Science.gov

Sample records for gas reactor simulator

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model

    NASA Technical Reports Server (NTRS)

    Kazeminezhad, F.; Anghaie, S.

    2008-01-01

    Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

  7. Modular high-temperature gas-cooled reactor core heatup accident simulations

    SciTech Connect

    Ball, S.J.; Conklin, J.C.

    1989-01-01

    The design features of the modular high-temperature gas-cooled reactor (HTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. Simulations of long-term loss-of-forced-convection (LOFC) accidents, both with and without depressurization of the primary coolant and with only passive cooling available to remove afterheat, have shown that maximum core temperatures stay below the point at which fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. 4 refs., 5 figs.

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. Reactor Simulator Testing Overview

    NASA Technical Reports Server (NTRS)

    Schoenfeld, Michael P.

    2013-01-01

    Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  14. Reactor Simulator Testing

    NASA Technical Reports Server (NTRS)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  15. Simulation and control of water-gas shift packed bed reactor with inter-stage cooling

    NASA Astrophysics Data System (ADS)

    Saw, S. Z.; Nandong, J.

    2016-03-01

    Water-Gas Shift Reaction (WGSR) has become one of the well-known pathways for H2 production in industries. The issue with WGSR is that it is kinetically favored at high temperatures but thermodynamically favored at low temperatures, thus requiring careful consideration in the control design in order to ensure that the temperature used does not deactivate the catalyst. This paper studies the effect of a reactor arrangement with an inter-stage cooling implemented in the packed bed reactor to look at its effect on outlet temperature. A mathematical model is developed based on one-dimensional heat and mass transfers which incorporate the intra-particle effects. It is shown that the placement of the inter-stage cooling and the outlet temperature exiting the inter-stage cooling have strong influence on the reaction conversion. Several control strategies are explored for the process. It is shown that a feedback- feedforward control strategy using Multi-scale Control (MSC) is effective to regulate the reactor temperature profile which is critical to maintaining the catalysts activity.

  16. Plug Flow Reactor Simulator

    SciTech Connect

    Larson, Richard S.

    1996-07-30

    PLUG is a computer program that solves the coupled steady state continuity, momentum, energy, and species balance equations for a plug flow reactor. Both homogeneous (gas-phase) and heterogenous (surface) reactions can be accommodated. The reactor may be either isothermal or adiabatic or may have a specified axial temperature or heat flux profile; alternatively, an ambient temperature and an overall heat-transfer coefficient can be specified. The crosssectional area and surface area may vary with axial position, and viscous drag is included. Ideal gas behavior and surface site conservation are assumed.

  17. Plug Flow Reactor Simulator

    1996-07-30

    PLUG is a computer program that solves the coupled steady state continuity, momentum, energy, and species balance equations for a plug flow reactor. Both homogeneous (gas-phase) and heterogenous (surface) reactions can be accommodated. The reactor may be either isothermal or adiabatic or may have a specified axial temperature or heat flux profile; alternatively, an ambient temperature and an overall heat-transfer coefficient can be specified. The crosssectional area and surface area may vary with axial position,more » and viscous drag is included. Ideal gas behavior and surface site conservation are assumed.« less

  18. Reactor Simulator Testing

    NASA Technical Reports Server (NTRS)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise Jon

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz. Keywords: fission, space power, nuclear, liquid metal, NaK.

  19. Simulator test to study hot-flow problems related to a gas cooled reactor

    NASA Technical Reports Server (NTRS)

    Poole, J. W.; Freeman, M. P.; Doak, K. W.; Thorpe, M. L.

    1973-01-01

    An advance study of materials, fuel injection, and hot flow problems related to the gas core nuclear rocket is reported. The first task was to test a previously constructed induction heated plasma GCNR simulator above 300 kW. A number of tests are reported operating in the range of 300 kW at 10,000 cps. A second simulator was designed but not constructed for cold-hot visualization studies using louvered walls. A third task was a paper investigation of practical uranium feed systems, including a detailed discussion of related problems. The last assignment resulted in two designs for plasma nozzle test devices that could be operated at 200 atm on hydrogen.

  20. 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.

  1. MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

    SciTech Connect

    Ball, S.J. )

    1991-10-01

    The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR.

  2. Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

    SciTech Connect

    McCulloch, R.W.; Post, D.W.; Lovell, R.T.; Snyder, S.D.

    1981-04-01

    Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relate this profile to that generated by the coils in completed fuel pin simulators.

  3. 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.

  4. Reactor Simulator Integration and Testing

    NASA Technical Reports Server (NTRS)

    Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  5. Study on bubbly flow behavior in natural circulation reactor by thermal-hydraulic simulation tests with SF6-Gas and ethanol liquid

    NASA Astrophysics Data System (ADS)

    Kondo, Yoshiyuki; Suga, Keishi; Hibi, Koki; Okazaki, Toshihiko; Komeno, Toshihiro; Kunugi, Tomoaki; Serizawa, Akimi; Yoneda, Kimitoshi; Arai, Takahiro

    2009-02-01

    An advanced experimental technique has been developed to simulate two-phase flow behavior in a light water reactor (LWR). The technique applies three kinds of methods; (1) use of sulfur-hexafluoride (SF6) gas and ethanol (C2H5OH) liquid at atmospheric temperature and a pressure less than 1.0MPa, where the fluid properties are similar to steam-water ones in the LWR, (2) generation of bubble with a sintering tube, which simulates bubble generation on heated surface in the LWR, (3) measurement of detailed bubble distribution data with a bi-optical probe (BOP), (4) and measurement of liquid velocities with the tracer liquid. This experimental technique provides easy visualization of flows by using a large scale experimental apparatus, which gives three-dimensional flows, and measurement of detailed spatial distributions of two-phase flow. With this technique, we have carried out experiments simulating two-phase flow behavior in a single-channel geometry, a multi-rod-bundle one, and a horizontal-tube-bundle one on a typical natural circulation reactor system. Those experiments have clarified a) a flow regime map in a rod bundle on the transient region between bubbly and churn flow, b) three-dimensional flow behaviour in rod-bundles where inter-subassembly cross-flow occurs, c) bubble-separation behavior with consideration of reactor internal structures. The data have given analysis models for the natural circulation reactor design with good extrapolation.

  6. Reactor Simulator Testing Overview

    NASA Technical Reports Server (NTRS)

    Schoenfeld, Michael P.

    2013-01-01

    OBJECTIVE: Integrated testing of the TDU components TESTING SUMMARY: a) Verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. b) Thermal test heat regeneration design aspect of a cold trap purification filter. c) Pump performance test at pump voltages up to 150 V (targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V). TESTING HIGHLIGHTS: a) Gas and vacuum ground support test equipment performed effectively for NaK fill, loop pressurization, and NaK drain operations. b) Instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  7. High cycle fatigue behavior of Incoloy 800H in a simulated high-temperature gas-cooled reactor helium environment

    SciTech Connect

    Soo, P.; Sabatini, R.L.; Epel, L.G.; Hare, J.R. Sr.

    1980-01-01

    The current study was an attempt to evaluate the high cycle fatigue strength of Incoloy 800H in a High-Temperature Gas-Cooled Reactor helium environment containing significant quantities of moisture. As-heat-treated and thermally-aged materials were tested to determine the effects of long term corrosion in the helium test gas. Results from in-helium tests were compared to those from a standard air environment. It was found that the mechanisms of fatigue failure were very complex and involved recovery/recrystallization of the surface ground layer on the specimens, sensitization, hardness changes, oxide scale integrity, and oxidation at the tips of propagation cracks. For certain situations a corrosion-fatigue process seems to be controlling. However, for the helium environment studied, there was usually no aging or test condition for which air gave a higher fatigue strength.

  8. Reactor simulations of the GEC reference cell reactor with an ICP source for Ar/Cl2 and BCl3/Cl2 gas mixtures.

    NASA Astrophysics Data System (ADS)

    Veerasingam, Ramana; Choi, Seung J.; Riley, Merle; Hoekstra, Robert; Kushner, Mark

    1996-10-01

    The ICP (inductively coupled plasma) device is a widely researched plasma etching technology to meet the stringent requirements of dielectric and metal etch for the next generation semiconductor wafers. At Sandia, the GEC reference cell has been modified to include a planar coil geometry to couple the RF power to the plasma inductively. Measurements of densities of electrons, CL-, Ar+, and recently of ion current flux have been made. In this paper, we will present results of simulations modeling the GEC cell for Ar/CL2 and BCL3/Cl2 gas mixtures using the HPEM and GEMINI code packages. Results will be parametrized with power, pressure, and gas mixture. In addition, simulations of the ion current flux using a sheath model developed at Sandia and the HPEM will be performed and compared to data.

  9. Interactive Reactor Simulation.

    ERIC Educational Resources Information Center

    Nuttall, Herbert E., Jr.; Himmelblau, David M.

    In the field of chemical engineering, interactive process models can simulate the dynamic behavior and analysis of chemical processes. DYFLO was the process simulation program selected as a foundation for development of interactive programs for computer-assisted instruction (CAI) in chemical engineering. Interactive Computing and time sharing…

  10. Packed-Bed Reactor Study of NETL Sample 196c for the Removal of Carbon Dioxide from Simulated Flue Gas Mixture

    SciTech Connect

    Hoffman, James S.; Hammache, Sonia; Gray, McMahan L.; Fauth Daniel J.; Pennline, Henry W.

    2012-04-24

    An amine-based solid sorbent process to remove CO2 from flue gas has been investigated. The sorbent consists of polyethylenimine (PEI) immobilized onto silica (SiO2) support. Experiments were conducted in a packed-bed reactor and exit gas composition was monitored using mass spectrometry. The effects of feed gas composition (CO2 and H2O), temperature, and simulated steam regeneration were examined for both the silica support as well as the PEI-based sorbent. The artifact of the empty reactor was also quantified. Sorbent CO2 capacity loading was compared to thermogravimetric (TGA) results to further characterize adsorption isotherms and better define CO2 working capacity. Sorbent stability was monitored by periodically repeating baseline conditions throughout the parametric testing and replacing with fresh sorbent as needed. The concept of the Basic Immobilized Amine Sorbent (BIAS) Process using this sorbent within a system where sorbent continuously flows between the absorber and regenerator was introduced. The basic tenet is to manipulate or control the level of moisture on the sorbent as it travels around the sorbent circulation path between absorption and regeneration stages to minimize its effect on regeneration heat duty.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Computer simulation of FCC riser reactors.

    SciTech Connect

    Chang, S. L.; Golchert, B.; Lottes, S. A.; Petrick, M.; Zhou, C. Q.

    1999-04-20

    A three-dimensional computational fluid dynamics (CFD) code, ICRKFLO, was developed to simulate the multiphase reacting flow system in a fluid catalytic cracking (FCC) riser reactor. The code solve flow properties based on fundamental conservation laws of mass, momentum, and energy for gas, liquid, and solid phases. Useful phenomenological models were developed to represent the controlling FCC processes, including droplet dispersion and evaporation, particle-solid interactions, and interfacial heat transfer between gas, droplets, and particles. Techniques were also developed to facilitate numerical calculations. These techniques include a hybrid flow-kinetic treatment to include detailed kinetic calculations, a time-integral approach to overcome numerical stiffness problems of chemical reactions, and a sectional coupling and blocked-cell technique for handling complex geometry. The copyrighted ICRKFLO software has been validated with experimental data from pilot- and commercial-scale FCC units. The code can be used to evaluate the impacts of design and operating conditions on the production of gasoline and other oil products.

  18. 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.

  19. Simulation of a marine nuclear reactor

    SciTech Connect

    Kusunoki, Tsuyoshi; Kyouya, Masahiko; Kobayashi, Hideo; Ochiai, Masaaki

    1995-02-01

    A Nuclear-powered ship Engineering Simulation SYstem (NESSY) has been developed by the Japan Atomic Energy Research Institute as an advanced design tool for research and development of future marine reactors. A marine reactor must respond to changing loads and to the ship`s motions because of the ship`s maneuvering and its presence in a marine environment. The NESSY has combined programs for the reactor plant behavior calculations and the ship`s motion calculations. Thus, it can simulate reactor power fluctuations caused by changing loads and the ship`s motions. It can also simulate the behavior of water in the pressurizer and steam generators. This water sloshes in response to the ship`s motions. The performance of NESSY has been verified by comparing the simulation calculations with the measured data obtained by experiments performed using the nuclear ship Mutsu. The effects of changing loads and the ship`s motions on the reactor behavior can be accurately simulated by NESSY.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  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. 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.

  7. 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.

  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. 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.

  11. G2 Gas Cloud Simulation

    NASA Video Gallery

    This simulation shows the future behavior of the G2 gas cloud now approaching Sgr A*, the supermassive black hole at the center of the Milky Way. X-ray emission from the cloud's tidal interaction w...

  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. High-Density Plasma Reactors: Simulations for Design

    NASA Technical Reports Server (NTRS)

    Hash, David B.; Meyyappan, Meyya; Arnold, James O. (Technical Monitor)

    1998-01-01

    The development of improved and more efficient plasma reactors is a costly process for the semiconductor industry. Until five years ago, the Industry made most of its advancements through a trial and error approach. More recently, the role of computational modeling in the design process has increased. Both conventional computational fluid dynamics (CFD) techniques like Navier-Stokes solvers as well as particle simulation methods are used to model plasma reactor flowfields. However, since high-density plasma reactors generally operate at low gas pressures on the order of 1 to 10 mTorr, a particle simulation may be necessary because of the failure of CFD techniques to model rarefaction effects. The direct simulation Monte Carlo method is the most widely accepted and employed particle simulation tool and has previously been used to investigate plasma reactor flowfields. A plasma DSMC code is currently under development at NASA Ames Research Center with its foundation as the object-oriented parallel Cornell DSMC code, MONACO. The present investigation is a follow up of a neutral flow investigation of the effects of process parameters as well as reactor design on etch rate and etch rate uniformity. The previous work concentrated on silicon etch of a chlorine flow in a configuration typical of electron cyclotron resonance (ECR) or helical resonator type reactors. The effects of the plasma on the dissociation chemistry were modeled by making assumptions about the electron temperature and number density. The electrons or ions themselves were not simulated.The present work extends these results by simulating the charged species.The electromagnetic fields are calculated such that power deposition is modeled self-consistently. Electron impact reactions are modeled along with mechanisms for charge exchange. An bipolar diffusion assumption is made whereby electrons remain tied to the ions. However, the velocities of tile electrons are allowed to be modified during collisions

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

    SciTech Connect

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

    1983-03-01

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

  17. Computational Fluid Dynamics Simulation of Fluidized Bed Polymerization Reactors

    SciTech Connect

    Fan, Rong

    2006-01-01

    , monovariate population balance, bivariate population balance, aggregation and breakage equation and DQMOM-Multi-Fluid model are described. In the last section of Chapter 3, numerical methods involved in the multi-fluid model and time-splitting method are presented. Chapter 4 is based on a paper about application of DQMOM to polydisperse gas-solid fluidized beds. Results for a constant aggregation and breakage kernel and a kernel developed from kinetic theory are shown. The effect of the aggregation success factor and the fragment distribution function are investigated. Chapter 5 shows the work on validation of mixing and segregation phenomena in gas-solid fluidized beds with a binary mixture or a continuous size distribution. The simulation results are compared with available experiment data and discrete-particle simulation. Chapter 6 presents the project with Univation Technologies on CFD simulation of a Polyethylene pilot-scale FB reactor, The fluid dynamics, mass/heat transfer and particle size distribution are investigated through CFD simulation and validated with available experimental data. The conclusions of this study and future work are discussed in Chapter 7.

