destruction of chlorinated solvents is substantial in the Li(2)CO(3)-Na(2)CO(3) eutectic molten salt, irrespective of the tested condition. However, further oxidation of CO, which is found to be the major destruction product, is not substantial due to the limited temperature and gas residence time in the MSO reactor. Increases in the reactor temperature as well as those in the oxidizing air feed rate consistently lead to decreased emissions of carbon monoxide. No significant influence of the MSO reactor operating condition on the chlorine capturing efficiency was found. Over 99.95% and 99.997% of the chlorine was captured in the hot MSO reactors during the C(6)H(5)Cl and CCl(4) destructions, respectively. This result suggests a relatively low potential of the MSO system in the recombination of chlorinated organics, when compared to a conventional incineration system.

  5. Measurement of Flow Phenomena in a Lower Plenum Model of a Prismatic Gas-Cooled Reactor

    SciTech Connect

    Hugh M. McIlroy, Jr.; Donald M. McEligot; Robert J. Pink

    2008-05-01

    Mean-velocity-field and turbulence data are presented that measure turbulent flow phenomena in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. The data were obtained in the Matched-Index-of-Refraction (MIR) facility at Idaho National Laboratory (INL) and are offered for assessing computational fluid dynamics (CFD) software. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. This paper reviews the experimental apparatus and procedures, presents a sample of the data set, and reviews the INL Standard Problem. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum consists of multiple jets injected into a confined cross flow - with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the mineral oil working fluid so that optical techniques may be employed for the measurements. The benefit of the MIR technique is that it permits optical measurements to determine flow characteristics in complex passages in and around objects to be obtained without locating intrusive transducers that will disturb the flow field and without distortion of the optical paths. An advantage of the INL system is its large size, leading to improved spatial and temporal resolution compared to similar facilities at smaller scales. A three-dimensional (3-D) Particle Image Velocimetry (PIV) system was used to collect the data. Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12

  6. Measurement of Flow Phenomena in a Lower Plenum Model of a Prismatic Gas-Cooled Reactor

    SciTech Connect

    Hugh M. McIlroy, Jr.; Doanld M. McEligot; Robert J. Pink

    2010-02-01

    Mean-velocity-field and turbulence data are presented that measure turbulent flow phenomena in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. The data were obtained in the Matched-Index-of-Refraction (MIR) facility at Idaho National Laboratory (INL) and are offered for assessing computational fluid dynamics (CFD) software. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum consists of multiple jets injected into a confined cross flow - with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate geometry scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the working fluid so that optical techniques may be employed for the measurements. The benefit of the MIR technique is that it permits optical measurements to determine flow characteristics in complex passages in and around objects to be obtained without locating intrusive transducers that will disturb the flow field and without distortion of the optical paths. An advantage of the INL system is its large size, leading to improved spatial and temporal resolution compared to similar facilities at smaller scales. A three-dimensional (3-D) Particle Image Velocimetry (PIV) system was used to collect the data. Inlet jet Reynolds numbers (based on the jet diameter and the time-mean bulk velocity) are approximately 4,300 and 12,400. Uncertainty analyses and a discussion of the standard problem are included. The measurements reveal developing, non-uniform, turbulent flow in the

  7. Measurement of Turbulent Flow Phenomena for the Lower Plenum of a Prismatic Gas-Cooled Reactor

    SciTech Connect

    Hugh M. McIlroy, Jr.; Donald M. McEligot; Robert J. Pink

    2010-02-01

    Mean velocity field and turbulence data are presented that measure turbulent flow phenomena in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics design (Gas-Turbine-Modular Helium Reactor). The datawere obtained in the Matched-Index-of-Refraction (MIR) facility at Idaho National Laboratory (INL) and are offered as a benchmark for assessing computational fluid dynamics (CFD) software. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The primary objective of this paper is to document the experiment and present a sample of the data set that has been established for this standard problem. Present results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). The flowin the lower plenum consists of multiple jets injected into a confined crossflow—with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. Posts, side walls and end walls are fabricated from clear, fused quartz to match the refractive index of the mineral oil working fluid so that optical techniques may be employed for the measurements. The benefit of the MIR technique is that it permits optical measurements to determine flow characteristics in complex passages and around objects to be obtained without locating intrusive transducers that will disturb the flow field and without distortion of the optical paths. An advantage of the INL system is its large size, leading to improved spatial and temporal resolution compared to similar facilities at smaller scales. A three-dimensional (3D) particle image velocimetry (PIV) system was used to collect the data. Inlet-jet Reynolds numbers (based on the hydraulic diameter of the jet

  8. Experimental Plans for Subsystems of a Shock Wave Driven Gas Core Reactor

    NASA Technical Reports Server (NTRS)

    Kazeminezhad, F.; Anghai, S.

    2008-01-01

    This Contractor Report proposes a number of plans for experiments on subsystems of a shock wave driven pulsed magnetic induction gas core reactor (PMI-GCR, or PMD-GCR pulsed magnet driven gas core reactor). Computer models of shock generation and collision in a large-scale PMI-GCR shock tube have been performed. Based upon the simulation results a number of issues arose that can only be addressed adequately by capturing experimental data on high pressure (approx.1 atmosphere or greater) partial plasma shock wave effects in large bore shock tubes ( 10 cm radius). There are three main subsystems that are of immediate interest (for appraisal of the concept viability). These are (1) the shock generation in a high pressure gas using either a plasma thruster or pulsed high magnetic field, (2) collision of MHD or gas dynamic shocks, their interaction time, and collision pile-up region thickness, and (3) magnetic flux compression power generation (not included here).

  9. Ambient Laboratory Coater for Advanced Gas Reactor Fuel Development

    SciTech Connect

    Duane D. Bruns; Robert M. Counce; Irma D. Lima Rojas

    2010-06-09

    this research is targeted at developing improved experimentally-based scaling relationships for the hydrodynamics of shallow, gas-spouted beds of dense particles. The work is motivated by the need to more effctively scale up shallow spouted beds used in processes such as in the coating of nuclear fuel particles where precise control of solids and gas circulation is critically important. Experimental results reported here are for a 50 mm diameter spouted bed containing two different types of bed solids (alumina and zirconia) at different static bed depths and fluidized by air and helium. Measurements of multiple local average pressures, inlet gas pressure fluctuations, and spout height were used to characterize the bed hydrodynamics for each operating condition. Follow-on studies are planned that include additional variations in bed size, particle properties, and fluidizing gas. The ultimate objective is to identify the most important non-dimensional hydrodynamic scaling groups and possible spouted-bed design correlations based on these groups.

  10. High pressure synthesis gas conversion

    SciTech Connect

    Not Available

    1992-01-01

    A high pressure gas phase fermentation system has been constructed for the biological production of ethanol from coal synthesis gas. The reactors in the system consist of a 650 mL continuous stirred tank reactor and a 1 L continuous column reactor. The reactors are designed for individual or dual operation in series or parallel, with continuous gas and liquid feed. The system is housed in a constant temperature, explosion-proof room, equipped with gas leak detectors.

  11. Adaptation of gas tagging for failed fuel identification in light water reactors

    SciTech Connect

    Lambert, J.D.B.; Gross, K.C.; Depiante, E.V.; Callis, E.L.; Egebrecht, P.M.

    1996-03-01

    This paper discusses experience with noble gas tagging and its adaptation to commercial reactors. It reviews the recent incidence of fuel failures in light water reactors, and methods used to identify failures, and concludes that the on-line technique of gas tagging could significantly augment present flux tilting, sipping and ultrasonic testing of assemblies. The paper describes calculations on tag gas stability in-reactor, and tag injection tests that were carried out collaboratively with Commonwealth Edison Company in the Byron-2 pressurized water reactor (P%a) and with Duke Power Company and Babcock and Wilcox Fuel Company in the Oconee-2 PWM. The tests gave information on: (a) noble gas concentration dynamics as the tag gases were dissolved in and eventually removed from subsystems of the RCS; and (b) the suitability of candidate Ar, Ne, Kr and Xe isotopes for tagging PWR fuel. It was found that the activity of Xe{sup 125} (the activation product of the tag isotope Xe{sup 124}) acted as a ``tag of a tag`` and tracked gas through the reactor; measured activities are being used to model gas movement in the RCS. Several interference molecules (trace contaminants normally present at sub-ppM concentrations in RCS samples) and entrained air in the RCS were found to affect mass spectrometer sensitivity for tag isotopes. In all instances the contaminants could be differentiated from the tag isotopes by operating the mass spectrometer at high resolution (2500). Similarly, it was possible to distinguish all the candidate tag gases against a high background of air. The test results suggested, however, that for routine analysis a high resolution static mass spectrometer will be preferable to the dynamic instrument used for the present analyses.