  18. 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.

  19. Simulating VIIRS Observed Gas Flare

    NASA Astrophysics Data System (ADS)

    Hsu, F. C.

    2015-12-01

    VIIRS Nightfire (VNF) had been proved being able to effectively detect gas flares at night, and characterize their temperature and source size. [1] However, limited access to generally confidential gas flare operation measurements made it difficult to verify the output. Although flared gas volume is occasionally available, it is not common to log the temperature and flames size which directly links to VNF output. To understand the mechanism of gas flare and how VIIRS perceives the event, a platform is proposed to simulate the gas flare being observed by VIIRS. The methodology can be described in three steps. (1) Use CFD simulation software ISIS-3D to simulate a simple gas flare. [2] Scalar fields of temperature and species concentration related to combustion are extracted from the simulation. The instantaneous scalar can be determined from time-averaging or guess by stochastic time and space series (TASS) from single-point statistics [3]. (2) Model spectral radiance intensity of simulated gas flare using RADCAL. [4] RADCAL developed by NIST can accurately model the spectral radiance emitted on the direction of lineof-sight given the spatial profile of temperature and concentration of species. (3) Use radiative transfer modeling to calculate the energy propagated to VIIRS. The modeled radiation will then be weighted by the MODTRAN [5] modeled transmissivity over predefined atmosphere to the satellite, with geometrical effects considered. Such platform can help understanding how exactly VNF is measuring gas flares, and thus lead to more precise characterization of combustion events. [1] C. D. Elvidge et al, Remote Sensing, 2013[2] IRSN ISIS-3D[3] M. E. Kounalakis et al, ASME J. Heat Transfer, 1991 [4] W. L. Grosshandler, NIST Technical Note 1402, 1993 [5] A. Berk et al, MODTRAN 5.2.0.0 User's Manual

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Coupling Schemes for Multiphysics Reactor Simulation

    SciTech Connect

    Vijay Mahadeven; Jean Ragusa

    2007-11-01

    This report documents the progress of the student Vijay S. Mahadevan from the Nuclear Engineering Department of Texas A&M University over the summer of 2007 during his visit to the INL. The purpose of his visit was to investigate the physics-based preconditioned Jacobian-free Newton-Krylov method applied to physics relevant to nuclear reactor simulation. To this end he studied two test problems that represented reaction-diffusion and advection-reaction. These two test problems will provide the basis for future work in which neutron diffusion, nonlinear heat conduction, and a twophase flow model will be tightly coupled to provide an accurate model of a BWR core.

  7. 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.

  8. 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

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    NASA Astrophysics Data System (ADS)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  14. Radiation Hydrodynamic Simulations of an Inertial Fusion Energy Reactor Chamber

    NASA Astrophysics Data System (ADS)

    Sacks, Ryan Foster

    Inertial fusion energy reactors present great promise for the future as they are capable of providing baseline power with no carbon footprint. Simulation work regarding the chamber response and first wall insult is carried out using the 1-D BUCKY radiation hydrodynamics code for a variety of differing chamber fills, radii, chamber obstructions and first wall materials. Discussion of the first wall temperature rise, x-ray spectrum incident on the wall, shock timing and maximum overpressure are presented. An additional discussion of the impact of different gas opacities and their effect on overall chamber dynamics, including the formation of two shock fronts, is also presented. This work is performed under collaboration with Lawrence Livermore National Laboratory at the University of Wisconsin-Madison's Fusion Technology Institute.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Hydraulic characteristics simulation of an innovative self-agitation anaerobic baffled reactor (SA-ABR).

    PubMed

    Qi, Wei-Kang; Hojo, Toshimasa; Li, Yu-You

    2013-05-01

    An investigation was conducted on a self-agitation anaerobic baffled reactor (SA-ABR) with agitation caused solely by the release of stored gas. The compound in the reactor is mixed without the use of any mechanical equipment and electricity. The computational fluid dynamics (CFD) simulation used to provide details of the flow pattern and information about the agitation process and a solid basis for design and optimization purposes. Every self-agitation cycle could be separated into the pressure energy storage process, the exergonic process and the buffer stage. The reactor is regarded as the combination of continuous stirred tank reactor and a small plug flow reactor. The liquid level and diffusion varies widely depending on the length of the U-tube. The compound transition phenomenon in the 1st chamber mainly occurs during the energy exergonic process and buffer stage. The fluid-diffusion in the 3rd and 4th chambers mainly happens after the buffer period.

  20. 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.

  1. 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.

  2. 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).

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. HTGR (High Temperature Gas-Cooled Reactor) ingress analysis using MINET

    SciTech Connect

    Van Tuyle, G.J.; Yang, J.W.; Kroeger, P.G.; Mallen, A.N.; Aronson, A.L.

    1989-04-01

    Modeling of water/steam ingress into the primary (helium) cooling circuit of a High Temperature Gas-Cooled Reactor (HTGR) is described. This modeling was implemented in the MINET Code, which is a program for analyzing transients in intricate fluid flow and heat transfer networks. Results from the simulation of a water ingress event postulated for the Modular HTGR are discussed. 27 refs., 6 figs., 6 tabs.

  17. 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.

  18. Dynamic Simulation of Batch Photocatalytic Reactor (BPR) for Wastewater Treatment

    NASA Astrophysics Data System (ADS)

    Dutta, Suman

    2012-08-01

    Reactive dyes discharged from dyehouse causes a serious environmental problem. UV/TiO2 photocatalysis has been employed effectively for these organic dyes removal from dye-house effluent. This process produces less amount of non-toxic final product. In this paper a photocatalytic reactor has been designed for Reactive red 198 (RR198) removal from aqueous solution. The reactor is operating in batch mode. After each batch, TiO2 catalyst has been separated and recycled in the next batch. Mathematical model equation of this batch photocatalytic reactor (BPR) has been developed considering Langmuir-Hinshelwood kinetics. Simulation of BPR has been carried out using fourth order Runge-Kutta (RK) method and fifth order RK method (Butcher method). This simulation results can be used to develop an automatic photocatlytic reactor for industrial wastewater treatment. Catalyst activity decay and its effect on each batch have been incorporated in this model.

  19. 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.

  20. 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.

  1. 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.

  2. Retained Gas Sampler Calibration and Simulant Tests

    SciTech Connect

    CRAWFORD, B.A.

    2000-01-05

    This test plan provides a method for calibration of the retained gas sampler (RGS) for ammonia gas analysis. Simulant solutions of ammonium hydroxide at known concentrations will be diluted with isotopically labeled 0.04 M ammonium hydroxide solution. Sea sand solids will also be mixed with ammonium hydroxide solution and diluent to determine the accuracy of the system for ammonia gas analysis.

  3. 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)

  4. Simulator for SUPO, a Benchmark Aqueous Homogeneous Reactor (AHR)

    SciTech Connect

    Klein, Steven Karl; Determan, John C.

    2015-10-14

    A simulator has been developed for SUPO (Super Power) an aqueous homogeneous reactor (AHR) that operated at Los Alamos National Laboratory (LANL) from 1951 to 1974. During that period SUPO accumulated approximately 600,000 kWh of operation. It is considered the benchmark for steady-state operation of an AHR. The SUPO simulator was developed using the process that resulted in a simulator for an accelerator-driven subcritical system, which has been previously reported.

  5. 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.

  6. Morphology of Gas Release in Physical Simulants

    SciTech Connect

    Daniel, Richard C.; Burns, Carolyn A.; Crawford, Amanda D.; Hylden, Laura R.; Bryan, Samuel A.; MacFarlan, Paul J.; Gauglitz, Phillip A.

    2014-07-03

    This report documents testing activities conducted as part of the Deep Sludge Gas Release Event Project (DSGREP). The testing described in this report focused on evaluating the potential retention and release mechanisms of hydrogen bubbles in underground radioactive waste storage tanks at Hanford. The goal of the testing was to evaluate the rate, extent, and morphology of gas release events in simulant materials. Previous, undocumented scoping tests have evidenced dramatically different gas release behavior from simulants with similar physical properties. Specifically, previous gas release tests have evaluated the extent of release of 30 Pa kaolin and 30 Pa bentonite clay slurries. While both materials are clays and both have equivalent material shear strength using a shear vane, it was found that upon stirring, gas was released immediately and completely from bentonite clay slurry while little if any gas was released from the kaolin slurry. The motivation for the current work is to replicate these tests in a controlled quality test environment and to evaluate the release behavior for another simulant used in DSGREP testing. Three simulant materials were evaluated: 1) a 30 Pa kaolin clay slurry, 2) a 30 Pa bentonite clay slurry, and 3) Rayleigh-Taylor (RT) Simulant (a simulant designed to support DSGREP RT instability testing. Entrained gas was generated in these simulant materials using two methods: 1) application of vacuum over about a 1-minute period to nucleate dissolved gas within the simulant and 2) addition of hydrogen peroxide to generate gas by peroxide decomposition in the simulants over about a 16-hour period. Bubble release was effected by vibrating the test material using an external vibrating table. When testing with hydrogen peroxide, gas release was also accomplished by stirring of the simulant.

  7. Scalable parallel solution coupling for multi-physics reactor simulation.

    SciTech Connect

    Tautges, T. J.; Caceres, A.; Mathematics and Computer Science

    2009-01-01

    Reactor simulation depends on the coupled solution of various physics types, including neutronics, thermal/hydraulics, and structural mechanics. This paper describes the formulation and implementation of a parallel solution coupling capability being developed for reactor simulation. The coupling process consists of mesh and coupler initialization, point location, field interpolation, and field normalization. We report here our test of this capability on an example problem, namely, a reflector assembly from an advanced burner test reactor. Performance of this coupler in parallel is reasonable for the chosen problem size and range of processor counts. The runtime is dominated by startup costs, which amortize over the entire coupled simulation. Future efforts will include adding more sophisticated interpolation and normalization methods, to accommodate different numerical solvers used in various physics modules and to obtain better conservation properties for certain field types.

  8. CASL: The Consortium for Advanced Simulation of Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Kothe, Douglas B.

    2010-11-01

    Like the fusion community, the nuclear engineering community is embarking on a new computational effort to create integrated, multiphysics simulations. The Consortium for Advanced Simulation of Light Water Reactors (CASL), one of 3 newly-funded DOE Energy Innovation Hubs, brings together an exceptionally capable team that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated the Virtual Reactor (VR), will: 1) Enable the use of leadership-class computing for engineering design and analysis to improve reactor capabilities, 2) Promote an enhanced scientific basis and understanding by replacing empirically based design and analysis tools with predictive capabilities, 3) Develop a highly integrated multiphysics environment for engineering analysis through increased fidelity methods, and 4) Incorporate UQ as a basis for developing priorities and supporting, application of the VR tools for predictive simulation. In this presentation, we present the plans for CASL and comment on the similarity and differences with the proposed Fusion Simulation Project (FSP).

  9. 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.

  10. 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.

  11. Parallelization and automatic data distribution for nuclear reactor simulations

    SciTech Connect

    Liebrock, L.M.

    1997-07-01

    Detailed attempts at realistic nuclear reactor simulations currently take many times real time to execute on high performance workstations. Even the fastest sequential machine can not run these simulations fast enough to ensure that the best corrective measure is used during a nuclear accident to prevent a minor malfunction from becoming a major catastrophe. Since sequential computers have nearly reached the speed of light barrier, these simulations will have to be run in parallel to make significant improvements in speed. In physical reactor plants, parallelism abounds. Fluids flow, controls change, and reactions occur in parallel with only adjacent components directly affecting each other. These do not occur in the sequentialized manner, with global instantaneous effects, that is often used in simulators. Development of parallel algorithms that more closely approximate the real-world operation of a reactor may, in addition to speeding up the simulations, actually improve the accuracy and reliability of the predictions generated. Three types of parallel architecture (shared memory machines, distributed memory multicomputers, and distributed networks) are briefly reviewed as targets for parallelization of nuclear reactor simulation. Various parallelization models (loop-based model, shared memory model, functional model, data parallel model, and a combined functional and data parallel model) are discussed along with their advantages and disadvantages for nuclear reactor simulation. A variety of tools are introduced for each of the models. Emphasis is placed on the data parallel model as the primary focus for two-phase flow simulation. Tools to support data parallel programming for multiple component applications and special parallelization considerations are also discussed.

  12. 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.

  13. 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.

  14. The Consortium for Advanced Simulation of Light Water Reactors

    SciTech Connect

    Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

    2011-10-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  15. Application of Gamma code coupled with turbomachinery models for high temperature gas-cooled reactors

    SciTech Connect

    Chang Oh

    2008-02-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.

  16. 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

  17. Simulator platform for fast reactor operation and safety technology demonstration

    SciTech Connect

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.}

  3. 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.

  4. Simulation and optimization of an organic-impurity oxidization reactor with a fixed porous bed and an electric heating element

    NASA Astrophysics Data System (ADS)

    Gnezdilov, N. N.; Dobrego, K. V.; Kozlov, I. M.; Shmelev, E. S.

    2006-09-01

    A reactor for oxidization of low-caloric-value organic impurities contained in the air has been simulated. It comprises a tube with a recuperator, filled with a porous carcass mix, and includes a heating element. The influence of the heating-element placement, the heat losses through the upper cover of the reactor, the flow rate of a gas mixture, and the power of the heater on the maximum temperatures of the porous carcass and the gas and on the concentration of the incompletely oxidized organic impurity at the output of the reactor has been investigated. It is shown that, to burn an impurity completely, it will suffice to heat the gas δTe to 300 K. It has been established that it is best to place a heater at the level of the upper cut of the inner tube of the reactor.

  5. 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.

  6. Designing a SCADA system simulator for fast breeder reactor

    NASA Astrophysics Data System (ADS)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

    2016-04-01

    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  7. Materials Test-2 LOCA Simulation in the NRU Reactor

    SciTech Connect

    Barner, J. O.; Hesson, G. M.; King, I. L.; Marshall, R. K.; Parchen, L. J.; Pilger, J. P.; Rausch, W. N.; Russcher, G. E.; Webb, B. J.; Wildung, N. J.; Wilson, C. L.; Wismer, M. D.; Mohr, C. L.

    1982-03-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This third experiment of the program produced fuel cladding temperatures exceeding 1033 K (1400°F) for 155 s and resulted in eight ruptured fuel rods. Experiment data and initial results are presented in the form of photographs and graphical summaries.

  8. Coupled simulation of the reactor core using CUPID/MASTER

    SciTech Connect

    Lee, J. R.; Cho, H. K.; Yoon, H. Y.; Jeong, J. J.

    2012-07-01

    The CUPID is a component-scale thermal hydraulics code which is aimed for the analysis of transient two-phase flows in nuclear reactor components such as the reactor vessel, steam generator, containment. This code adopts a three-dimensional, transient, two-phase and three-field model, and includes physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. In the present paper, a multi-physics simulation was performed by coupling CUPID with a three dimensional neutron kinetics code, MASTER. MASTER is merged into CUPID as a dynamic link library (DLL). The APR1400 reactor core during a control rod drop/ejection accident was simulated as an example by adopting a porous media approach to employ a fuel assembly. The following sections present the numerical modeling for the reactor core, coupling of the kinetics code, and the simulation results. And also, a preliminary study for multi-scale simulation between CUPID and system-scaled thermal hydraulics code, MARS will be introduced as well. (authors)

  9. A MODEL TO ESTIMATE VOLUME CHANGE DUE TO RADIOLYTIC GAS BUBBLES AND THERMAL EXPANSION IN SOLUTION REACTORS

    SciTech Connect

    F. SOUTO; A HEGER

    2001-02-01

    Aqueous homogeneous solution reactors have been proposed for the production of medical isotopes. However, the reactivity effects of fuel solution volume change, due to formation of radiolytic gas bubbles and thermal expansion, have to be mitigated to allow steady-state operation of solution reactors. The results of the free run experiments analyzed indicate that the proposed model to estimate the void volume due to radiolytic gas bubbles and thermal expansion in solution reactors can accurately describe the observed behavior during the experiments. This void volume due to radiolytic gas bubbles and fuel solution thermal expansion can then be used in the investigation of reactivity effects in fissile solutions. In addition, these experiments confirm that the radiolytic gas bubbles are formed at a higher temperature than the fuel solution temperature. These experiments also indicate that the mole-weighted average for the radiolytic gas bubbles in uranyl fluoride solutions is about 1 {micro}m. Finally, it should be noted that another model, currently under development, would simulate the power behavior during the transient given the initial fuel solution level and density. The model is based on Monte Carlo simulation with the MCNP computer code [Briesmeister, 1997] to obtain the reactor reactivity as a function of the fuel solution density, which, in turn, changes due to thermal expansion and radiolytic gas bubble formation.

  10. Simulation of Ultrasonic-driven Gas Separations

    SciTech Connect

    Rector, David R.; Greenwood, Margaret S.; Ahmed, Salahuddin; Doctor, Steven R.; Posakony, Gerald J.; Stenkamp, Victoria S.

    2007-06-01

    The separation of components in a gas mixture is important for a wide range of applications. One method for achieving this separation is by passing a traveling acoustic wave through the gas mixture, which creates a flux of the lighter components away from the transducer. A series of simulation were performed to assess the effectiveness of this method for separating a binary mixture of argon and helium using the lattice kinetics method. The energy transport equation was modified to account for adiabatic expansion and compression. The species transport equation was modified to include a barodiffusion term. Simulations were performed on two different scales; detailed acoustic wave simulations to determine the net component flux as a function of local concentration, pressure, etc., and device scale simulations to predict the gas composition as a function of time inside a gas separation cylinder. The method is first validated using data from literature and then applied to mixtures of argon and helium. Results are presented and discussed.

  11. Simulation of ultrasonic-driven gas separations.

    PubMed

    Rector, David R; Greenwood, Margaret S; Ahmed, Salahuddin; Doctor, Steven R; Posakony, Gerry J; Stenkamp, V Susan

    2007-06-01

    The separation of components in a gas mixture is important for a wide range of applications. One method for achieving this separation is by passing a traveling acoustic wave through the gas mixture, which creates a flux of the lighter components away from the transducer. A series of simulations was performed to assess the effectiveness of this method for separating a binary mixture of argon and helium using the lattice kinetics method. The energy transport equation was modified to account for adiabatic expansion and compression. The species transport equation was modified to include a barodiffusion term. Simulations were performed on two different scales; detailed acoustic wave simulations to determine the net component flux as a function of local concentration, pressure, etc. and device scale simulations to predict the gas composition as a function of time inside a gas separation cylinder. The method is first validated using data from literature and then applied to mixtures of argon and helium. Results are presented and discussed.

  12. Thermohydraulic modeling and simulation of breeder reactors

    SciTech Connect

    Agrawal, A.K.; Khatib-Rahbar, M.; Curtis, R.T.; Hetrick, D.L.; Girijashankar, P.V.

    1982-01-01

    This paper deals with the modeling and simulation of system-wide transients in LMFBRs. Unprotected events (i.e., the presumption of failure of the plant protection system) leading to core-melt are not considered in this paper. The existing computational capabilities in the area of protected transients in the US are noted. Various physical and numerical approximations that are made in these codes are discussed. Finally, the future direction in the area of model verification and improvements is discussed.

  13. 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.

  14. 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.

  15. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    SciTech Connect

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

  16. Development of GAMMA Code and Evaluation for a Very High Temperature gas-Cooled Reactor

    SciTech Connect

    Oh, Chang H; Lim, H.S.; Kim, E.S.; NO, H.C.

    2007-06-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. This paper will also include what improvements will be made in the Gamma code for the VHTR.

  17. Reactor Subsystem Simulation for Nuclear Hybrid Energy Systems

    SciTech Connect

    Shannon Bragg-Sitton; J. Michael Doster; Alan Rominger

    2012-09-01

    Preliminary system models have been developed by Idaho National Laboratory researchers and are currently being enhanced to assess integrated system performance given multiple sources (e.g., nuclear + wind) and multiple applications (i.e., electricity + process heat). Initial efforts to integrate a Fortran-based simulation of a small modular reactor (SMR) with the balance of plant model have been completed in FY12. This initial effort takes advantage of an existing SMR model developed at North Carolina State University to provide initial integrated system simulation for a relatively low cost. The SMR subsystem simulation details are discussed in this report.

  18. Disruption simulation experiments and extrapolation to reactor conditions

    SciTech Connect

    Hassanein, A.; Konkashbaev, I.K.

    1998-08-01

    Laboratory experiments to simulate plasma disruptions have contributed significantly in many aspects to the understanding of the physical processes occurring during high-energy deposition on target material surfaces due to plasma instabilities. Laser light, electron beams, and plasma guns have been used worldwide to study disruption effects and erosion damage of candidate divertor materials. The differences among these simulation experiments are examined. The net power flux reaching the originally exposed surface depends on many parameters, such as type of energy deposited, target material, pulse duration, and geometrical factors. Experimental results have been evaluated and compared with theoretical predictions, and the overall relevance of simulation experiments to reactor conditions has been critically examined.

  19. 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.

  20. CFD Simulation of Slug Mixing in VVER-1000 Reactor

    SciTech Connect

    Vyskocil, Ladislav

    2006-07-01

    Recently, the safety analyses of VVER and PWR reactors have dealt with the possibility of reactivity-induced accidents related to the penetration of a water slug with low boron concentration into the reactor core. Loop seals at the reactor coolant pump (RCP) suction are the most likely places for the formation of these slugs. The slug is formed in the loop when there is neither natural nor forced circulation. When the circulation is restored, the slug travels towards the reactor and causes an insertion of positive reactivity in the core. This report deals with a CFD simulation of the most dangerous event - the start-up of the first RCP. Only several seconds are needed for slug to reach the core and the operator has no time for corrective action. Mixing of slug on its way to the core can reduce the danger of core recriticality. The primary objective of this study was to find out whether the FLUENT 6 CFD code is capable of predicting the mixing in the cold leg, downcomer and lower plenum as the slug moves toward the reactor core. Numerical simulations were based on mixing tests performed on 1:5 scale model of VVER-1000 reactor at the Gidropress Design Bureau, Russia. In the physical mixing tests, temperature was substituted for Boron concentration through the use of hot and cold water. The time history of core inlet average temperature was calculated by FLUENT and was found to be in good qualitative agreement with experimental data. This work was carried out as part of the EU project FLOMIX-R, Work Package 4. (author)

  1. 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.

  2. STUDY OF MERCURY OXIDATION BY SCR CATALYST IN AN ENTRAINED-FLOW REACTOR UNDER SIMULATED PRB CONDITIONS

    EPA Science Inventory

    A bench-scale entrained-flow reactor system was constructed for studying elemental mercury oxidation under selective catalytic reduction (SCR) reaction conditions. Simulated flue gas was doped with fly ash collected from a subbituminous Powder River Basin (PRB) coal-fired boiler ...

  3. 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.

  4. 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.

  5. Large-Scale Simulations of Realistic Fluidized Bed Reactors using Novel Numerical Methods

    NASA Astrophysics Data System (ADS)

    Capecelatro, Jesse; Desjardins, Olivier; Pepiot, Perrine; National Renewable Energy Lab Collaboration

    2011-11-01

    Turbulent particle-laden flows in the form of fluidized bed reactors display good mixing properties, low pressure drops, and a fairly uniform temperature distribution. Understanding and predicting the flow dynamics within the reactor is necessary for improving the efficiency, and providing technologies for large-scale industrialization. A numerical strategy based on an Eulerian representation of the gas phase and Lagrangian tracking of the particles is developed in the framework of NGA, a high- order fully conservative parallel code tailored for turbulent flows. The particles are accounted for using a point-particle assumption. Once the gas-phase quantities are mapped to the particle location a conservative, implicit diffusion operation smoothes the field. Normal and tangential collisions are handled via soft-sphere model, modified to allow the bed to reach close packing at rest. The pressure drop across the bed is compared with theory to accurately predict the minimum fluidization velocity. 3D simulations of the National Renewable Energy Lab's 4-inch reactor are then conducted. Tens of millions of particles are tracked. The reactor's geometry is modeled using an immersed boundary scheme. Statistics for volume fraction, velocities, bed expansion, and bubble characteristics are analyzed and compared with experimental data.

  6. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    NASA Technical Reports Server (NTRS)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  7. Hydraulic characteristics simulation of an innovative self-agitation anaerobic baffled reactor (SA-ABR).

    PubMed

    Qi, Wei-Kang; Hojo, Toshimasa; Li, Yu-You

    2013-05-01

    An investigation was conducted on a self-agitation anaerobic baffled reactor (SA-ABR) with agitation caused solely by the release of stored gas. The compound in the reactor is mixed without the use of any mechanical equipment and electricity. The computational fluid dynamics (CFD) simulation used to provide details of the flow pattern and information about the agitation process and a solid basis for design and optimization purposes. Every self-agitation cycle could be separated into the pressure energy storage process, the exergonic process and the buffer stage. The reactor is regarded as the combination of continuous stirred tank reactor and a small plug flow reactor. The liquid level and diffusion varies widely depending on the length of the U-tube. The compound transition phenomenon in the 1st chamber mainly occurs during the energy exergonic process and buffer stage. The fluid-diffusion in the 3rd and 4th chambers mainly happens after the buffer period. PMID:23567669

  8. METC CFD simulations of hot gas filtration

    SciTech Connect

    O`Brien, T.J.

    1995-06-01

    Computational Fluid Dynamic (CFD) simulations of the fluid/particle flow in several hot gas filtration vessels will be presented. These simulations have been useful in designing filtration vessels and in diagnosing problems with filter operation. The simulations were performed using the commercial code FLUENT and the METC-developed code MFIX. Simulations of the initial configuration of the Karhula facility indicated that the dirty gas flow over the filter assemblage was very non-uniform. The force of the dirty gas inlet flow was inducing a large circulation pattern that caused flow around the candles to be in opposite directions on opposite sides of the vessel. By introducing a system of baffles, a more uniform flow pattern was developed. This modification may have contributed to the success of the project. Several simulations of configurations proposed by Industrial Filter and Pump were performed, varying the position of the inlet. A detailed resolution of the geometry of the candles allowed determination of the flow between the individual candles. Recent simulations in support of the METC/CeraMem Cooperative Research and Development Agreement have analyzed the flow in the vessel during the cleaning back-pulse. Visualization of experiments at the CeraMem cold-flow facility provided confidence in the use of CFD. Extensive simulations were then performed to assist in the design of the hot test facility being built by Ahlstrom/Pyropower. These tests are intended to demonstrate the CeraMem technology.

  9. 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)

  10. 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

  11. Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

    NASA Technical Reports Server (NTRS)

    Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

    2007-01-01

    A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

  12. 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)

  13. A Computational Fluid Dynamic and Heat Transfer Model for Gaseous Core and Gas Cooled Space Power and Propulsion Reactors

    NASA Technical Reports Server (NTRS)

    Anghaie, S.; Chen, G.