  12. Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems

    NASA Astrophysics Data System (ADS)

    Was, Gary S.

    2007-08-01

    The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems.

  13. Transient thermal analysis of a space reactor power system

    SciTech Connect

    Gaeta, M.J.; Best, F.R. . Dept. of Nuclear Engineering)

    1993-07-01

    Space nuclear power systems utilize materials and processes that are completely different from terrestrial reactor systems. Therefore, the tools used to analyze ground-based systems are inappropriate for space reactor design and analysis. The purpose of this study was to develop a space reactor transient analysis tool and to apply this tool to scenarios of interest. The scope of the simulation includes the thermal and neutronic behavior of a liquid-metal-cooled fast reactor, the electrical and thermal performance of the thermoelectric generators, the thermal dynamics of heat pipe radiators, and the thermal behavior of the coolant piping between major components. The thermal model of the system is explicitly coupled to a momentum model of the primary and secondary coolant loops. A one-dimensional conduction model is employed in all solid component models. The reactor model includes an expression for energy generation due to fission and decay heat. The thermoelectric heat exchanger model accounts for thermal energy conversion to useful electrical output. The two-node radiator heat pipe model includes normal operation as well as limited heat pipe operation under sonic limit conditions. The reactor, thermoelectric heat exchanger, and heat pipe models are coupled explicitly by the coolant piping thermal model. The computer program is used to simulate a variety of transients including reactor power changer, degradation of the radiator, and a temporary open circuit condition on the thermoelectrics.

  14. Robotic system for remote maintenance of a pulsed nuclear reactor

    SciTech Connect

    Thunborg, S.

    1986-01-01

    Guidelines recently established for occupational radiation exposure specify that exposure should be as low as reasonably achievable. In conformance with these guidelines, SNL has developed a remote maintenance robot (RMR) system for use in the Sandia Pulse Reactor III (SPR III) facility. The RMR should reduce occupational radiation exposure by a factor of 4 and decrease reactor downtime. Other goals include developing a technology base for a more advanced pulse reactor and for the nuclear fuel cycle programs of the US Department of Energy and US Nuclear Regulatory Commission. The RMR has five major subsystems: (a) a chain-driven cart to bring the system into the reactor room; (b) a Puma 560 robot to perform dextrous operations; (c) a programmable turntable to orient the robot to any of the reactor's four sides; (d) a programmable overhead hoist for lifting components weighing up to 400 lb onto or off of the reactor; and (e) a supervisory control console for the system operator. Figure 1 is a schematic diagram of the turntable, hoist, and robot system in position around the SPR III reactor.

  15. a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

    NASA Astrophysics Data System (ADS)

    Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

    2009-08-01

    This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

  16. The characteristics of gas-solid flow and wall heat transfer in a fluidized bed reactor

    NASA Astrophysics Data System (ADS)

    Choi, Hang Seok; Meier, Dietrich

    2012-09-01

    Numerical study using computational fluid dynamics has been carried out to investigate the heat transfer characteristics of a laboratory fluidized bed reactor. The fluidized bed reactor of vTI (Johann Heinrich von Thünen-Institute)-Institute of Wood Technology and Wood Biology is modeled. For the simulation of multiphase flow and thermal fields, an Eulerian-Eulerian approach is applied. The flow and thermal characteristics of the reactor are fully investigated for the wide range of superficial gas velocities and two different particle diameters. In particular, the contributions of the gas bubble and emulsion phase flows on the wall heat transfer are scrutinized. From the predicted results, it is fully elucidated that particular near-wall bubble motions mainly govern the wall heat transfer.

  17. Simple automatic device for real time sampling of gas production by a reactor

    SciTech Connect

    Frattolillo, A.

    2006-06-15

    An innovative automatic device, allowing periodically drawing samples of the gases produced by a generic reactor, is presented. The gases evolving during the reaction are collected in a storage manifold, equipped with a variable volume consisting of a stainless steel bellow, whose expansion or contraction is driven by a linear step motor. A capacitive gauge monitors the pressure inside the storage manifold, while a feedback control loop reacts to any pressure change adjusting the variable volume (by means of the step motor) in such a way to keep the pressure at a desired set point P{sub 0}. As long as the reaction proceeds, the gas production results in a progressive expansion of the variable volume, whose instantaneous value is constantly monitored by means of a slide potentiometer, whose lever is rigidly connected to the bellow's moving extremity. Once the bellow's expansion has reached a predetermined volume increment {delta}V, which means that an amount of gas P{sub 0}{delta}V has been produced and collected in the storage circuit, a quantity P{sub 0}V{sub S}{<=}P{sub 0}{delta}V of gas is released to the analysis system. A set of electropneumatic valves, automatically operated by the control system, allows for gas delivery to the analysis equipment and retrieval of the set point pressure, by compression of the variable volume, with no influence on the reaction. All relevant parameters are monitored and logged on a personal computer. The control and data acquisition software, made out using National Instrument LABVIEW trade mark sign , also provides control of the analysis equipment. The ability of the proposed setup to not affect the ongoing process allows real time monitoring (by drawing samples at regular time intervals during the reaction) of the gas production. Moreover, since the amount of gas P{sub 0}V{sub S} drawn at each sampling is always the same, it is possible to establish at a glance whether or not there are changes in the concentration of any

  18. Emergency heat removal system for a nuclear reactor

    DOEpatents

    Dunckel, Thomas L.

    1976-01-01

    A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

  19. Autonomous Control and Diagnostics of Space Reactor Systems

    SciTech Connect

    Upadhyaya, B.R.; Xu, X.; Perillo, S.R.P.; Na, M.G.

    2006-07-01

    This paper describes three key features of the development of an autonomous control strategy for space reactor systems. These include the development of a reactor simulation model for transient analysis, development of model-predictive control as part of the autonomous control strategy, and a fault detection and isolation module. The latter is interfaced with the control supervisor as part of a hierarchical control system. The approach has been applied to the nodal model of the SP-100 reactor with a thermo-electric generator. The results of application demonstrate the effectiveness of the control approach and its ability to reconfigure the control mode under fault conditions. (authors)

  20. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    SciTech Connect

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  1. Microprocessor tester for the treat upgrade reactor trip system

    SciTech Connect

    Lenkszus, F.R.; Bucher, R.G.

    1984-01-01

    The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.

  2. ADVANCED GAS TURBINE SYSTEMS RESEARCH

    SciTech Connect

    Unknown

    2000-01-01

    The activities of the Advanced Gas Turbine Systems Research (AGRSR) program are described in the quarterly report. The report is divided into discussions of Membership, Administration, Technology Transfer (Workshop/Education) and Research. Items worthy of note are presented in extended bullet format following the appropriate heading.

  3. ADVANCED GAS TURBINE SYSTEMS RESEARCH

    SciTech Connect

    Unknown

    2002-04-01

    The activities of the Advanced Gas Turbine Systems Research (AGTSR) program for this reporting period are described in this quarterly report. The report is divided into discussions of Membership, Administration, Technology Transfer (Workshop/Education), Research and Miscellaneous Related Activity. Items worthy of note are presented in extended bullet format following the appropriate heading.

  4. ADVANCED GAS TURBINE SYSTEMS RESEARCH

    SciTech Connect

    Unknown

    2002-02-01

    The activities of the Advanced Gas Turbine Systems Research (AGTSR) program for this reporting period are described in this quarterly report. The report is divided into discussions of Membership, Administration, Technology Transfer (Workshop/Education), Research and Miscellaneous Related Activity. Items worthy of note are presented in extended bullet format following the appropriate heading.