    1996-01-01

    A computational model based on the axisymmetric, thin-layer Navier-Stokes equations is developed to predict the convective, radiation and conductive heat transfer in high temperature space nuclear reactors. An implicit-explicit, finite volume, MacCormack method in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve the thermal and fluid governing equations. Simulation of coolant and propellant flows in these reactors involves the subsonic and supersonic flows of hydrogen, helium and uranium tetrafluoride under variable boundary conditions. An enthalpy-rebalancing scheme is developed and implemented to enhance and accelerate the rate of convergence when a wall heat flux boundary condition is used. The model also incorporated the Baldwin and Lomax two-layer algebraic turbulence scheme for the calculation of the turbulent kinetic energy and eddy diffusivity of energy. The Rosseland diffusion approximation is used to simulate the radiative energy transfer in the optically thick environment of gas core reactors. The computational model is benchmarked with experimental data on flow separation angle and drag force acting on a suspended sphere in a cylindrical tube. The heat transfer is validated by comparing the computed results with the standard heat transfer correlations predictions. The model is used to simulate flow and heat transfer under a variety of design conditions. The effect of internal heat generation on the heat transfer in the gas core reactors is examined for a variety of power densities, 100 W/cc, 500 W/cc and 1000 W/cc. The maximum temperature, corresponding with the heat generation rates, are 2150 K, 2750 K and 3550 K, respectively. This analysis shows that the maximum temperature is strongly dependent on the value of heat generation rate. It also indicates that a heat generation rate higher than 1000 W/cc is necessary to maintain the gas temperature at about 3500 K, which is typical design temperature required to achieve high

  14. 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.

  15. 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.

  16. 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

  17. High-Temperature Jet Spray Reactor for the Preparation of Rare Earth Oxides by Pyrolysis: Computer Simulation

    NASA Astrophysics Data System (ADS)

    Zhao, Qiu-yue; Lv, Chao; Zhang, Zi-mu; Dou, Zhi-he; Zhang, Ting-an; Liu, Yan; Lv, Guo-zhi

    2014-09-01

    A new type of high-temperature jet spray pyrolysis (SP) reactor is investigated in this article as part of studies on the preparation of rare earth oxides at Northeastern University (NEU), Shenyang, China. The jet spray reactor examined here is a horizontal, tubular reactor conveying the hot products of the combustion of methane and oxygen with a converging-diverging jet section in an arrangement that provides for inspiration of LaCl3 solution to pyrolyze to La2O3 with the hot gas. The present article is concerned with a computer simulation using a computational fluid dynamic model to develop the velocity, temperature, and pressure profiles in the jet reactor since direct measurement is difficult. The article includes brief comments on a room-temperature model designed to examine the flow characteristics of the jet SP reactor. It was found that the velocity decreased at first, and then it increased near the jet throat. The highest velocity occurred at the throat of jet SP reactor where the LaCl3 enters the unit. Along the reactor axis, the temperature decreases with distance from the gas inlet. The lowest temperature zone was near the wall before the throat of the reactor due to wall heat losses. The temperature was estimated to be close to 1700 K at the throat of the reactor, and it was about 1300 K toward the exit of the reactor. It was shown that a reaction would take place mainly in the throat and in the vicinity of first contact between gas and induced spray. A negative pressure was produced as gas passes through the converging-diverging throat of the jet SP reactor that causes the LaCl3 solution to enter the throat of the reactor. While the investigations of this type of reactor are at an early stage, the results look promising. NEU continues to investigate this approach for the preparation of La2O3 based on high-temperature testwork and physical modeling techniques.

  18. High-Temperature Jet Spray Reactor for the Preparation of Rare Earth Oxides by Pyrolysis: Computer Simulation

    NASA Astrophysics Data System (ADS)

    Zhao, Qiu-yue; Lv, Chao; Zhang, Zi-mu; Dou, Zhi-he; Zhang, Ting-an; Liu, Yan; Lv, Guo-zhi

    2014-08-01

    A new type of high-temperature jet spray pyrolysis (SP) reactor is investigated in this article as part of studies on the preparation of rare earth oxides at Northeastern University (NEU), Shenyang, China. The jet spray reactor examined here is a horizontal, tubular reactor conveying the hot products of the combustion of methane and oxygen with a converging-diverging jet section in an arrangement that provides for inspiration of LaCl3 solution to pyrolyze to La2O3 with the hot gas. The present article is concerned with a computer simulation using a computational fluid dynamic model to develop the velocity, temperature, and pressure profiles in the jet reactor since direct measurement is difficult. The article includes brief comments on a room-temperature model designed to examine the flow characteristics of the jet SP reactor. It was found that the velocity decreased at first, and then it increased near the jet throat. The highest velocity occurred at the throat of jet SP reactor where the LaCl3 enters the unit. Along the reactor axis, the temperature decreases with distance from the gas inlet. The lowest temperature zone was near the wall before the throat of the reactor due to wall heat losses. The temperature was estimated to be close to 1700 K at the throat of the reactor, and it was about 1300 K toward the exit of the reactor. It was shown that a reaction would take place mainly in the throat and in the vicinity of first contact between gas and induced spray. A negative pressure was produced as gas passes through the converging-diverging throat of the jet SP reactor that causes the LaCl3 solution to enter the throat of the reactor. While the investigations of this type of reactor are at an early stage, the results look promising. NEU continues to investigate this approach for the preparation of La2O3 based on high-temperature testwork and physical modeling techniques.

  19. 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.

  20. Numerical simulation of synthesis gas incineration

    NASA Astrophysics Data System (ADS)

    Kazakov, A. V.; Khaustov, S. A.; Tabakaev, R. B.; Belousova, Y. A.

    2016-04-01

    The authors have analysed the expediency of the suggested low-grade fuels application method. Thermal processing of solid raw materials in the gaseous fuel, called synthesis gas, is investigated. The technical challenges concerning the applicability of the existing gas equipment developed and extensively tested exclusively for natural gas were considered. For this purpose computer simulation of three-dimensional syngas-incinerating flame dynamics was performed by means of the ANSYS Multiphysics engineering software. The subjects of studying were: a three-dimensional aerodynamic flame structure, heat-release and temperature fields, a set of combustion properties: a flare range and the concentration distribution of burnout reagents. The obtained results were presented in the form of a time-averaged pathlines with color indexing. The obtained results can be used for qualitative and quantitative evaluation of complex multicomponent gas incineration singularities.

  1. 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

  2. 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.

  3. LOCA simulation in the NRU reactor: materials test-1

    SciTech Connect

    Russcher, G.E.; Marshall, R.K.; Hesson, G.M.; Wildung, N.J.; Rausch, W.N.

    1981-10-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607/sup 0/F) and resulted in six ruptured fuel rods. Test data and initial results from the experiment are presented here in the form of photographs and graphical summaries. These results are also compared with the preceding prototypic thermal-hydraulic test results and with computer model test predictions.

  4. 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.

  5. Neutronic analysis stochastic distribution of fuel particles in Very High Temperature Gas-Cooled Reactors

    NASA Astrophysics Data System (ADS)

    Ji, Wei

    The Very High Temperature Gas-Cooled Reactor (VHTR) is a promising candidate for Generation IV designs due to its inherent safety, efficiency, and its proliferation-resistant and waste minimizing fuel cycle. A number of these advantages stem from its unique fuel design, consisting of a stochastic mixture of tiny (0.78mm diameter) microspheres with multiple coatings. However, the microsphere fuel regions represent point absorbers for resonance energy neutrons, resulting in the "double heterogeneity" for particle fuel. Special care must be taken to analyze this fuel in order to predict the spatial and spectral dependence of the neutron population in a steady-state reactor configuration. The challenges are considerable and resist brute force computation: there are over 1010 microspheres in a typical reactor configuration, with no hope of identifying individual microspheres in this stochastic mixture. Moreover, when individual microspheres "deplete" (e.g., burn the fissile isotope U-235 or transmute the fertile isotope U-238 (eventually) to Pu-239), the stochastic time-dependent nature of the depletion compounds the difficulty posed by the stochastic spatial mixture of the fuel, resulting in a prohibitive computational challenge. The goal of this research is to develop a methodology to analyze particle fuel randomly distributed in the reactor, accounting for the kernel absorptions as well as the stochastic depletion of the fuel mixture. This Ph.D. dissertation will address these challenges by developing a methodology for analyzing particle fuel that will be accurate enough to properly model stochastic particle fuel in both static and time-dependent configurations and yet be efficient enough to be used for routine analyses. This effort includes creation of a new physical model, development of a simulation algorithm, and application to real reactor configurations.

  6. An Implicit Solution Framework for Reactor Fuel Performance Simulation

    SciTech Connect

    Glen Hansen; Chris Newman; Derek Gaston; Cody Permann

    2009-08-01

    The simulation of nuclear reactor fuel performance involves complex thermomechanical processes between fuel pellets, made of fissile material, and the protective cladding that surrounds the pellets. An important design goal for a fuel is to maximize the life of the cladding thereby allowing the fuel to remain in the reactor for a longer period of time to achieve higher degrees of burnup. This presentation presents an initial approach for modeling the thermomechanical response of reactor fuel, and details of the solution method employed within INL's fuel performance code, BISON. The code employs advanced methods for solving coupled partial differential equation systems that describe multidimensional fuel thermomechanics, heat generation, and oxygen transport within the fuel. This discussion explores the effectiveness of a JFNK-based solution of a problem involving three dimensional fully coupled, nonlinear transient heat conduction and that includes pellet displacement and oxygen diffusion effects. These equations are closed using empirical data that is a function of temperature, density, and oxygen hyperstoichiometry. The method appears quite effective for the fuel pellet / cladding configurations examined, with excellent nonlinear convergence properties exhibited on the combined system. In closing, fully coupled solutions of three dimensional thermomechanics coupled with oxygen diffusion appear quite attractive using the JFNK approach described here, at least for configurations similar to those examined in this report.

  7. 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.

  8. Simulation of subsea gas hydrate exploitation

    NASA Astrophysics Data System (ADS)

    Janicki, Georg; Schlüter, Stefan; Hennig, Torsten; Deerberg, Görge

    2014-05-01

    The recovery of methane from gas hydrate layers that have been detected in several subsea sediments and permafrost regions around the world is a promising perspective to overcome future shortages in natural gas supply. Being aware that conventional natural gas resources are limited, research is going on to develop technologies for the production of natural gas from such new sources. Thus various research programs have started since the early 1990s in Japan, USA, Canada, India, and Germany to investigate hydrate deposits and develop required technologies. In recent years, intensive research has focussed on the capture and storage of CO2 from combustion processes to reduce climate impact. While different natural or man-made reservoirs like deep aquifers, exhausted oil and gas deposits or other geological formations are considered to store gaseous or liquid CO2, the storage of CO2 as hydrate in former methane hydrate fields is another promising alternative. Due to beneficial stability conditions, methane recovery may be well combined with CO2 storage in the form of hydrates. Regarding technological implementation many problems have to be overcome. Especially mixing, heat and mass transfer in the reservoir are limiting factors causing very long process times. Within the scope of the German research project »SUGAR« different technological approaches for the optimized exploitation of gas hydrate deposits are evaluated and compared by means of dynamic system simulations and analysis. Detailed mathematical models for the most relevant chemical and physical processes are developed. The basic mechanisms of gas hydrate formation/dissociation and heat and mass transport in porous media are considered and implemented into simulation programs. Simulations based on geological field data have been carried out. The studies focus on the potential of gas production from turbidites and their fitness for CO2 storage. The effects occurring during gas production and CO2 storage within

  9. Dynamic Modeling and Simulation Based Analysis of an Ammonia Borane (AB) Reactor System for Hydrogen Storage

    SciTech Connect

    Devarakonda, Maruthi N.; Holladay, Jamelyn D.; Brooks, Kriston P.; Rassat, Scot D.; Herling, Darrell R.

    2010-10-02

    Research on ammonia borane (AB, NH3BH3) has shown it to be a promising material for chemical hydrogen storage in PEM fuel cell applications. AB was selected by DOE’s Hydrogen Storage Engineering Center of Excellence (HSECoE) as the initial chemical hydride of study because of its high hydrogen storage capacity (up to 19.6% by weight for the release of three molar equivalents of hydrogen gas) and its stability under typical ambient conditions. A model of a bead reactor system which includes feed and product tanks, hot and cold augers, a ballast tank/reactor, a H2 burner and a radiator was developed to study AB system performance in an automotive application and estimate the energy, mass, and volume requirements for this off-board regenerable hydrogen storage material. Preliminary system simulation results for a start-up case and for a transient drive cycle indicate appropriate trends in the reactor system dynamics. A new controller was developed and validated in simulation for a couple of H2 demand cases.

  10. 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.

  11. 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.

  12. STUDY OF THE EFFECT OF CHLORINE ADDITION ON MERCURY OXIDATION BY SCR CATALYST UNDER SIMULATED SUBBITUMINOUS COAL FLUE GAS

    EPA Science Inventory

    An entrained flow reactor is used to study the effect of addition of chlorine-containing species on the oxidation of elemental mercury (Hgo)by a selective catalytic reduction (SCR) catalyst in simulated subbituminous coal combustion flue gas. The combustion flue gas was doped wit...

  13. Lattice gas simulations of replicating domains

    SciTech Connect

    Dawson, S.P.; Hasslacher, B.; Pearson, J.E.

    1993-12-31

    We use the lattice gas cellular automation (LGCA) developed to simulate a process of pattern-formation recently observed in reaction-diffusion systems. We study the reaction mechanism, which is an extension of the Selkov model for glycolytic oscillations. We are able to reproduce the self-replicating domains observed in this work. We use the LGCA simulation to estimate the smallest length-scale on which this process can occur under conditions encountered in the cell. These estimates are similar to those obtained for Turing patterns in the same setting.

  14. 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

  15. Simulations of gas clouds in interacting galaxies

    NASA Technical Reports Server (NTRS)

    Thomasson, Magnus

    1990-01-01

    A companion can induce a variety of morphological changes in a galaxy. The author uses N-body simulations to study the effects of different kinds of perturbations on the dynamics of a disk galaxy. The model is two-dimensional, with a disk consisting of about 60,000 particles. Most of the particles (80%) represent the old stellar population with a high velocity dispersion, while the rest (20%) represent gas clouds with a low velocity dispersion. Initially, the velocity dispersion corresponds to Q = 1 for the star particles, and Q = O for the gas particles, where Q is Toomre's (1964) stability parameter. The gas clouds can collide inelastically. The disk is stabilized by a rigid halo potential, and by the random motions of the old star particles. To simulate the effect of an encounter on the disk, a companion galaxy, modelled as a point mass, can move in a co-planar orbit around the disk. A complete description of the N-body code is found in Thomasson (1989). The spiral structures caused by a companion in first a direct and then a retrograde (with respect to the rotation of the disk) parabolic orbit are presented. The associated velocity fields suggest a way to observationally distinguish between leading and trailing spiral arms. The stability of the gas component in a disk in which tidally triggered infall of gas to the center occurs is studied. Finally, the author shows how a ring of gas can form in a disk as a result of a co-planar encounter with another galaxy.

  16. Use of computational fluid dynamics simulations for design of a pretreatment screw conveyor reactor.

    PubMed

    Berson, R Eric; Hanley, Thomas R

    2005-01-01

    Computational fluid dynamics simulations were employed to compare performance of various designs of a pretreatment screw conveyor reactor. The reactor consisted of a vertical screw used to create cross flow between the upward conveying solids and the downward flow of acid. Simulations were performed with the original screw design and a modified design in which the upper flights of the screw were removed. Results of the simulations show visually that the modified design provided favorable plug flow behavior within the reactor. Pressure drop across the length of the reactor without the upper screws in place was predicted by the simulations to be 5 vs 40 kPa for the original design.

  17. High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

    SciTech Connect

    Guillen, Donna; Greenwood, Lawrence R.; Parry, James

    2014-06-22

    A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

  18. Simulation results of liquid and plastic scintillator detectors for reactor antineutrino detection - A comparison

    NASA Astrophysics Data System (ADS)

    Kashyap, V. K. S.; Pant, L. M.; Mohanty, A. K.; Datar, V. M.

    2016-03-01

    A simulation study of two kinds of scintillation detectors has been done using GEANT4. We compare plastic scintillator and liquid scintillator based designs for detecting electron antineutrinos emitted from the core of reactors. The motivation for this study is to set up an experiment at the research reactor facility at BARC for very short baseline neutrino oscillation study and remote reactor monitoring.

  19. Numerical modeling of carrier gas flow in atomic layer deposition vacuum reactor: A comparative study of lattice Boltzmann models

    SciTech Connect

    Pan, Dongqing; Chien Jen, Tien; Li, Tao; Yuan, Chris

    2014-01-15

    This paper characterizes the carrier gas flow in the atomic layer deposition (ALD) vacuum reactor by introducing Lattice Boltzmann Method (LBM) to the ALD simulation through a comparative study of two LBM models. Numerical models of gas flow are constructed and implemented in two-dimensional geometry based on lattice Bhatnagar–Gross–Krook (LBGK)-D2Q9 model and two-relaxation-time (TRT) model. Both incompressible and compressible scenarios are simulated and the two models are compared in the aspects of flow features, stability, and efficiency. Our simulation outcome reveals that, for our specific ALD vacuum reactor, TRT model generates better steady laminar flow features all over the domain with better stability and reliability than LBGK-D2Q9 model especially when considering the compressible effects of the gas flow. The LBM-TRT is verified indirectly by comparing the numerical result with conventional continuum-based computational fluid dynamics solvers, and it shows very good agreement with these conventional methods. The velocity field of carrier gas flow through ALD vacuum reactor was characterized by LBM-TRT model finally. The flow in ALD is in a laminar steady state with velocity concentrated at the corners and around the wafer. The effects of flow fields on precursor distributions, surface absorptions, and surface reactions are discussed in detail. Steady and evenly distributed velocity field contribute to higher precursor concentration near the wafer and relatively lower particle velocities help to achieve better surface adsorption and deposition. The ALD reactor geometry needs to be considered carefully if a steady and laminar flow field around the wafer and better surface deposition are desired.

  20. Supercomputer applications in gas turbine flowfield simulation

    NASA Technical Reports Server (NTRS)

    Madavan, Nateri K.; Kelaita, Paul; Gavali, Sharad

    1990-01-01

    The numerical simulation of the unsteady three-dimensional viscous flow in a gas turbine stage is considered. Results from a three-dimensional time-accurate Navier-Stokes simulation of rotor-stator interaction in an axial turbine stage are presented. The present study uses a fine grid in the spanwise direction to better resolve the complex three-dimensional flowfield, and complements earlier reported coarse-grid calculations. Several different features of the flowfield are analyzed and compared to earlier calculations and to experimental data whenever possible. Computer animation techniques are used to visualize various unsteady three-dimensional features of the flow. The results demonstrate the capabilities of current computing hardware in obtaining accurate simulations of unsteady flows in turbomachines.

  1. Measurement of photon flux with a miniature gas ionization chamber in a Material Testing Reactor

    NASA Astrophysics Data System (ADS)

    Fourmentel, D.; Filliatre, P.; Villard, J. F.; Lyoussi, A.; Reynard-Carette, C.; Carcreff, H.

    2013-10-01

    Nuclear heating measurements in Material Testing Reactors (MTR) are crucial for the design of the experimental devices and the prediction of the temperature of the hosted samples. Nuclear heating in MTR materials (except fuel) is mainly due to the energy deposition by the photon flux. Therefore, the photon flux is a key input parameter for the computer codes which simulate nuclear heating and temperature reached by samples/devices under irradiation. In the Jules Horowitz MTR under construction at the CEA Cadarache, the maximal expected nuclear heating levels will be about 15 to 18 W g-1 and it will be necessary to assess this parameter with the best accuracy. An experiment was performed at the OSIRIS reactor to combine neutron flux, photon flux and nuclear heating measurements to improve the knowledge of the nuclear heating in MTR. There are few appropriate sensors for selective measurement of the photon flux in MTR even if studies and developments are ongoing. An experiment, called CARMEN-1, was conducted at the OSIRIS MTR and we used in particular a gas ionization chamber based on miniature fission chamber design to measure the photon flux. In this paper, we detail Monte-Carlo simulations to analyze the photon fluxes with ionization chamber measurements and we compare the photon flux calculations to the nuclear heating measurements. These results show a good accordance between photon flux measurements and nuclear heating measurement and allow improving the knowledge of these parameters.

  2. 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.

  3. 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.

  4. 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.

  5. Simulating regolith ejecta due to gas impingement

    NASA Astrophysics Data System (ADS)

    Chambers, Wesley Allen; Metzger, Philip; Dove, Adrienne; Britt, Daniel

    2016-10-01

    Space missions operating at or near the surface of a planet or small body must consider possible gas-regolith interactions, as they can cause hazardous effects or, conversely, be employed to accomplish mission goals. They are also directly related to a body's surface properties; thus understanding these interactions could provide an additional tool to analyze mission data. The Python Regolith Interaction Calculator (PyRIC), built upon a computational technique developed in the Apollo era, was used to assess interactions between rocket exhaust and an asteroid's surface. It focused specifically on threshold conditions for causing regolith ejecta. To improve this model, and learn more about the underlying physics, we have begun ground-based experiments studying the interaction between gas impingement and regolith simulant. Compressed air, initially standing in for rocket exhaust, is directed through a rocket nozzle at a bed of simulant. We assess the qualitative behavior of various simulants when subjected to a known maximum surface pressure, both in atmosphere and in a chamber initially at vacuum. These behaviors are compared to prior computational results, and possible flow patterns are inferred. Our future work will continue these experiments in microgravity through the use of a drop tower. These will use several simulant types and various pressure levels to observe the effects gas flow can have on target surfaces. Combining this with a characterization of the surface pressure distribution, tighter bounds can be set on the cohesive threshold necessary to maintain regolith integrity. This will aid the characterization of actual regolith distributions, as well as informing the surface operation phase of mission design.

  6. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    SciTech Connect

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J; Wilson, C L

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  7. 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...

  8. 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...

  9. 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...

  10. Large eddy simulation applications in gas turbines.

    PubMed

    Menzies, Kevin

    2009-07-28

    The gas turbine presents significant challenges to any computational fluid dynamics techniques. The combination of a wide range of flow phenomena with complex geometry is difficult to model in the context of Reynolds-averaged Navier-Stokes (RANS) solvers. We review the potential for large eddy simulation (LES) in modelling the flow in the different components of the gas turbine during a practical engineering design cycle. We show that while LES has demonstrated considerable promise for reliable prediction of many flows in the engine that are difficult for RANS it is not a panacea and considerable application challenges remain. However, for many flows, especially those dominated by shear layer mixing such as in combustion chambers and exhausts, LES has demonstrated a clear superiority over RANS for moderately complex geometries although at significantly higher cost which will remain an issue in making the calculations relevant within the design cycle. PMID:19531505

  11. 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.

  12. CASL Virtual Reactor Predictive Simulation: Grid-to-Rod Fretting Wear

    SciTech Connect

    Roger, Lu Y.; Karoutas, Zeses; Sham, Sam

    2011-01-01

    Grid-to-Rod Fretting (GTRF) wear is currently one of the main causes of fuel rod leaking in pressurized water reactors. The Consortium for Advanced Simulation of Light Water Reactors (CASL) has identified GTRF as one of the Challenge Problems that drive the requirement for the development and application of a modeling and simulation computational environment for predictive simulation of light water reactors. This paper presents fretting wear simulation methodology currently employed by Westinghouse, a CASL industrial partner, to address GTRF. The required advancements in the computational and materials science modeling areas to develop a predictive simulation environment by CASL to address GTRF are outlined.

  13. 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.

  14. 2D fluid model analysis for the effect of 3D gas flow on a capacitively coupled plasma deposition reactor

    NASA Astrophysics Data System (ADS)

    Kim, Ho Jun; Lee, Hae June

    2016-06-01

    The wide applicability of capacitively coupled plasma (CCP) deposition has increased the interest in developing comprehensive numerical models, but CCP imposes a tremendous computational cost when conducting a transient analysis in a three-dimensional (3D) model which reflects the real geometry of reactors. In particular, the detailed flow features of reactive gases induced by 3D geometric effects need to be considered for the precise calculation of radical distribution of reactive species. Thus, an alternative inclusive method for the numerical simulation of CCP deposition is proposed to simulate a two-dimensional (2D) CCP model based on the 3D gas flow results by simulating flow, temperature, and species fields in a 3D space at first without calculating the plasma chemistry. A numerical study of a cylindrical showerhead-electrode CCP reactor was conducted for particular cases of SiH4/NH3/N2/He gas mixture to deposit a hydrogenated silicon nitride (SiN x H y ) film. The proposed methodology produces numerical results for a 300 mm wafer deposition reactor which agree very well with the deposition rate profile measured experimentally along the wafer radius.

  15. Experimental and simulated dosimetry of the university of Utah TRIGA reactor

    NASA Astrophysics Data System (ADS)

    Marble, Benjamin James

    Simulated neutron and gamma transport enable the gamma dose to be estimated at the surface of the University of Utah TRIGA Reactor UUTR pool. These results are benchmarked against experimental results for model verification. This model is useful for future licensing and possible reactor power upgrades. MCNP5 was utilized for the UUTR simulation and comparison with thermoluminescent detectors TLDs.

  16. Long-term testing of the zinc titanate for desulfurization of hot coal gas in a fluidized-bed reactor

    SciTech Connect

    Jain, S.C.; Gupta, R.; Gangwal, S.K.

    1993-12-31

    Research Triangle Institute (RTI) under contract to the US Department of Energy (DOE), Morgantown energy Technology Center has recently completed a long-term test consisting of 100 sulfidation-regeneration cycles on a zinc titanate material intended for use as a high-temperature, regenerable sorbent to desulfurize coal-derived gas. The primary motivation for this development is to generate a more economical, environmentally superior, and reliable process to purify the product gas of coal gasifiers for use in gas turbines and fuel cells. This zinc titanate formulation (designated as ZT-4 and containing Zn-to-Ti in a molar ratio of 1.5) exhibited the best overall performance in terms of chemical reactivity, sulfur capacity, regenerability, structural properties and, most importantly, the attrition resistance based on multicycle testing of a number of sorbent formulations in a bench scale fluidized-bed reactor. The conditions in the test were -- desulfurization temperature: 750C (1382F); pressure: 1.52 MPa (220 psia); coal gas: simulated Texaco entrained-bed oxygen-blown gasifier gas containing 12,000 ppmv of H{sub 2}S; superficial gas velocity: 15 cm/s (0.49 ft/s). The ZT-4 sorbent used in this test was prepared using a granulation technique and 500 g of the sorbent in the 100 to 300 microns particle diameter range were used in a 5.1-cm (2-inch) i.d. stainless steel reactor.

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. Laboratory simulation of cometary neutral gas ionization

    NASA Astrophysics Data System (ADS)

    Chang, T.-F.; Rahman, H. U.; White, R. S.

    1989-05-01

    The laboratory simulation of the interaction of the solar wind with a comet is used to study the cometary neural gas ionization. The experiment is carried out in the UCR T-1 facility with an ice ball as the comet model. Photographs and data are taken with a variety of values of the solar wind velocity, interplanetary magnetic field (IMF), and comet configurations. The results show that the cometary neutral gas ionization depends on both the velocity of the solar wind and the interplanetary magnetic field. The plasma cloud surrounding the comet is visible only when the solar wind velocity and IMF are both above certain minimum values. This velocity dependent phenomena is explained by Alfven's critical ionization velocity effect. The critical magnetic field may be explained by assuming two stream lower hybrid instability as a triggering mechanism for the ionization of the neutral gas by plasma flow. Critical upper and lower limits for the magnetic field, required by anomalous ionization, are also derived that satisfy the experimental observations.

  3. Simulation of the Two-Dimensional Gasdynamic, Temperature, and Concentration Fields in an Injection Reactor of Chemical Vapor Deposition for Synthesis of Carbon Nanotube Arrays

    NASA Astrophysics Data System (ADS)

    Futko, S. I.; Chornyi, A. D.; Shulitskii, B. G.; Labunov, V. A.

    2016-01-01

    The two-dimensional axisymmetric gasdynamic, concentration, and heat fields arising in an injection reactor of chemical vapor deposition in the process of synthesis of arrays of carbon nanotubes in it from hydrocarbons and organometallic compounds were numerically simulated for the purpose of investigating the features of these fields. It was established that, even in the case of laminar flow of a gas mixture over the surface of a substrate positioned in this reactor, in it there arise vortices introducing a significant heterogeneity into the gas flow. The influence of changes in the gasdynamic and temperature fields in the indicated reactor on the characteristics of an array of carbon nanotubes grown on the surface of the substrate was analyzed. Parametric calculations of the dependences of the velocity of the gas flow, the gas temperature, and the concentration of reagents in the reactor on the hydrocarbon flow rate, the temperature of the process, and the length of the injection needle have been performed. These calculations have shown that the regimes of heating and mixing of reagents in an injection reactor of chemical vapor deposition correspond to those of an ideal-mixing reactor. The results obtained can be used for determining the conditions necessary for the growth of homogeneous arrays of carbon nanotubes with a high rate on the surface of a substrate in a reactor of chemical vapor deposition.