  5. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    SciTech Connect

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

  6. Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications

    SciTech Connect

    Lee Nelson

    2011-09-01

    This report is a summary of analyses performed by the NGNP project to determine whether it is technically and economically feasible to integrate high temperature gas cooled reactor (HTGR) technology into industrial processes. To avoid an overly optimistic environmental and economic baseline for comparing nuclear integrated and conventional processes, a conservative approach was used for the assumptions and calculations.

  7. Advanced gas turbine systems program

    SciTech Connect

    Zeh, C.M.

    1995-06-01

    The U.S. Department of Energy (DOE) is sponsoring a program to develop fuel-efficient gas turbine-based power systems with low emissions. DOE`s Office of Fossil Energy (DOE/FE) and Office of Energy Efficiency and Renewable Energy (DOE/EE) have initiated an 8-year program to develop high-efficiency, natural gas-fired advanced gas turbine power systems. The Advanced Turbine Systems (ATS) Program will support full-scale prototype demonstration of both industrial- and utility-scale systems that will provide commercial marketplace entries by the year 2000. When the program targets are met, power system emissions will be lower than from the best technology in use today. Efficiency of the utility-scale units will be greater than 60 percent on a lower heating value basis, and emissions of carbon dioxide will be reduced inversely with this increase. Industrial systems will also see an improvement of at least 15 percent in efficiency. Nitrogen oxides will be reduced by at least 10 percent, and carbon monoxide and hydrocarbon emissions will each be kept below 20 parts per million, for both utility and industrial systems.

  8. Gas Hydrate Petroleum System Analysis

    NASA Astrophysics Data System (ADS)

    Collett, T. S.

    2012-12-01

    In a gas hydrate petroleum system, the individual factors that contribute to the formation of gas hydrate accumulations, such as (1) gas hydrate pressure-temperature stability conditions, (2) gas source, (3) gas migration, and (4) the growth of the gas hydrate in suitable host sediment can identified and quantified. The study of know and inferred gas hydrate accumulations reveal the occurrence of concentrated gas hydrate is mostly controlled by the presence of fractures and/or coarser grained sediments. Field studies have concluded that hydrate grows preferentially in coarse-grained sediments because lower capillary pressures in these sediments permit the migration of gas and nucleation of hydrate. Due to the relatively distal nature of the deep marine geologic settings, the overall abundance of sand within the shallow geologic section is usually low. However, drilling projects in the offshore of Japan, Korea, and in the Gulf of Mexico has revealed the occurrence of significant hydrate-bearing sand reservoirs. The 1999/2000 Japan Nankai Trough drilling confirmed occurrence of hydrate-bearing sand-rich intervals (interpreted as turbidite fan deposits). Gas hydrate was determined to fill the pore spaces in these deposits, reaching saturations up to 80% in some layers. A multi-well drilling program titled "METI Toaki-oki to Kumano-nada" also identified sand-rich reservoirs with pore-filling hydrate. The recovered hydrate-bearing sand layers were described as very-fine- to fine-grained turbidite sand layers measuring from several centimeters up to a meter thick. However, the gross thickness of the hydrate-bearing sand layers were up to 50 m. In 2010, the Republic of Korea conducted the Second Ulleung Basin Gas Hydrate (UBGH2) Drilling Expedition. Seismic data clearly showed the development of a thick, potential basin wide, sedimentary sections characterized by mostly debris flows. The downhole LWD logs and core data from Site UBGH2-5 reveal that each debris flows is

  9. Multi-Megawatt Gas Turbine Power Systems for Lunar Colonies

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.

    2006-01-01

    A concept for development of second generation 10 MWe prototype lunar power plant utilizing a gas cooled fission reactor supplying heated helium working fluid to two parallel 5 MWe closed cycle gas turbines is presented. Such a power system is expected to supply the energy needs for an initial lunar colony with a crew of up to 50 persons engaged in mining and manufacturing activities. System performance and mass details were generated by an author developed code (BRMAPS). The proposed pilot power plant can be a model for future plants of the same capacity that could be tied to an evolutionary lunar power grid.

  10. Atmospheric pressure flow reactor: Gas phase chemical kinetics under tropospheric conditions without wall effects

    NASA Technical Reports Server (NTRS)

    Koontz, Steven L. (Inventor); Davis, Dennis D. (Inventor)

    1991-01-01

    A flow reactor for simulating the interaction in the troposphere is set forth. A first reactant mixed with a carrier gas is delivered from a pump and flows through a duct having louvers therein. The louvers straighten out the flow, reduce turbulence and provide laminar flow discharge from the duct. A second reactant delivered from a source through a pump is input into the flowing stream, the second reactant being diffused through a plurality of small diffusion tubes to avoid disturbing the laminar flow. The commingled first and second reactants in the carrier gas are then directed along an elongated duct where the walls are spaced away from the flow of reactants to avoid wall interference, disturbance or turbulence arising from the walls. A probe connected with a measuring device can be inserted through various sampling ports in the second duct to complete measurements of the first and second reactants and the product of their reaction at selected XYZ locations relative to the flowing system.

  11. Development of ceramic membrane reactors for high temperature gas cleanup. Final report

    SciTech Connect

    Roberts, D.L.; Abraham, I.C.; Blum, Y.; Gottschlich, D.E.; Hirschon, A.; Way, J.D.; Collins, J.

    1993-06-01

    The objective of this project was to develop high temperature, high pressure catalytic ceramic membrane reactors and to demonstrate the feasibility of using these membrane reactors to control gaseous contaminants (hydrogen sulfide and ammonia) in integrated gasification combined cycle (IGCC) systems. Our strategy was to first develop catalysts and membranes suitable for the IGCC application and then combine these two components as a complete membrane reactor system. We also developed a computer model of the membrane reactor and used it, along with experimental data, to perform an economic analysis of the IGCC application. Our results have demonstrated the concept of using a membrane reactor to remove trace contaminants from an IGCC process. Experiments showed that NH{sub 3} decomposition efficiencies of 95% can be achieved. Our economic evaluation predicts ammonia decomposition costs of less than 1% of the total cost of electricity; improved membranes would give even higher conversions and lower costs.

  12. Tritium Formation and Mitigation in High-Temperature Reactor Systems

    SciTech Connect

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  13. Treatment of Dye Wastewater by Using a Hybrid Gas/Liquid Pulsed Discharge Plasma Reactor

    NASA Astrophysics Data System (ADS)

    Lu, Na; Li, Jie; Wu, Yan; Masayuki, Sato

    2012-02-01

    A hybrid gas/liquid pulsed discharge plasma reactor using a porous ceramic tube is proposed for dye wastewater treatment. High voltage pulsed discharge plasma was generated in the gas phase and simultaneously the plasma channel was permeated through the tiny holes of the ceramic tube into the water phase accompanied by gas bubbles. The porous ceramic tube not only separated the gas phase and liquid phase but also offered an effective plasma spreading channel. The effects of the peak pulse voltage, additive gas varieties, gas bubbling rate, solution conductivity and TiO2 addition were investigated. The results showed that this reactor was effective for dye wastewater treatment. The decoloration efficiency of Acid Orange II was enhanced with an increase in the power supplied. Under the studied conditions, 97% of Acid Orange II in aqueous solution was effectively decolored with additive oxygen gas, which was 51% higher than that with argon gas, and the increasing O2 bubbling rate also benefited the decoloration of dye wastewater. Water conductivity had a small effect on the level of decoloration. Catalysis of TiO2 could be induced by the pulsed discharge plasma and addition of TiO2 aided the decoloration of Acid Orange II.

  14. Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

    SciTech Connect

    Conklin, J.C.

    1986-08-01

    The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory.

  15. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, October 1, 1979-December 31, 1979

    SciTech Connect

    Not Available

    1980-04-18

    This report presents the results of work performed from October 1, 1979 through December 31, 1979. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described. This includes: screening creep results, weight gain and post-exposure mechanical properties for materials thermally exposed at 750/sup 0/ and 850/sup 0/C (1382/sup 0/ and 1562/sup 0/F). In addition, the status of the data management system is described.