  4. Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)

    NASA Astrophysics Data System (ADS)

    Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

    2014-06-01

    Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated.

  5. 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.

  6. 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.

  7. 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.

  8. 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...

  9. 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

  10. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    SciTech Connect

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  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. 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

  13. Gas turbine system simulation: An object-oriented approach

    NASA Technical Reports Server (NTRS)

    Drummond, Colin K.; Follen, Gregory J.; Putt, Charles W.

    1993-01-01

    A prototype gas turbine engine simulation has been developed that offers a generalized framework for the simulation of engines subject to steady-state and transient operating conditions. The prototype is in preliminary form, but it successfully demonstrates the viability of an object-oriented approach for generalized simulation applications. Although object oriented programming languages are-relative to FORTRAN-somewhat austere, it is proposed that gas turbine simulations of an interdisciplinary nature will benefit significantly in terms of code reliability, maintainability, and manageability. This report elucidates specific gas turbine simulation obstacles that an object-oriented framework can overcome and describes the opportunity for interdisciplinary simulation that the approach offers.

  14. Gas-Grain Simulation Facility (GGSF)

    NASA Technical Reports Server (NTRS)

    Greenwald, Ken

    1992-01-01

    The goal of the Gas-Grain Simulation Facility project is to provide a microgravity laboratory to facilitate research relevant to exobiology (the study of the origin and evolution of life in the universe). Such a facility will also be useful in other areas of study important to NASA including planetary science, biology, atmospheric science, astrophysics, chemistry, and physics. To achieve this goal, the project will develop and support the GGSF, a modular facility-class payload planned for inclusion on Space Station Freedom. The GGSF will consist of the following: an experiment chamber(s) supported by subsystems that provide chamber environment regulation and monitoring capabilities; sample generation, injection, positioning, and retrieval capabilities; and computer control, data acquisition, and housekeeping capabilities. The facility will also provide analytical tools such as light-scattering measurement systems, aerosol size-spectrum measurement devices, and optical imaging systems.

  15. The computer simulation of 3d gas dynamics in a gas centrifuge

    NASA Astrophysics Data System (ADS)

    Borman, V. D.; Bogovalov, S. V.; Borisevich, V. D.; Tronin, I. V.; Tronin, V. N.

    2016-09-01

    We argue on the basis of the results of 2D analysis of the gas flow in gas centrifuges that a reliable calculation of the circulation of the gas and gas content in the gas centrifuge is possible only in frameworks of 3D numerical simulation of gas dynamics in the gas centrifuge (hereafter GC). The group from National research nuclear university, MEPhI, has created a computer code for 3D simulation of the gas flow in GC. The results of the computer simulations of the gas flows in GC are presented. A model Iguassu centrifuge is explored for the simulations. A nonaxisymmetric gas flow is produced due to interaction of the hypersonic rotating flow with the scoops for extraction of the product and waste flows from the GC. The scoops produce shock waves penetrating into a working camera of the GC and form spiral waves there.

  16. Finite Element Based Stress Analysis of Graphite Component in High Temperature Gas Cooled Reactor Core Using Linear and Nonlinear Irradiation Creep Models

    SciTech Connect

    Mohanty, Subhasish; Majumdar, Saurindranath

    2015-01-01

    Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

  17. System startup simulation for an in-core thermionic reactor with heat pipe cooling

    NASA Astrophysics Data System (ADS)

    Determan, William R.; Otting, William D.

    1992-01-01

    The heat pipe cooled thermionic (HPTI) reactor relies on in-core sodium heat pipes to provide a redundant means of cooling the 72 thermionic fuel elements (TFEs) which comprise the 40-kWe reactor core assembly. In-core heat pipe cooling was selected for the reactor design due to a requirement for multiple system on-orbit restarts over its lifetime. Powering up the reactor requires the in-core and radiator heat pipes to undergo a thaw cycle with a rapid ascension in power to their operating temperatures. The present study considers how fast the thaw-out and power ascension cycle can be safely accomplished within a reactor core. As part of the study, a transient startup simulator model of the heat pipe cooled reactor system was developed. Results of the startup transient simulation are provided.

  18. 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

  19. 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.

  20. 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.

  1. 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

  2. Programmable AC power supply for simulating power transient expected in fusion reactor

    SciTech Connect

    Halimi, B.; Suh, K. Y.

    2012-07-01

    This paper focus on control engineering of the programmable AC power source which has capability to simulate power transient expected in fusion reactor. To generate the programmable power source, AC-AC power electronics converter is adopted to control the power of a set of heaters to represent the transient phenomena of heat exchangers or heat sources of a fusion reactor. The International Thermonuclear Experimental Reactor (ITER) plasma operation scenario is used as the basic reference for producing this transient power source. (authors)

  3. System simulation of a multicell thermionic space power reactor

    NASA Astrophysics Data System (ADS)

    von Arx, Alan Vincent

    For many years, thermionic power has been considered for space application. The prominent feature of the power conversion system is that there are no moving parts. Although designs have been developed by various organizations, no comprehensive system models are known to exist which can simulate transient behavior of a multicell design nor is there a method to directly couple these models to other codes that can calculate variations in reactivity. Thus, a procedure has been developed to couple the performance calculations of a space nuclear reactor thermal/hydraulics code with a neutron diffusion code to analyze temperature feedback. Thermionic power is based on the thermionic emissions principle where free electrons in a conductor have sufficient energy to escape the surface. Kinetic energy is given to the electrons by heating the conductor. Specifically, a 48 kWe thermionic power converter system model has been developed and used to model startup and other transients. Less than 10% of the fuel heat is converted to electricity, and the rest is rejected to space via a heat pipe radiator. An electromagnetic pump circulates the liquid metal coolant. First, a startup transient model was developed which showed stable operation through ignition of the Thermionic Fuel Elements (TFEs) and thawing of the radiator heat pipes. Also, the model's capability was expanded to include two-phase heat transfer to model boiling using coupled mass and thermal energy conservation equations. The next step incorporated effects of reactivity feedback---showing that various mechanisms will prevent power and temperature run-up for a flow reduction scenario where the reactor control systems fail to respond. In particular, the Doppler effect was shown to counter a positive worth due to partial core voiding although steps must be taken to preclude film boiling in that high superheats will result in TFE failures. Finally, analysis of the core grid spacer location suggests it should be located at

  4. 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.

  5. 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.

  6. MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS

    SciTech Connect

    Cao, Yan; Gohar, Yousry

    2015-11-01

    In a periodic pulsed reactor, the reactor state varies periodically from slightly subcritical to slightly prompt supercritical for producing periodic power pulses. Such periodic state change is accomplished by a periodic movement of specific reactor parts, such as control rods or reflector sections. The analysis of such reactor is difficult to perform with the current reactor physics computer programs. Based on past experience, the utilization of the point kinetics approximations gives considerable errors in predicting the magnitude and the shape of the power pulse if the reactor has significantly different neutron life times in different zones. To accurately simulate the dynamics of this type of reactor, a Monte Carlo procedure using the transfer function TRCL/TR of the MCNP/MCNPX computer programs is utilized to model the movable reactor parts. In this paper, two algorithms simulating the geometry part movements during a neutron history tracking have been developed. Several test cases have been developed to evaluate these procedures. The numerical test cases have shown that the developed algorithms can be utilized to simulate the reactor dynamics with movable geometry parts.

  7. 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.

  8. 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.

  9. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Jumel, Stéphanie; Van-Duysen, Jean Claude

    2005-04-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called ';Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, …) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program.

  10. 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.

  11. Computer simulation of magnetization-controlled shunt reactors for calculating electromagnetic transients in power systems

    SciTech Connect

    Karpov, A. S.

    2013-01-15

    A computer procedure for simulating magnetization-controlled dc shunt reactors is described, which enables the electromagnetic transients in electric power systems to be calculated. It is shown that, by taking technically simple measures in the control system, one can obtain high-speed reactors sufficient for many purposes, and dispense with the use of high-power devices for compensating higher harmonic components.

  12. Interactive simulations of gas-turbine modular HTGR transients and heatup accidents

    SciTech Connect

    Ball, S.J.; Nypaver, D.J.

    1994-06-01

    An interactive workstation-based simulator has been developed for performing analyses of modular high-temperature gas-cooled reactor (MHTGR) core transients and accidents. It was originally developed at Oak Ridge National Laboratory for the US Nuclear Regulatory Commission to assess the licensability of the US Department of Energy (DOE) steam cycle design 350-MW(t) MHTGR. Subsequently, the code was modified under DOE sponsorship to simulate the 450-MW(t) Gas Turbine (GT) design and to aid in development and design studies. Features of the code (MORECA-GT) include detailed modeling of 3-D core thermal-hydraulics, interactive workstation capabilities that allow user/analyst or ``operator`` involvement in accident scenarios, and options for studying anticipated transients without scram (ATWS) events. In addition to the detailed models for the core, MORECA includes models for the vessel, Shutdown Cooling System (SCS), and Reactor Cavity Cooling System (RCCS), and core point kinetics to accommodate ATWS events. The balance of plant (BOP) is currently not modeled. The interactive workstation features include options for on-line parameter plots and 3-D graphic temperature profiling. The studies to date show that the proposed MHTGR designs are very robust and can generally withstand the consequences of even the extremely low probability postulated accidents with little or no damage to the reactor`s fuel or metallic components.

  13. 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

  14. High-temperature gas-cooled reactor safety studies for the Division of Reactor Safety Research. Quarterly progress report, July 1-September 30, 1981

    SciTech Connect

    Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Harrington, R.M.

    1982-01-01

    Development work continued on the accident dynamics simulation codes ORTAP, BLAST, and ORECA for the Fort St. Vrain (FSV) reactor. New steam line and main steam bypass system models were developed and incorporated into ORTAP. An initial simulation of the FSV prestressed concrete reactor vessel and liner cooling system was developed and tested for use in the severe accident sequence analysis task.

  15. 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

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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

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

    SciTech Connect

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

    1997-05-31

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

  2. Computational Fluid Dynamics Simulation of the Hydrogen Reduction of Magnetite Concentrate in a Laboratory Flash Reactor

    NASA Astrophysics Data System (ADS)

    Fan, De-Qiu; Sohn, H. Y.; Mohassab, Yousef; Elzohiery, Mohamed

    2016-08-01

    A three-dimensional computational fluid dynamics (CFD) model was developed to study the hydrogen reduction of magnetite concentrate particles in a laboratory flash reactor representing a novel flash ironmaking process. The model was used to simulate the fluid flow, heat transfer, and chemical reactions involved. The governing equations for the gas phase were solved in the Eulerian frame of reference while the particles were tracked in the Lagrangian framework. The change in the particle mass was related to the chemical reaction and the particle temperature was calculated by taking into consideration the heat of reaction, convection, and radiation. The stochastic trajectory model was used to describe particle dispersion due to turbulence. Partial combustion of H2 by O2 injected through a non-premixed burner was also simulated in this study. The partial combustion mechanism used in this model consisted of seven chemical reactions involving six species. The temperature profiles and reduction degrees obtained from the simulations satisfactorily agreed with the experimental measurements.

  3. 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.

  4. 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.

  5. 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.

  6. Self-consistent microwave field and plasma discharge simulations for a moderate pressure hydrogen discharge reactor

    NASA Astrophysics Data System (ADS)

    Hassouni, K.; Grotjohn, T. A.; Gicquel, A.

    1999-07-01

    A self-consistent two-dimensional model of the electromagnetic field and the plasma in a hydrogen discharge system has been developed and tested in comparison to experimental measurements. The reactor studied is a 25 cm diameter resonant cavity structure operating at 2.45 GHz with a silica belljar of 10 cm diameter and 17 cm height contained within the microwave cavity. The inside of the belljar where the discharge occurs contains a substrate holder of 5 cm diameter that is used to hold substrates for diamond deposition. The electromagnetic field model solves for the microwave fields using a finite difference time-domain solution of Maxwell's equations. The plasma model is a three energy mode (gas, molecular vibration, and electron) and nine species (H2, H, H(n=2), H(n=3), H+, H2+, H3+, H-, electron) model which accounts for non-Boltzmann electron distribution function and has 35 reactions. Simulated characteristics of the reactor in two dimensions include gas temperature, electron temperature, electron density, atomic hydrogen molar fraction, microwave power absorption, and microwave fields. Comparisons of the model are made with close agreement to several experimental measurements including coherent anti-Stokes Raman Spectroscopy measurement of H2 temperature versus position above the substrate, Doppler broadening optical emission spectroscopy (OES) measurements of H temperature versus pressure, actinometry measurements of the relative H atom concentration, Hα OES intensity measurements versus position, and microwave electric field measurements. The parameter range studied includes pressures of 2500-11 000 Pa, microwave powers of 300-2000 W, and three vertical positions of the substrate holder.

  7. A computer program for engineering simulations of space reactor system performance

    SciTech Connect

    Dobranich, D.

    1992-07-01

    Nuclear thermal propulsion systems are envisioned as a fast and efficient form of transportation for the exploration of space. Several nuclear reactor concepts have been proposed. This document discusses SAFSIM (System Analysis Flow SIMulator) which is an engineering computer program that allows the fluid mechanic, heat transfer, and reactor dynamic simulation of the entire propulsion system. SAFSIM currently contains three basic physics modules: (1) fluid mechanics, (2) heat transfer, and (3) reactor dynamics. All three modules are coupled to allow the prediction of system performance. The analyst can employ any or all of the physics modules as the problem dictates.

  8. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    SciTech Connect

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

    2010-09-28

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  9. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1980-September 30, 1980

    SciTech Connect

    Not Available

    1980-12-12

    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 progress in the screening test program is described: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, 950 and 1050/sup 0/C. Initiation of controlled purity helium creep-rupture testing in the intensive screening test program is discussed. In addition, the results of 1000-hour exposures at 750 and 850/sup 0/C on several experimental alloys are discussed.

  10. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1980-June 30, 1980

    SciTech Connect

    Not Available

    1980-11-14

    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 progress in the screening test program is described; this includes: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850 and 950/sup 0/C. The initiation of air creep-rupture testing in the intensive screening test program is discussed. In addition, the status of the data management system is described.