  16. Autonomous Control of Space Reactor Systems

    SciTech Connect

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

    2007-11-30

    Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

  17. Reactor Lithium Heat Pipes for HP-STMCs Space Reactor Power System

    NASA Astrophysics Data System (ADS)

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2004-02-01

    Design and performance analysis of the nuclear reactor's lithium heat pipes for a 110-kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The evaporator length of the heat pipes is the same as the active core height (0.45 m) and the C-C finned condenser is of the same length as the STMC panels (1.5 m). The C-C finned condenser section is radiatively coupled to the collector shoes of the STMCs placed on both sides. The lengths of the adiabatic section, the values of the power throughput and the evaporator wall temperature depend on the radial location of the heat pipe in the reactor core and the number and dimensions of the potassium heat pipes in the heat rejection radiator. The reactor heat pipes have a total length that varies from 7.57 to 7.73 m, and a 0.2 mm thick Mo-14%Re wick with an average pore radius of 12 μm. The wick is separated from the Mo-14%Re wall by a 0.5 mm annulus filled with liquid lithium, to raise the prevailing capillary limit. The nominal evaporator (or reactor) temperature varies from 1513 to 1591 K and the thermal power of the reactor is 1.6 MW, which averages 12.7 kW for each of the 126 reactor heat pipes. The power throughput per heat pipe increase to a nominal 15.24 kW at the location of the peak power in the core and to 20.31 kW when an adjacent heat pipe fails. The prevailing capillary limit of the reactor heat pipes is 28.3 kW, providing a design margin >= 28%.

  18. Reactor Lithium Heat Pipes for HP-STMCs Space Reactor Power System

    SciTech Connect

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2004-02-04

    Design and performance analysis of the nuclear reactor's lithium heat pipes for a 110-kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The evaporator length of the heat pipes is the same as the active core height (0.45 m) and the C-C finned condenser is of the same length as the STMC panels (1.5 m). The C-C finned condenser section is radiatively coupled to the collector shoes of the STMCs placed on both sides. The lengths of the adiabatic section, the values of the power throughput and the evaporator wall temperature depend on the radial location of the heat pipe in the reactor core and the number and dimensions of the potassium heat pipes in the heat rejection radiator. The reactor heat pipes have a total length that varies from 7.57 to 7.73 m, and a 0.2 mm thick Mo-14%Re wick with an average pore radius of 12 {mu}m. The wick is separated from the Mo-14%Re wall by a 0.5 mm annulus filled with liquid lithium, to raise the prevailing capillary limit. The nominal evaporator (or reactor) temperature varies from 1513 to 1591 K and the thermal power of the reactor is 1.6 MW, which averages 12.7 kW for each of the 126 reactor heat pipes. The power throughput per heat pipe increase to a nominal 15.24 kW at the location of the peak power in the core and to 20.31 kW when an adjacent heat pipe fails. The prevailing capillary limit of the reactor heat pipes is 28.3 kW, providing a design margin {>=} 28%.

  19. Reactors for Catalytic Methanation in the Conversion of Biomass to Synthetic Natural Gas (SNG).

    PubMed

    Schildhauer, Tilman J; Biollaz, Serge M A

    2015-01-01

    Production of Synthetic Natural Gas (SNG) from biomass is an important step to decouple the use of bioenergy from the biomass production with respect to both time and place. While anaerobic digestion of wet biomass is a state-of-the art process, wood gasification to producer gas followed by gas cleaning and methanation has only just entered the demonstration scale. Power-to-Gas applications using biogas from biomass fermentation or producer gas from wood gasification as carbon oxide source are under development. Due to the importance of the (catalytic) methanation step in the production of SNG from dry biomass or within Power-to-Gas applications, the specific challenges of this step and the developed reactor types are discussed in this review. PMID:26598404

  20. Reactors for Catalytic Methanation in the Conversion of Biomass to Synthetic Natural Gas (SNG).

    PubMed

    Schildhauer, Tilman J; Biollaz, Serge M A

    2015-01-01

    Production of Synthetic Natural Gas (SNG) from biomass is an important step to decouple the use of bioenergy from the biomass production with respect to both time and place. While anaerobic digestion of wet biomass is a state-of-the art process, wood gasification to producer gas followed by gas cleaning and methanation has only just entered the demonstration scale. Power-to-Gas applications using biogas from biomass fermentation or producer gas from wood gasification as carbon oxide source are under development. Due to the importance of the (catalytic) methanation step in the production of SNG from dry biomass or within Power-to-Gas applications, the specific challenges of this step and the developed reactor types are discussed in this review.

  1. Modeling Reactor Coolant Systems Thermal-Hydraulic Transients

    1999-10-05

    RELAP5/MOD3.2* is used to model reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transients without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal-hydraulic systems. Control system and secondary system components are included to allow modeling of themore » plant controls, turbines, condensers, and secondary feedwater systems.« less

  2. A high energy neutral beam system for reactors

    SciTech Connect

    Anderson, O.A.; Chan, C.F.; Cooper, W.S.; Leung, K.N.; Lietzke, A.F.; Kim, C.H.; Kunkel, W.B.; Kwan, J.W.; Purgalis, P.; Schlachter, A.S.

    1988-09-01

    High energy neutral beams provide a promising method of heating and driving current in steady-state tokamak fusion reactors. As an example, we have made a conceptual design of a neutral beam system for current drive on the International Thermonuclear Experimental Reactor (ITER). The system, based on electrostatic acceleration of D/sup /minus// ions, can deliver up to 100 MW of 1.6 MeV D/sup 0/ neutrals through three ports. Radiation protection is provided by locating sensitive beamline components 35 to 50 m from the reactor. In an application to a 3300 MW power reactor, a system delivering 120 MW of 2-2.4 MeV deuterium beams assisted by 21 MW of lower hybrid wave power drives 25 MA and provides an adequate plasma power gain (Q = 24) for a commercial fusion power plant. 8 refs., 1 fig., 2 tabs.

  3. Code System for Reactor Physics and Fuel Cycle Simulation.

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterativemore » processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.« less

  4. Code System for Reactor Physics and Fuel Cycle Simulation.

    SciTech Connect

    TEUCHERT, E.

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.

  5. Summary of HTGR (high-temperature gas-cooled reactor) benchmark data from the high temperature lattice test reactor

    SciTech Connect

    Newman, D.F.

    1989-10-01

    The High Temperature Lattice Test Reactor (HTLTR) was a unique critical facility specifically built and operated to measure variations in neutronic characteristics of high temperature gas cooled reactor (HTGR) lattices at temperatures up to 1000{degree}C. The Los Alamos National Laboratory commissioned Pacific Northwest Laboratory (PNL) to prepare this summary reference report on the HTLTR benchmark data and its associated documentation. In the initial stages of the program, the principle of the measurement of k{sub {infinity}} using the unpoisoned technique (developed by R.E. Heineman of PNL) was subjected to extensive peer review within PNL and the General Atomic Company. A number of experiments were conducted at PNL in the Physical Constants Testing Reactor (PCTR) using both the unpoisoned technique and the well-established null reactivity technique that substantiated the equivalence of the measurements by direct comparison. Records of all data from fuel fabrication, the reactor experiments, and the analytical results were compiled and maintained to meet applicable quality assurance standards in place at PNL. Sensitivity of comparisons between measured and calculated k{sub {infinity}}(T) data for various HTGR lattices to changes in neutron cross section data, graphite scattering kernel models, and fuel block loading variations, were analyzed by PNL for the Electric Power Research Institute. As a part of this effort, the fuel rod composition in the dilute {sup 233}UO{sub 2}-ThO{sub 2} HTGR central cell (HTLTR Lattice {number sign}3) was sampled and analyzed by mass spectrometry. Values of k{sub {infinity}} calculated for that lattice were about 5% higher than those measured. Trace quantities of sodium chloride were found in the fuel rod that were equivalent to 22 atom parts-per-million of natural boron.

  6. A review of existing gas-cooled reactor circulators with application of the lessons learned to the new production reactor circulators

    SciTech Connect

    White, L.S.

    1990-07-01

    This report presents the results of a study of the lessons learned during the design, testing, and operation of gas-cooled reactor coolant circulators. The intent of this study is to identify failure modes and problem areas of the existing circulators so this information can be incorporated into the design of the circulators for the New Production Reactor (NPR)-Modular High-Temperature Gas Cooled Reactor (MHTGR). The information for this study was obtained primarily from open literature and includes data on high-pressure, high-temperature helium test loop circulators as well as the existing gas cooled reactors worldwide. This investigation indicates that trouble free circulator performance can only be expected when the design program includes a comprehensive prototypical test program, with the results of this test program factored into the final circulator design. 43 refs., 7 tabs.

  7. Purge gas recovery of ammonia synthesis plant by integrated configuration of catalytic hydrogen-permselective membrane reactor and solid oxide fuel cell as a novel technology

    NASA Astrophysics Data System (ADS)

    Siavashi, Fakhteh; Saidi, Majid; Rahimpour, Mohammad Reza

    2014-12-01

    The purge gas emission of ammonia synthesis plant which contains hazardous components is one of the major sources of environmental pollution. Using integrated configuration of catalytic hydrogen-permselective membrane reactor and solid oxide fuel cell (SOFC) system is a new approach which has a great impact to reduce the pollutant emission. By application of this method, not only emission of ammonia and methane in the atmosphere is prevented, hydrogen is produced through the methane steam reforming and ammonia decomposition reactions that take place simultaneously in a catalytic membrane reactor. The pure generated hydrogen by recovery of the purge gas in the Pd-Ag membrane reactor is used as a feed of SOFC. Since water is the only byproduct of the electrochemical reaction in the SOFC, it is recycled to the reactor for providing the required water of the reforming reaction. Performance investigation of the reactor represents that the rate of hydrogen permeation increases with enhancing the reactor temperature and pressure. Also modeling results indicate that the SOFC performance improves with increasing the temperature and fuel utilization ratio. The generated power by recovery of the purging gas stream of ammonia synthesis plant in the Razi petrochemical complex is about 8 MW.

  8. Measuring carbon monoxide gas-liquid mass transfer in a stirred tank reactor for syngas fermentation.

    PubMed

    Riggs, Seth S; Heindel, Theodore J

    2006-01-01

    Carbon monoxide-liquid mass transfer rates in a water-filled stirred tank reactor are determined using a myoglobin protein method to measure dissolved carbon monoxide concentrations as a function of time. Data are acquired over a range of stirrer speeds (200 < or = N < or = 600 rpm) and gas flow rates (1 < or = Q < or = 6 L/min), corresponding to a gas retention time range of 1.2-7 min. Volumetric CO-water mass transfer coefficients range from 0.003 to 0.043 s(-1) and are well-correlated using the power density and the superficial gas velocity.

  9. High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

    SciTech Connect

    Larry Demick

    2010-08-01

    The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

  10. Reference Reactor Module for the Affordable Fission Surface Power System

    NASA Astrophysics Data System (ADS)

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

    2008-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO2-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important ``affordability'' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  11. Reference Reactor Module for the Affordable Fission Surface Power System

    SciTech Connect

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

    2008-01-21

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO{sub 2}-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important 'affordability' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  12. Passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  13. Natural circulating passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  14. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program: Topical report I, selection of candidate alloys. Volume 3. Selection of surface coating/substrate systems for screening creep and structural stability studies

    SciTech Connect

    Not Available

    1980-06-20

    Considering the high temperature, low O/sub 2/, high C environment of operation in the Very High Temperature Reactor (VHTR) Systems, the utilization of coatings is envisaged to hold potential for extending component lifetimes through the formation of stable and continuous oxide films with enhanced resistance to C diffusion. A survey of the current state of technology for high temperature coatings has been performed. The usefulness of these coatings on the Mo, Ni, and Fe base alloys is discussed. Specifically, no coating substitute was identified for TZM other than the well known W-3 (pack silicide) and Al/sub 2/O/sub 3/ forming coatings were recommended for the Fe and Ni base structural materials. Recommendations as to coating types and processng have been made based on the predicted VHTR component size, shape, base metal and operational environment. Four tests designed to evaluate the effects of selected combinations of coatings and substrate matrices are recommended for consideration.

  15. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

    NASA Astrophysics Data System (ADS)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

    1992-10-01

    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  16. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

    1992-01-01

    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  17. Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK™

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Sanchez, Travis

    2005-02-01

    The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK™ (Simulink, 2004). SIMULINK™ is a development environment packaged with MatLab™ (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion components such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK™ models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK™ modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator).

  18. Upgrading producer gas quality from rubber wood gasification in a radio frequency tar thermocatalytic treatment reactor.

    PubMed

    Anis, Samsudin; Zainal, Z A

    2013-12-01

    This study focused on improving the producer gas quality using radio frequency (RF) tar thermocatalytic treatment reactor. The producer gas containing tar, particles and water was directly passed at a particular flow rate into the RF reactor at various temperatures for catalytic and thermal treatments. Thermal treatment generates higher heating value of 5.76 MJ Nm(-3) at 1200°C. Catalytic treatments using both dolomite and Y-zeolite provide high tar and particles conversion efficiencies of about 97% on average. The result also showed that light poly-aromatic hydrocarbons especially naphthalene and aromatic compounds particularly benzene and toluene were still found even at higher reaction temperatures. Low energy intensive RF tar thermocatalytic treatment was found to be effective for upgrading the producer gas quality to meet the end user requirements and increasing its energy content.

  19. Upgrading producer gas quality from rubber wood gasification in a radio frequency tar thermocatalytic treatment reactor.

    PubMed

    Anis, Samsudin; Zainal, Z A

    2013-12-01

    This study focused on improving the producer gas quality using radio frequency (RF) tar thermocatalytic treatment reactor. The producer gas containing tar, particles and water was directly passed at a particular flow rate into the RF reactor at various temperatures for catalytic and thermal treatments. Thermal treatment generates higher heating value of 5.76 MJ Nm(-3) at 1200°C. Catalytic treatments using both dolomite and Y-zeolite provide high tar and particles conversion efficiencies of about 97% on average. The result also showed that light poly-aromatic hydrocarbons especially naphthalene and aromatic compounds particularly benzene and toluene were still found even at higher reaction temperatures. Low energy intensive RF tar thermocatalytic treatment was found to be effective for upgrading the producer gas quality to meet the end user requirements and increasing its energy content. PMID:24185417

  20. STUDY ON AIR INGRESS MITIGATION METHODS IN THE VERY HIGH TEMPERATURE GAS COOLED REACTOR (VHTR)

    SciTech Connect

    Chang H. Oh

    2011-03-01

    An air-ingress accident followed by a pipe break is considered as a critical event for a very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break leading to oxidation of the in-core graphite structure. Thus, without mitigation features, this accident might lead to severe exothermic chemical reactions of graphite and oxygen. Under extreme circumstances, a loss of core structural integrity may occur along with excessive release of radiological inventory. Idaho National Laboratory under the auspices of the U.S. Department of Energy is performing research and development (R&D) that focuses on key phenomena important during challenging scenarios that may occur in the VHTR. Phenomena Identification and Ranking Table (PIRT) studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Oh et al. 2006, Schultz et al. 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) requirements are part of the experimental validation plan. This paper discusses about various air-ingress mitigation concepts applicable for the VHTRs. The study begins with identifying important factors (or phenomena) associated with the air-ingress accident by using a root-cause analysis. By preventing main causes of the important events identified in the root-cause diagram, the basic air-ingress mitigation ideas can be conceptually derived. The main concepts include (1) preventing structural degradation of graphite supporters; (2) preventing local stress concentration in the supporter; (3) preventing graphite oxidation; (4) preventing air ingress; (5) preventing density gradient driven flow; (4) preventing fluid density gradient; (5) preventing fluid temperature gradient; (6) preventing high temperature. Based on the basic concepts listed above, various air

  1. System and method for the analysis of one or more compounds and/or species produced by a solution-based nuclear reactor

    DOEpatents

    Policke, Timothy A; Nygaard, Eric T

    2014-05-06

    The present invention relates generally to both a system and method for determining the composition of an off-gas from a solution nuclear reactor (e.g., an Aqueous Homogeneous Reactor (AHR)) and the composition of the fissioning solution from those measurements. In one embodiment, the present invention utilizes at least one quadrupole mass spectrometer (QMS) in a system and/or method designed to determine at least one or more of: (i) the rate of production of at least one gas and/or gas species from a nuclear reactor; (ii) the effect on pH by one or more nitrogen species; (iii) the rate of production of one or more fission gases; and/or (iv) the effect on pH of at least one gas and/or gas species other than one or more nitrogen species from a nuclear reactor.