  11. Process Modeling Phase I Summary Report for the Advanced Gas Reactor Fuel Development and Qualification Program

    SciTech Connect

    Pannala, Sreekanth; Daw, C Stuart; Boyalakuntla, Dhanunjay S; FINNEY, Charles E A

    2006-09-01

    This report summarizes the results of preliminary work at Oak Ridge National Laboratory (ORNL) to demonstrate application of computational fluid dynamics modeling to the scale-up of a Fluidized Bed Chemical Vapor Deposition (FBCVD) process for nuclear fuels coating. Specifically, this work, referred to as Modeling Scale-Up Phase I, was conducted between January 1, 2006 and March 31, 2006 in support of the Advanced Gas Reactor (AGR) Program. The objective was to develop, demonstrate and "freeze" a version of ORNL's computational model of the TRI ISOtropic (TRISO) fuel-particle coating process that can be specifically used to assist coater scale-up activities as part of the production of AGR-2 fuel. The results in this report are intended to serve as input for making decisions about initiating additional FBCVD modeling work (referred to as Modeling Scale-Up Phase II) in support of AGR-2. The main computational tool used to implement the model is the general-purpose multiphase fluid-dynamics computer code known as MFIX (Multiphase Flow with Interphase eXchanges), which is documented in detail on the DOE-sponsored website http://www.mfix.org. Additional computational tools are also being developed by ORNL for post-processing MFIX output to efficiently summarize the important information generated by the coater simulations. The summarized information includes quantitative spatial and temporal measures (referred to as discriminating characteristics, or DCs) by which different coater designs and operating conditions can be compared and correlated with trends in product quality. The ORNL FBCVD modeling work is being conducted in conjunction with experimental coater studies at ORNL with natural uranium CO (NUCO) and surrogate fuel kernels. Data are also being obtained from ambient-temperature, spouted-bed characterization experiments at the University of Tennessee and theoretical studies of carbon and silicon carbide chemical vapor deposition kinetics at Iowa State

  12. 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.

  13. 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.

  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. 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.

  16. Coupled high fidelity thermal hydraulics and neutronics for reactor safety simulations

    SciTech Connect

    Vincent A. Mousseau; Hongbin Zhang; Haihua Zhao

    2008-09-01

    This work is a continuation of previous work on the importance of accuracy in the simulation of nuclear reactor safety transients. This work is qualitative in nature and future work will be more quantitative. The focus of this work will be on a simplified single phase nuclear reactor primary. The transient of interest investigates the importance of accuracy related to passive (inherent) safety systems. The transient run here will be an Unprotected Loss of Flow (ULOF) transient. Here the coolant pump is turned off and the un’SCRAM’ed reactor transitions from forced to free convection (Natural circulation). Results will be presented that show the difference that the first order in time truncation physics makes on the transient. The purpose of this document is to illuminate a possible problem in traditional reactor simulation approaches. Detailed studies need to be done on each simulation code for each transient analyzed to determine if the first order truncation physics plays an important role.

  17. 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.

  18. 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.

  19. Package Equivalent Reactor Networks as Reduced Order Models for Use with CAPE-OPEN Compliant Simulation

    SciTech Connect

    Meeks, E.; Chou, C. -P.; Garratt, T.

    2013-03-31

    Engineering simulations of coal gasifiers are typically performed using computational fluid dynamics (CFD) software, where a 3-D representation of the gasifier equipment is used to model the fluid flow in the gasifier and source terms from the coal gasification process are captured using discrete-phase model source terms. Simulations using this approach can be very time consuming, making it difficult to imbed such models into overall system simulations for plant design and optimization. For such system-level designs, process flowsheet software is typically used, such as Aspen Plus® [1], where each component where each component is modeled using a reduced-order model. For advanced power-generation systems, such as integrated gasifier/gas-turbine combined-cycle systems (IGCC), the critical components determining overall process efficiency and emissions are usually the gasifier and combustor. Providing more accurate and more computationally efficient reduced-order models for these components, then, enables much more effective plant-level design optimization and design for control. Based on the CHEMKIN-PRO and ENERGICO software, we have developed an automated methodology for generating an advanced form of reduced-order model for gasifiers and combustors. The reducedorder model offers representation of key unit operations in flowsheet simulations, while allowing simulation that is fast enough to be used in iterative flowsheet calculations. Using high-fidelity fluiddynamics models as input, Reaction Design’s ENERGICO® [2] software can automatically extract equivalent reactor networks (ERNs) from a CFD solution. For the advanced reduced-order concept, we introduce into the ERN a much more detailed kinetics model than can be included practically in the CFD simulation. The state-of-the-art chemistry solver technology within CHEMKIN-PRO allows that to be accomplished while still maintaining a very fast model turn-around time. In this way, the ERN becomes the basis for

  20. 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)

  1. 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.

  2. 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.

  3. 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)

  4. 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.

  5. Recirculation of reverse osmosis concentrate in lab-scale anaerobic and aerobic landfill simulation reactors.

    PubMed

    Morello, Luca; Cossu, Raffaello; Raga, Roberto; Pivato, Alberto; Lavagnolo, Maria Cristina

    2016-10-01

    Leachate treatment is a major issue in the context of landfill management, particularly in view of the consistent changes manifested over time in the quality and quantity of leachate produced, linked to both waste and landfill characteristics, which renders the procedure technically difficult and expensive. Leachate recirculation may afford a series of potential advantages, including improvement of leachate quality, enhancement of gas production, acceleration of biochemical processes, control of moisture content, as well as nutrients and microbe migration within the landfill. Recirculation of the products of leachate treatment, such as reverse osmosis (RO) concentrate, is a less common practice, with widespread controversy relating to its suitability, potential impacts on landfill management and future gaseous and leachable emissions. Scientific literature provides the results of only a few full-scale applications of concentrate recirculation. In some cases, an increase of COD and ammonium nitrogen in leachate was observed, coupled with an increase of salinity; which, additionally, might negatively affect performance of the RO plant itself. In other cases, not only did leachate production not increase significantly but the characteristics of leachate extracted from the well closest to the re-injection point also remained unchanged. This paper presents the results of lab-scale tests conducted in landfill simulation reactors, in which the effects of injection of municipal solid waste (MSW) landfill leachate RO concentrate were evaluated. Six reactors were managed with different weekly concentrate inputs, under both anaerobic and aerobic conditions, with the aim of investigating the short and long-term effects of this practice on landfill emissions. Lab-scale tests resulted in a more reliable identification of compound accumulation and kinetic changes than full-scale applications, further enhancing the development of a mass balance in which gaseous emissions and waste

  6. Recirculation of reverse osmosis concentrate in lab-scale anaerobic and aerobic landfill simulation reactors.

    PubMed

    Morello, Luca; Cossu, Raffaello; Raga, Roberto; Pivato, Alberto; Lavagnolo, Maria Cristina

    2016-10-01

    Leachate treatment is a major issue in the context of landfill management, particularly in view of the consistent changes manifested over time in the quality and quantity of leachate produced, linked to both waste and landfill characteristics, which renders the procedure technically difficult and expensive. Leachate recirculation may afford a series of potential advantages, including improvement of leachate quality, enhancement of gas production, acceleration of biochemical processes, control of moisture content, as well as nutrients and microbe migration within the landfill. Recirculation of the products of leachate treatment, such as reverse osmosis (RO) concentrate, is a less common practice, with widespread controversy relating to its suitability, potential impacts on landfill management and future gaseous and leachable emissions. Scientific literature provides the results of only a few full-scale applications of concentrate recirculation. In some cases, an increase of COD and ammonium nitrogen in leachate was observed, coupled with an increase of salinity; which, additionally, might negatively affect performance of the RO plant itself. In other cases, not only did leachate production not increase significantly but the characteristics of leachate extracted from the well closest to the re-injection point also remained unchanged. This paper presents the results of lab-scale tests conducted in landfill simulation reactors, in which the effects of injection of municipal solid waste (MSW) landfill leachate RO concentrate were evaluated. Six reactors were managed with different weekly concentrate inputs, under both anaerobic and aerobic conditions, with the aim of investigating the short and long-term effects of this practice on landfill emissions. Lab-scale tests resulted in a more reliable identification of compound accumulation and kinetic changes than full-scale applications, further enhancing the development of a mass balance in which gaseous emissions and waste

  7. The Addition of Noncondensable Gases into RELAP5-3D for Analysis of High Temperature Gas-Cooled Reactors

    SciTech Connect

    C. B. Davis; C. H. Oh

    2003-08-01

    Oxygen, carbon dioxide, and carbon monoxide have been added to the RELAP5-3D computer code as noncondensable gases to support analysis of high temperature gas-cooled reactors. Models of these gases are required to simulate the effects of air ingress on graphite oxidation following a loss-of-coolant accident. Correlations were developed for specific internal energy, thermal conductivity, and viscosity for each gas at temperatures up to 3000 K. The existing model for internal energy (a quadratic function of temperature) was not sufficiently accurate at these high temperatures and was replaced by a more general, fourth-order polynomial. The maximum deviation between the correlations and the underlying data was 2.2% for the specific internal energy and 7% for the specific heat capacity at constant volume. The maximum deviation in the transport properties was 4% for oxygen and carbon monoxide and 12% for carbon dioxide.

  8. Numerical simulation of boiling water reactor ventclearing hydrodynamics

    SciTech Connect

    Nichols, B.D.; Hirt, C.W.

    1980-02-01

    Pressure suppression pools used in nuclear reactors are subject to hydrodynamic processes involving complicated free surface configurations. A new numerical method, SOLA-VOF, developed to handle such problems is described and evaluated through comparisons with laboratory test data. Results from numerous computations provide a detailed understanding of the hydrodynamic phenomena associated with boiling water reactor vent-clearing processes. In addition, calculations show the sensitivity of the results to variations in the water vapor content, vent submergence depth, vent orifice size, and to the influence of fluid-structure interactions.

  9. Overview of the Consortium for the Advanced Simulation of Light Water Reactors (CASL)

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Franceschini, Fausto; Evans, Thomas M.; Gehin, Jess C.

    2016-02-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR) operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a first-principles approach. In order to pursue these goals, CASL has participation from laboratory, academic, and industry partners. These partners are pursuing the solution of ten major "Challenge Problems" in order to advance the state-of-the-art in reactor design and analysis to permit power uprates, higher burnup, life extension, and increased safety. At present, the problems being addressed by CASL are primarily reactor physics-oriented; however, this paper is intended to introduce CASL to the reactor dosimetry community because of the importance of reactor physics modelling and nuclear data to define the source term for that community and the applicability and extensibility of the transport methods being developed.

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. Benchmark for evaluation and validation of reactor simulations (BEAVRS)

    SciTech Connect

    Horelik, N.; Herman, B.; Forget, B.; Smith, K.

    2013-07-01

    Advances in parallel computing have made possible the development of high-fidelity tools for the design and analysis of nuclear reactor cores, and such tools require extensive verification and validation. This paper introduces BEAVRS, a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading patterns, and numerous in-vessel components. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from fifty-eight instrumented assemblies. Initial comparisons between calculations performed with MIT's OpenMC Monte Carlo neutron transport code and measured cycle 1 HZP test data are presented, and these results display an average deviation of approximately 100 pcm for the various critical configurations and control rod worth measurements. Computed HZP radial fission detector flux maps also agree reasonably well with the available measured data. All results indicate that this benchmark will be extremely useful in validation of coupled-physics codes and uncertainty quantification of in-core physics computational predictions. The detailed BEAVRS specification and its associated data package is hosted online at the MIT Computational Reactor Physics Group web site (http://crpg.mit.edu/), where future revisions and refinements to the benchmark specification will be made publicly available. (authors)

  15. 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.

  16. SAFSIM: A computer program for engineering simulations of space reactor system performance

    SciTech Connect

    Dobranich, D.

    1992-07-01

    SAFSIM (System Analysis Flow SIMulator) is a FORTRAN computer program that provides engineering simulations of user-specified flow networks at the system level. It includes fluid mechanics, heat transfer, and reactor dynamics capabilities. SAFSIM provides sufficient versatility to allow the simulation of almost any flow system, from a backyard sprinkler system to a clustered nuclear reactor propulsion system. In addition to versatility, speed and robustness are primary goals of SAFSIM. The current capabilities of SAFSIM are summarized, and some illustrative example results are presented.

  17. 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.

  18. 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.

  19. 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.

  20. Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

    NASA Astrophysics Data System (ADS)

    Turinsky, Paul J.; Kothe, Douglas B.

    2016-05-01

    The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics "core simulator" based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M

  1. 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.

  2. 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.

  3. 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.

  4. 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

  5. Study on Applicability of Numerical Simulation to Evaluation of Gas Entrainment From Free Surface

    SciTech Connect

    Kei Ito; Takaaki Sakai; Hiroyuki Ohshima

    2006-07-01

    An onset condition of gas entrainment (GE) due to free surface vortex has been studied to establish a design of fast breeder reactor with higher coolant velocity than conventional designs, because the GE might cause the reactor operation instability and therefore should be avoided. The onset condition of the GE has been investigated experimentally and theoretically, however, dependency of the vortex type GE on local geometry configuration of each experimental system and local velocity distribution has prevented researchers from formulating the universal onset condition of the vortex type GE. A real scale test is considered as an accurate method to evaluate the occurrence of the vortex type GE, but the real scale test is generally expensive and not useful in the design study of large and complicated FBR systems, because frequent displacement of inner equipments accompanied by the design change is difficult in the real scale test. Numerical simulation seems to be promising method as an alternative to the real scale test. In this research, to evaluate the applicability of the numerical simulation to the design work, numerical simulations were conducted on the basic experimental system of the vortex type GE. This basic experiment consisted of rectangular flow channel and two important equipments for vortex type GE in the channel, i.e. vortex generation and suction equipments. Generated vortex grew rapidly interacting with the suction flow and the grown vortex formed a free surface dent (gas core). When the tip of the gas core or the bubbles detached from the tip of the gas core reached the suction mouth, the gas was entrained to the suction tube. The results of numerical simulation under the experimental conditions were compared to the experiment in terms of velocity distributions and free surface shape. As a result, the numerical simulation showed qualitatively good agreement with experimental data. The numerical simulation results were similar to the experimental

  6. 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.