  2. Oxidation of ammonium sulfite by a multi-needle-to-plate gas phase pulsed corona discharge reactor

    NASA Astrophysics Data System (ADS)

    Ren, Hua; Lu, Na; Shang, Kefeng; Li, Jie; Wu, Yan

    2013-03-01

    The oxidation of ammonium sulfite in the ammonia-based flue gas desulfurization (FGD) process was investigated in a multi-needle-to-plate gas phase pulsed corona discharge reactor in this paper. The effect of several parameters, including capacitance and peak pulse voltage of discharge system, electrode gap and bubbling gas flow rate on the oxidation rate of ammonium sulfite was reviewed. The oxidation rate of ammonium sulfite could reach 47.2% at the capacitance, the peak pulse voltage, electrode gap and bubbling gas flow rate equal to 2 nF, -24.6 k V, 35 mm and 4 L min-1 within treatment time of 40 min The experimental results indicate that the gas phase pulsed discharge system with a multi-needle-to-plate electrode can oxide the ammonium sulfite. The oxidation rate increased with the applied capacitance and peak pulse voltage and decreased with the electrode gap. As the bubbling gas flow rate increased, the oxidation rate increased first and then tended to reach a stationary value. These results would be important for the process optimization of the (NH4)2SO3 to (NH4)2SO4 oxidation.

  3. CONCEPTUAL DESIGN OF A LUNAR REGOLITH CLUSTERED-REACTOR SYSTEM

    SciTech Connect

    John Darrell Bess

    2009-06-01

    It is proposed that a fast-fission, heatpipe-cooled, lunar-surface power reactor system be divided into subcritical units that could be launched safely without the incorporation of additional spectral shift absorbers or other complex means of control. The reactor subunits are to be emplaced directly into the lunar regolith utilizing the regolith not just for shielding but as the reflector material to increase the neutron economy of the system. While a single subunit cannot achieve criticality by itself, coordinated placement of additional subunits will provide a critical reactor system for lunar surface power generation. A lunar regolith clustered-reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of a slight increase in launch mass per rated power level and an overall reduction in neutron economy when compared to a single-reactor system. Additional subunits may be launched with future missions to increase the cluster size and power according to desired lunar base power demand and lifetime. The results address the potential uncertainties associated with the lunar regolith material and emplacement of the subunit systems. Physical distance between subunits within the clustered emplacement exhibits the most significant feedback regarding changes in overall system reactivity. Narrow, deep holes will be the most effective in reducing axial neutron leakage from the core. The variation in iron concentration in the lunar regolith can directly influence the overall system reactivity although its effects are less than the more dominant factors of subunit emplacement.

  4. New reactor technology: safety improvements in nuclear power systems.

    PubMed

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

  5. New reactor technology: safety improvements in nuclear power systems.

    PubMed

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems. PMID:18049233

  6. Deployment history and design considerations for space reactor power systems

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.

    2009-05-01

    The history of the deployment of nuclear reactors in Earth orbits is reviewed with emphases on lessons learned and the operation and safety experiences. The former Soviet Union's "BUK" power systems, with SiGe thermoelectric conversion and fast neutron energy spectrum reactors, powered a total of 31 Radar Ocean Reconnaissance Satellites (RORSATs) from 1970 to 1988 in 260 km orbit. Two of the former Soviet Union's TOPAZ reactors, with in-core thermionic conversion and epithermal neutron energy spectrum, powered two Cosmos missions launched in 1987 in ˜800 km orbit. The US' SNAP-10A system, with SiGe energy conversion and a thermal neutron energy spectrum reactor, was launched in 1965 in 1300 km orbit. The three reactor systems used liquid NaK-78 coolant, stainless steel structure and highly enriched uranium fuel (90-96 wt%) and operated at a reactor exit temperature of 833-973 K. The BUK reactors used U-Mo fuel rods, TOPAZ used UO 2 fuel rods and four ZrH moderator disks, and the SNAP-10A used moderated U-ZrH fuel rods. These low power space reactor systems were designed for short missions (˜0.5 kW e and ˜1 year for SNAP-10A, <3.0 kW e and <6 months for BUK, and ˜5.5 kW e and up to 1 year for TOPAZ). The deactivated BUK reactors at the end of mission, which varied in duration from a few hours to ˜4.5 months, were boosted into ˜800 km storage orbit with a decay life of more than 600 year. The ejection of the last 16 BUK reactor fuel cores caused significant contamination of Earth orbits with NaK droplets that varied in sizes from a few microns to 5 cm. Power systems to enhance or enable future interplanetary exploration, in-situ resources utilization on Mars and the Moon, and civilian missions in 1000-3000 km orbits would generate significantly more power of 10's to 100's kW e for 5-10 years, or even longer. A number of design options to enhance the operation reliability and safety of these high power space reactor power systems are presented and discussed.

  7. Dual annular rotating "windowed" nuclear reflector reactor control system

    DOEpatents

    Jacox, Michael G.; Drexler, Robert L.; Hunt, Robert N. M.; Lake, James A.

    1994-01-01

    A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

  8. OPERATION OF FUSION REACTORS IN ONE ATMOSPHERE OF AIR INSTEAD OF VACUUM SYSTEMS

    SciTech Connect

    Roth, J. Reece

    2009-07-26

    Engineering design studies of both magnetic and inertial fusion power plants have assumed that the plasma will undergo fusion reactions in a vacuum environment. Operation under vacuum requires an expensive additional major system for the reactor-a vacuum vessel with vacuum pumping, and raises the possibility of sudden unplanned outages if the vacuum containment is breached. It would be desirable in many respects if fusion reactors could be made to operate at one atmosphere with air surrounding the plasma, thus eliminating the requirement of a pressure vessel and vacuum pumping. This would have obvious economic, reliability, and engineering advantages for currently envisaged power plant reactors; it would make possible forms of reactor control not possible under vacuum conditions (i.e. adiabatic compression of the fusion plasma by increasing the pressure of surrounding gas); it would allow reactors used as aircraft engines to operate as turbojets or ramjets in the atmosphere, and it would allow reactors used as fusion rockets to take off from the surface of the earth instead of low earth orbit.

  9. Modular high temperature gas-cooled reactor plant design duty cycle. Revision 3

    SciTech Connect

    Chan, T.

    1989-12-31

    This document defines the Plant Design Duty Cycle (PCDC) for the Modular High Temperature Gas-cooled Reactor (MHTGR). The duty cycle is a set of events and their design number of occurrences over the life of the plant for which the MHTGR plant shall be designed to ensure that the plant meets all the top-level requirements. The duty cycle is representative of the types of events to be expected in multiple reactor module-turbine plant configurations of the MHTGR. A synopsis of each PDDC event is presented to provide an overview of the plant response and consequence. 8 refs., 1 fig., 4 tabs.

  10. Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

    SciTech Connect

    Orvis, D.D.; Raabe, P.H.

    1980-01-01

    This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs.

  11. 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.

  12. Small space reactor power systems for unmanned solar system exploration missions

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

  13. Small space reactor power systems for unmanned solar system exploration missions

    SciTech Connect

    Bloomfield, H.S.

    1987-12-01

    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

  14. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  15. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

    SciTech Connect

    Perret, G.; Pattupara, R. M.; Girardin, G.; Chawla, R.

    2012-07-01

    The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

  16. Continuous crafting of uniform colloidal nanocrystals using an inert-gas-driven microflow reactor

    NASA Astrophysics Data System (ADS)

    Tang, Hailong; He, Yanjie; Li, Bo; Jung, Jaehan; Zhang, Chuchu; Liu, Xiaobo; Lin, Zhiqun

    2015-05-01

    Recent research has witnessed rapid advances in synthesis of nanocrystals, which has led to the development of a large variety of approaches for producing nanocrystals with controlled dimensions. However, most of these techniques lack the high-throughput production. Herein, we report on a viable and robust strategy based on an inert-gas-driven microflow reactor for continuous crafting of high-quality colloidal nanocrystals. With the judicious introduction of the inert-gas driven capability, the microflow reactor provides an attractive platform for continuous production of colloidal nanocrystals in large quantities, including easily-oxidized nanocrystals. The as-synthesized nanocrystals possessed a uniform size and shape. Intriguingly, the size of nanocrystals can be effectively tailored by varying the flow rate and the precursor concentration. We envision that the microflow reactor strategy is general and offers easy access to a wide range of scalable nanocrystals for potential applications in sensors, optics, optoelectronics, solar energy conversion, batteries, photocatalysis, and electronic devices.Recent research has witnessed rapid advances in synthesis of nanocrystals, which has led to the development of a large variety of approaches for producing nanocrystals with controlled dimensions. However, most of these techniques lack the high-throughput production. Herein, we report on a viable and robust strategy based on an inert-gas-driven microflow reactor for continuous crafting of high-quality colloidal nanocrystals. With the judicious introduction of the inert-gas driven capability, the microflow reactor provides an attractive platform for continuous production of colloidal nanocrystals in large quantities, including easily-oxidized nanocrystals. The as-synthesized nanocrystals possessed a uniform size and shape. Intriguingly, the size of nanocrystals can be effectively tailored by varying the flow rate and the precursor concentration. We envision that the

  17. Reference reactor module for NASA's lunar surface fission power system

    SciTech Connect

    Poston, David I; Kapernick, Richard J; Dixon, David D; Werner, James; Qualls, Louis; Radel, Ross

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  18. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOEpatents

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  19. [Phylogeny of gas exchange systems].

    PubMed

    Jürgens, K D; Gros, G

    2002-04-01

    Several systems of gas transport have developed during evolution, all of which are able to sufficiently supply oxygen to the tissues and eliminate the CO2 produced by the metabolism, in spite of great distances between the environment and the individual cells of the tissues. Almost all these systems utilize a combination of convection and diffusion steps. Convection achieves an efficient transport of gas over large distances, but requires energy and cannot occur across tissue barriers. Diffusion, on the other hand, achieves gas transport across barriers, but requires optimization of diffusion paths and diffusion areas. When two convectional gas flows are linked via a diffusional barrier (gas/fluid in the case of the avian lung, fluid/fluid in the case of gills), the directions in which the respective convectional movements pass each other are important determinants of gas exchange efficiency (concurrent, countercurrent and cross-current systems). The tracheal respiration found in insects has the advantage of circumventing the convective gas transport step in the blood, thereby avoiding the high energy expenditure of circulatory systems. This is made possible by a system of tracheae, ending in tracheoles, that reaches from the body surface to every cell within the body. The last step of gas transfer in these animals occurs by diffusion from the tracheoles ("air capillaries") to the mitochondria of cells. The disadvantage is that the tracheal system occupies a substantial fraction of body volume and that, due to limited mechanical stability of tracheal walls, this system would not be able to operate under conditions of high hydrostatic pressures, i. e. in large animals. Respiration in an "open" system, i. e. direct exposure of the diffusional barrier to the environmental air, eliminates the problem of bringing the oxygen to the barrier by convection, as is necessary in the avian and mammalian lung, in the insects' tracheal system and in the gills. An open system is

  20. EBR-II Cover Gas Cleanup System upgrade distributed control and front end computer systems

    SciTech Connect

    Carlson, R.B.

    1992-05-01

    The Experimental Breeder Reactor II (EBR-II) Cover Gas Cleanup System (CGCS) control system was upgraded in 1991 to improve control and provide a graphical operator interface. The upgrade consisted of a main control computer, a distributed control computer, a front end input/output computer, a main graphics interface terminal, and a remote graphics interface terminal. This paper briefly describes the Cover Gas Cleanup System and the overall control system; gives reasons behind the computer system structure; and then gives a detailed description of the distributed control computer, the front end computer, and how these computers interact with the main control computer. The descriptions cover both hardware and software.

  1. EBR-II Cover Gas Cleanup System upgrade distributed control and front end computer systems

    SciTech Connect

    Carlson, R.B.

    1992-01-01

    The Experimental Breeder Reactor II (EBR-II) Cover Gas Cleanup System (CGCS) control system was upgraded in 1991 to improve control and provide a graphical operator interface. The upgrade consisted of a main control computer, a distributed control computer, a front end input/output computer, a main graphics interface terminal, and a remote graphics interface terminal. This paper briefly describes the Cover Gas Cleanup System and the overall control system; gives reasons behind the computer system structure; and then gives a detailed description of the distributed control computer, the front end computer, and how these computers interact with the main control computer. The descriptions cover both hardware and software.

  2. A microcomputer-controlled gas phase microreactor system

    SciTech Connect

    Morris, R.M.

    1983-08-01

    Although automated reactors are effective tools for studying a single type of reaction or optimizing catalyst performance, they may not be well suited for exploratory research. These reactors generally have several shortcomings. First, they may have limited versatility since they are usually designed with a single application in mind. Second, computer systems used for process control and data acquisition are often expensive and complex, so that once they are set up for a given application, it is quite difficult to adapt them for another. Because of these restrictions, experimental reactors are often operated manually, requiring a full-time operator to monitor operations and acquire data. This is a greater problem in laboratories where projects are often short-term, and the costs of setting up an automated reactor may outweigh the benefits of automation. For an automated reactor to be cost-effective in such an environment, both reactor hardware and control software must be versatile enough that they can be easily modified and adapted for different experiments. An automated gas-flow microreactor has been designed and constructed which is both inexpensive and flexible. The reactor is capable of performing three different types of experiments, 1) continuous reagent feed with analysis of the product stream, 2) pulsed-flow experiments, and 3) temperature-programmed desorption (TPD) and reaction (TPR). Conversion of the reactor from one configuration to another requires less than one hour. Process control and data acquisition are performed using an Apple II Plus microcomputer (Apple Computer Corp., Cupertino, Calif.) and an ISAAC interface device (Cyborg Corp., Newton, Mass.).

  3. Comprehensive Thermal Hydraulics Research of the Very High Temperature Gas Cooled Reactor

    SciTech Connect

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; David Petti; Hyung Kang

    2010-10-01

    The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  4. Solid-Core, Gas-Cooled Reactor for Space and Surface Power

    NASA Astrophysics Data System (ADS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S∧4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S∧4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively.

  5. Solid-Core, Gas-Cooled Reactor for Space and Surface Power

    SciTech Connect

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-20

    The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respect0011ive.

  6. Gas expander based power plant system

    SciTech Connect

    Kucerija, Z.

    1991-04-02

    This patent describes a method for generating electricity where gas pressure from a high pressure gas transmission line is reduced with a gas expander system capable of operating with fluctuating high pressure gas conditions and maintaining required outlet gas conditions for delivery to a lower pressure gas distribution line. It comprises supplying a portion of the lower pressure gas to the combustion apparatus of a steam generator, heating feedwater in the steam generator and producing a low pressure steam, condensing the steam in a heat exchange using the heat of condensation of the steam to heat a portion of high pressure gas in a heat exchange, mixing the heated high pressure gas with unheated high pressure gas in proportion to controlled variable requirements, reducing the pressure of the combined high pressure gas by passing the gas through a gas expander which produces useful shaft power for driving an electrical generator or for mechanical drive applications.

  7. Gas sampling system for a mass spectrometer

    DOEpatents

    Taylor, Charles E; Ladner, Edward P

    2003-12-30

    The present invention relates generally to a gas sampling system, and specifically to a gas sampling system for transporting a hazardous process gas to a remotely located mass spectrometer. The gas sampling system includes a capillary tube having a predetermined capillary length and capillary diameter in communication with the supply of process gas and the mass spectrometer, a flexible tube surrounding and coaxial with the capillary tube intermediate the supply of process gas and the mass spectrometer, a heat transfer tube surrounding and coaxial with the capillary tube, and a heating device in communication the heat transfer tube for substantially preventing condensation of the process gas within the capillary tube.

  8. Design of virtual SCADA simulation system for pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  9. System aspects of a Space Nuclear Reactor Power System

    SciTech Connect

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Kia, T.; Nesmith, B.

    1988-01-01

    Selected systems aspects of a 300 kW nuclear reactor power system for spacecraft have been studied. The approach included examination of two candidate missions and their associated spacecraft, and a number of special topics dealing with the power system design and operation. The missions considered were a reusable orbital transfer vehicle and a space-based radar. The special topics included: power system configuration and scaling, launch vehicle integration, operating altitude, orbital storage, start-up, thawing, control, load following, procedures in case of malfunction, restart, thermal and nuclear radiation to other portions of the spacecraft, thermal stresses between subsystems, boom and cable designs, vibration modes, altitude control, reliability, and survivability. Among the findings are that the stowed length of the power system is important to mission design and that orbital storage for months to years may be needed for missions involving orbital assembly. The power system design evolved during the study and has continued to evolve; the current design differs somewhat from that examined in this paper.

  10. Enhancing VHTR passive safety and economy with thermal radiation based direct reactor auxiliary cooling system

    SciTech Connect

    Zhao, H.; Zhang, H.; Zou, L.; Sun, X.

    2012-07-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The RVACS can be characterized as a surface-based decay heat removal system. It is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to the core volume) and decay heat removal capability (proportional to the vessel surface area). Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environmental side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps or annular regions formed between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions among the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very

  11. One Step Biomass Gas Reforming-Shift Separation Membrane Reactor

    SciTech Connect

    Roberts, Michael J.; Souleimanova, Razima

    2012-12-28

    GTI developed a plan where efforts were concentrated in 4 major areas: membrane material development, membrane module development, membrane process development, and membrane gasifier scale-up. GTI assembled a team of researchers to work in each area. Task 1.1 Ceramic Membrane Synthesis and Testing was conducted by Arizona State University (ASU), Task 1.2 Metallic Membrane Synthesis and Testing was conducted by the U.S. National Energy Technology Laboratory (NETL), Task 1.3 was conducted by SCHOTT, and GTI was to test all membranes that showed potential. The initial focus of the project was concentrated on membrane material development. Metallic and glass-based membranes were identified as hydrogen selective membranes under the conditions of the biomass gasification, temperatures above 700C and pressures up to 30 atmospheres. Membranes were synthesized by arc-rolling for metallic type membranes and incorporating Pd into a glass matrix for glass membranes. Testing for hydrogen permeability properties were completed and the effects of hydrogen sulfide and carbon monoxide were investigated for perspective membranes. The initial candidate membrane of Pd80Cu20 chosen in 2008 was selected for preliminary reactor design and cost estimates. Although the H2A analysis results indicated a $1.96 cost per gge H2 based on a 5A (micron) thick PdCu membrane, there was not long-term operation at the required flux to satisfy the go/no go decision. Since the future PSA case yielded a $2.00/gge H2, DOE decided that there was insufficient savings compared with the already proven PSA technology to further pursue the membrane reactor design. All ceramic membranes synthesized by ASU during the project showed low hydrogen flux as compared with metallic membranes. The best ceramic membrane showed hydrogen permeation flux of 0.03 SCFH/ft2 at the required process conditions while the metallic membrane, Pd80Cu20 showed a flux of 47.2 SCFH/ft2 (3 orders of magnitude difference). Results from

  12. Gas characterization system software acceptance test procedure

    SciTech Connect

    Vo, C.V.

    1996-02-27

    This document details the Software Acceptance Testing of gas characterization systems. The gas characterization systems will be used to monitor the vapor spaces of waste tanks known to contain measurable concentrations of flammable gases.

  13. NEUTRONIC REACTOR COUNTER METHOD AND SYSTEM

    DOEpatents

    Graham, C.B.; Spiewak, I.

    1960-05-31

    An improved method is given for controlling the rate of fission in circulating-fuel neutronic reactors in which the fuel is a homogeneous liquid containing fissionable material and a neutron moderator. A change in the rate of flssion is effected by preferentially retaining apart from the circulating fuel a variable amount of either fissionable material or moderator, thereby varying the concentration of fissionable material in the fuel. In the case of an aqueous fuel solution a portion of the water may be continuously vaporized from the circulating solution and the amount of condensate, or condensate plus make-up water, returned to the solution is varied to control the fission rate.

  14. Different Mechanisms for Establishing Liquid Walls in Advanced Reactor Systems

    NASA Astrophysics Data System (ADS)

    Hançerlioğullari, Aybaba; Cini, Mesut

    2013-04-01

    The APEX study is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around a fusion plasma. In this study the modeling of APEX hybrid reactor produced by using ARIES-RS hybrid reactor technology, was performed by using the Monte Carlo code and ENF/B-V-VI nuclear data. The most important feature of APEX hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity, good power transformation productivity the magnitude of the reactor's operational duration, low failure percentage, short maintenance time and the inclusion of the system's simple technology and material. Around the fusion chamber, molten salt Li2BeF4 and natural lithium were used as cooling materials. The result of the study indicated that fissile material production UF4 and ThF4 heavy metal salt increased nearly at the same percentage.

  15. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    SciTech Connect

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  16. Reactor Controllability of 3-Region-Core Molten Salt Reactor System - A Study on Load Following Capability

    SciTech Connect

    Takahisa Yamamoto; Koshi Mitachi; Masatoshi Nishio

    2006-07-01

    The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of fissile fuels (breeding). Thorium (Th) and uranium-233 ({sup 233}U) are fertile and fissile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650 K), that is both the reactor nuclear fuel and the coolant. The MSR system is one of the six advanced reactor concepts identified by the Generation IV International Forum (GIF) as a candidate for cooperative development. In the MSR system, fuel salt flows through a fuel duct constructed around a reactor core and fuel channel of a graphite moderator accompanied by fission reaction and heat generation, and flows out to an external-loop system consisted of a heat exchanger and a circulation pump. Due to the motion of fuel salt, delayed neutron precursors that are one of the source of neutron production make to change their position between the fission reaction and neutron emission events and decay even occur in the external loop system. Hence the reactivity and effective delayed neutron precursor fraction of the MSR system are lower than those of solid fuel reactor systems such as Boiling Water Reactors (BWRs) and Pressurised Water Reactor (PWRs). Since all of the presently operating nuclear power reactors utilize solid fuel, little attention had been paid to the MSR analysis of the reactivity loss and reactor characteristics change caused by the fuel salt circulation. Sides et al. and Shimazu et al. developed MSR analytical models based on the point reactor kinetics model to consider the effect of fuel salt flow. Their models represented a reactor as having six zones for fuel salt and three zones for the graphite moderator. Since their models employed the point reactor kinetics model and the rough temperature approximation, their results were not sufficiently accurate to

  17. Gas sampling system for matrix of semiconductor gas sensors

    NASA Astrophysics Data System (ADS)

    Jasinski, Grzegorz; Strzelczyk, Anna; Koscinski, Piotr

    2016-01-01

    Semiconductor gas sensors are popular commercial sensors applied in numerous gas detection systems. They are reliable, small, rugged and inexpensive. However, there are a few problem limiting the wider use of such sensors. Semiconductor gas sensor usually exhibits a low selectivity, low repeatability, drift of response, strong temperature and moisture influence on sensor properties. Sample flow rate is one of the parameters that influence sensors response what should be considered in the measurement system. This paper describes low cost module for controlling measured gas flow rate. The proposed equipment will be used as a component of electronic nose system employed for classifying and distinguishing different levels of contamination in air.

  18. Gas fired Advanced Turbine System

    SciTech Connect

    LeCren, R.T.; White, D.J.

    1993-01-01

    The primary objective of the first phase of the Advanced Gas Turbine System (ATS) program was the concept definition of an advanced engine system that meets efficiency and emission goals far exceeding those that can be provided with today`s equipment. The thermal efficiency goal for such an advanced industrial engine was set at 50% some 15 percentage points higher than current equipment levels. Exhaust emissions goals for oxides of nitrogen (NO{sub x}), carbon monoxide (CO), and unburned hydrocarbons (UH) were fixed at 8 parts per million by volume (ppmv), 20 ppmv, and 20 ppmv respectively, corrected to 15% oxygen (O{sub 2}) levels. Other goals had to be addressed; these involved reducing the cost of power produced by 10 percent and improving or maintaining the reliability, availability, and maintainability (RAM) at current levels. This advanced gas turbine was to be fueled with natural gas, and it had to embody features that would allow it bum coal or coal derived fuels.

  19. Reactor technology assessment and selection utilizing systems engineering approach

    SciTech Connect

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-12

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  20. Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

    SciTech Connect

    Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.

    2011-04-06

    Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.