  7. Fusion core start-up, ignition and burn simulations of reversed-field pinch (RFP) reactors

    SciTech Connect

    Chu, Yuh-Yi

    1988-01-01

    A transient reactor simulation model is developed to investigate and simulate the start-up, ignition and burn of a reversed-field pinch reactor. The simulation is based upon a spatially averaged plasma balance model with field profiles obtained from MHD quasi-equilibrium analysis. Alpha particle heating is estimated from Fokker-Planck calculations. The instantaneous plasma current is derived from a self-consistent circuit analysis for plasma/coil/eddy current interactions. The simulation code is applied to the TITAN RFP reactor design which features a compact, high-power-density reversed-field pinch fusion system. A contour analysis is performed using the steady-state global plasma balance. The results are presented with contours of constant plasma current. A saddle point is identified in the contour plot which determines the minimum value of plasma current required to achieve ignition. An optimized start-up to ignition and burn path can be obtained by passing through the saddle point. The simulation code is used to study and optimize the start-up scenario. In the simulations of the TITAN RFP reactor, the OH-driven superconducting EF coils are found to deviate from the required equilibrium values as the induced plasma current increases. This results in the modification of superconducting EF coils and the addition of a set of EF trim coils. The design of the EF coil system is performed with the simulation code subject to the optimization of trim-coil power and current. In addition, the trim-coil design is subject to the constraints of vertical-field stability index and maintenance access. A power crowbar is also needed to prevent the superconducting EF coils from generating excessive vertical field. A set of basic results from the simulation of TITAN RFP reactor yield a picture of RFP plasma operation in a reactor. Investigations of eddy current are also presented. 145 refs., 37 figs., 2 tabs.

  8. 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.

  9. 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.

  10. Multi-physics nuclear reactor simulator for advanced nuclear engineering education

    SciTech Connect

    Yamamoto, A.

    2012-07-01

    Multi-physics nuclear reactor simulator, which aims to utilize for advanced nuclear engineering education, is being introduced to Nagoya Univ.. The simulator consists of the 'macroscopic' physics simulator and the 'microscopic' physics simulator. The former performs real time simulation of a whole nuclear power plant. The latter is responsible to more detail numerical simulations based on the sophisticated and precise numerical models, while taking into account the plant conditions obtained in the macroscopic physics simulator. Steady-state and kinetics core analyses, fuel mechanical analysis, fluid dynamics analysis, and sub-channel analysis can be carried out in the microscopic physics simulator. Simulation calculations are carried out through dedicated graphical user interface and the simulation results, i.e., spatial and temporal behaviors of major plant parameters are graphically shown. The simulator will provide a bridge between the 'theories' studied with textbooks and the 'physical behaviors' of actual nuclear power plants. (authors)

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Simulation of wastewater treatment by aerobic granules in a sequencing batch reactor based on cellular automata.

    PubMed

    Benzhai, Hai; Lei, Liu; Ge, Qin; Yuwan, Peng; Ping, Li; Qingxiang, Yang; Hailei, Wang

    2014-10-01

    In the present paper, aerobic granules were developed in a sequencing batch reactor (SBR) using synthetic wastewater, and 81 % of granular rate was obtained after 15-day cultivation. Aerobic granules have a 96 % BOD removal to the wastewater, and the reactor harbors a mount of biomass including bacteria, fungi and protozoa. In view of the complexity of kinetic behaviors of sludge and biological mechanisms of the granular SBR, a cellular automata model was established to simulate the process of wastewater treatment. The results indicate that the model not only visualized the complex adsorption and degradation process of aerobic granules, but also well described the BOD removal of wastewater and microbial growth in the reactor. Thus, CA model is suitable for simulation of synthetic wastewater treatment. This is the first report about dynamical and visual simulation of treatment process of synthetic wastewater in a granular SBR.

  20. 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

  1. Reducing numerical costs for core wide nuclear reactor CFD simulations by the Coarse-Grid-CFD

    NASA Astrophysics Data System (ADS)

    Viellieber, Mathias; Class, Andreas G.

    2013-11-01

    Traditionally complete nuclear reactor core simulations are performed with subchannel analysis codes, that rely on experimental and empirical input. The Coarse-Grid-CFD (CGCFD) intends to replace the experimental or empirical input with CFD data. The reactor core consists of repetitive flow patterns, allowing the general approach of creating a parametrized model for one segment and composing many of those to obtain the entire reactor simulation. The method is based on a detailed and well-resolved CFD simulation of one representative segment. From this simulation we extract so-called parametrized volumetric forces which close, an otherwise strongly under resolved, coarsely-meshed model of a complete reactor setup. While the formulation so far accounts for forces created internally in the fluid others e.g. obstruction and flow deviation through spacers and wire wraps, still need to be accounted for if the geometric details are not represented in the coarse mesh. These are modelled with an Anisotropic Porosity Formulation (APF). This work focuses on the application of the CGCFD to a complete reactor core setup and the accomplishment of the parametrization of the volumetric forces.

  2. Performance and microbial community analysis of the anaerobic reactor with coke oven gas biomethanation and in situ biogas upgrading.

    PubMed

    Wang, Wen; Xie, Li; Luo, Gang; Zhou, Qi; Angelidaki, Irini

    2013-10-01

    A new method for simultaneous coke oven gas (COG) biomethanation and in situ biogas upgrading in anaerobic reactor was developed in this study. The simulated coke oven gas (SCOG) (92% H2 and 8% CO) was injected directly into the anaerobic reactor treating sewage sludge through hollow fiber membrane (HFM). With pH control at 8.0, the added H2 and CO were fully consumed and no negative effects on the anaerobic degradation of sewage sludge were observed. The maximum CH4 content in the biogas was 99%. The addition of SCOG resulted in enrichment and dominance of homoacetogenetic genus Treponema and hydrogenotrophic genus Methanoculleus in the liquid, which indicated that H2 were converted to methane by both direct (hydrogenotrophic methanogenesis) and indirect (homoacetogenesis+aceticlastic methanogenesis) pathways in the liquid. However, the aceticlasitic genus Methanosaeta was dominant for archaea in the biofilm on the HFM, which indicated indirect (homoacetogenesis+aceticlastic methanogenesis) H2 conversion pathway on the biofilm.

  3. 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.

  4. 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.

  5. Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

    SciTech Connect

    Trond Bjornard; John Hockert

    2011-08-01

    The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC&A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC&A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC&A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR [Pty] and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC&A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR&D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present work summarized in

  6. 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.

  7. Gas-Liquid Two-Phase Flows Through Packed Bed Reactors in Microgravity

    NASA Technical Reports Server (NTRS)

    Motil, Brian J.; Balakotaiah, Vemuri

    2001-01-01

    The simultaneous flow of gas and liquid through a fixed bed of particles occurs in many unit operations of interest to the designers of space-based as well as terrestrial equipment. Examples include separation columns, gas-liquid reactors, humidification, drying, extraction, and leaching. These operations are critical to a wide variety of industries such as petroleum, pharmaceutical, mining, biological, and chemical. NASA recognizes that similar operations will need to be performed in space and on planetary bodies such as Mars if we are to achieve our goals of human exploration and the development of space. The goal of this research is to understand how to apply our current understanding of two-phase fluid flow through fixed-bed reactors to zero- or partial-gravity environments. Previous experiments by NASA have shown that reactors designed to work on Earth do not necessarily function in a similar manner in space. Two experiments, the Water Processor Assembly and the Volatile Removal Assembly have encountered difficulties in predicting and controlling the distribution of the phases (a crucial element in the operation of this type of reactor) as well as the overall pressure drop.

  8. Concept of an inherently-safe high temperature gas-cooled reactor

    SciTech Connect

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

    2012-06-06

    As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

  9. 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.

  10. Ignition of a combustible gas mixture by a laser spark excited in the reactor volume

    NASA Astrophysics Data System (ADS)

    Kazantsev, S. Yu.; Kononov, I. G.; Kossyi, I. A.; Tarasova, N. M.; Firsov, K. N.

    2009-03-01

    Ignition of a stoichiometric CH4: O2 mixture by a laser spark excited in the reactor volume is studied experimentally. It is found that the spark initiates a feebly radiating incomplete-combustion wave, which is much faster than the combustion wave, but is substantially slower than the detonation wave. With a time delay of 500-700 μs, a bright optical flash occupying the entire chamber volume is observed, which indicates fast (involving branching chain reactions) ignition of the gas mixture. A conclusion is drawn regarding the common nature of the process of ignition of a combustible gas mixture by a laser spark excited in the reactor volume and the previously investigated initiation of combustion by laser sparks excited at solid targets, high-power microwave discharges, and high-current gliding discharges.

  11. Ignition of a combustible gas mixture by a laser spark excited in the reactor volume

    SciTech Connect

    Kazantsev, S. Yu.; Kononov, I. G.; Kossyi, I. A.; Tarasova, N. M.; Firsov, K. N.

    2009-03-15

    Ignition of a stoichiometric CH{sub 4}: O{sub 2} mixture by a laser spark excited in the reactor volume is studied experimentally. It is found that the spark initiates a feebly radiating incomplete-combustion wave, which is much faster than the combustion wave, but is substantially slower than the detonation wave. With a time delay of 500-700 {mu}s, a bright optical flash occupying the entire chamber volume is observed, which indicates fast (involving branching chain reactions) ignition of the gas mixture. A conclusion is drawn regarding the common nature of the process of ignition of a combustible gas mixture by a laser spark excited in the reactor volume and the previously investigated initiation of combustion by laser sparks excited at solid targets, high-power microwave discharges, and high-current gliding discharges.

  12. Closures for Course-Grid Simulation of Fluidized Gas-Particle Flows

    SciTech Connect

    Sankaran Sundaresan

    2010-02-14

    Gas-particle flows in fluidized beds and riser reactors are inherently unstable, and they manifest fluctuations over a wide range of length and time scales. Two-fluid models for such flows reveal unstable modes whose length scale is as small as ten particle diameters. Yet, because of limited computational resources, gas-particle flows in large fluidized beds are invariably simulated by solving discretized versions of the two-fluid model equations over a coarse spatial grid. Such coarse-grid simulations do not resolve the small-scale spatial structures which are known to affect the macroscale flow structures both qualitatively and quantitatively. Thus there is a need to develop filtered two-fluid models which are suitable for coarse-grid simulations and capturing the effect of the small-scale structures through closures in terms of the filtered variables. The overall objective of the project is to develop validated closures for filtered two-fluid models for gas-particle flows, with the transport gasifier as a primary, motivating example. In this project, highly resolved three-dimensional simulations of a kinetic theory based two-fluid model for gas-particle flows have been performed and the statistical information on structures in the 100-1000 particle diameters length scale has been extracted. Based on these results, closures for filtered two-fluid models have been constructed. The filtered model equations and closures have been validated against experimental data and the results obtained in highly resolved simulations of gas-particle flows. The proposed project enables more accurate simulations of not only the transport gasifier, but also many other non-reacting and reacting gas-particle flows in a variety of chemical reactors. The results of this study are in the form of closures which can readily be incorporated into existing multi-phase flow codes such as MFIX (www.mfix.org). Therefore, the benefits of this study can be realized quickly. The training provided

  13. Representative Source Terms and the Influence of Reactor Attributes on Functional Containment in Modular High-Temperature Gas-Cooled Reactors

    SciTech Connect

    D. A. Petti; Hans Gougar; Dick Hobbins; Pete Lowry

    2013-11-01

    Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

  14. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    SciTech Connect

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  15. Comments on the critical issues for a high-density gas-embedded Z-pinch CTR reactor concept

    NASA Astrophysics Data System (ADS)

    Ekdahl, C. A.

    1981-02-01

    The issues considered critical for the realization of a CTR reactor are discussed for the high density gas embedded z pinch concept. A simple model of the pinch reactor is presented that gives estimates for Q in substantial agreement with elaborate computer studies. An attempt is made to identify such critical points that are amenable to exploratory experiments. There appears to be no fatal flaw in the issues covered that would preclude a viable reactor based on this concept.

  16. Computational fluid dynamics simulation and redesign of a screw conveyor reactor.

    PubMed

    Wan, Yinkun; Hanley, Thomas R

    2004-01-01

    National Renewable Energy Laboratory (NREL) designed a shrinking-bed reactor to maintain a constant bulk packing density of cellulosic biomass. The high solid-to-liquid ratio in the pretreatment process allows a high sugar yield and avoids the need to flush large volumes of solution through the reactor. To scale up the shrinking-bed reactor, NREL investigated a pilot-scale screw conveyor reactor in which an interrupted flight between screws was employed to mimic the "shrinking-bed" effect. In the experiments with the screw conveyor reactor, overmixing and uneven flow occurred. These phenomena produce negative effects on biomass hydrolysis. The flow behavior inside the reactor was analyzed to allow redesign of the screw to achieve adequate mixing and even flow. In the present study, computational fluid dynamics (CFD) was utilized to simulate the fluid flow in the porous media, and a new screw design was proposed. CFD analysis performed on the redesigned reactor indicated that an even flow pattern was achieved.

  17. 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.

  18. Neutronics analysis of an open-cycle high-impulse gas core reactor concept

    NASA Technical Reports Server (NTRS)

    Whitmarsh, C. L., Jr.

    1972-01-01

    A procedure was developed to calculate the critical fuel mass, including the effects of propellant pressure, for coaxial-flow gas-core reactors operating at 196,600 newtons thrust and 4400 seconds specific impulse. Data were generated for a range of cavity diameter, reflector-moderator thickness, and quantity of structural material. Also presented are such core characteristics as upper limits on cavity pressure, spectral hardening in very-high-temperature hydrogen, and reactivity coefficients.

  19. NGNP: High Temperature Gas-Cooled Reactor Key Definitions, Plant Capabilities, and Assumptions

    SciTech Connect

    Wayne Moe

    2013-05-01

    This document provides key definitions, plant capabilities, and inputs and assumptions related to the Next Generation Nuclear Plant to be used in ongoing efforts related to the licensing and deployment of a high temperature gas-cooled reactor. These definitions, capabilities, and assumptions were extracted from a number of NGNP Project sources such as licensing related white papers, previously issued requirement documents, and preapplication interactions with the Nuclear Regulatory Commission (NRC).

  20. INTEGRATION OF HIGH TEMPERATURE GAS REACTORS WITH IN SITU OIL SHALE RETORTING

    SciTech Connect

    Eric P. Robertson; Michael G. McKellar; Lee O. Nelson

    2011-05-01

    This paper evaluates the integration of a high-temperature gas-cooled reactor (HTGR) to an in situ oil shale retort operation producing 7950 m3/D (50,000 bbl/day). The large amount of heat required to pyrolyze the oil shale and produce oil would typically be provided by combustion of fossil fuels, but can also be delivered by an HTGR. Two cases were considered: a base case which includes no nuclear integration, and an HTGR-integrated case.

  1. 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.

  2. 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.

  3. 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.

  4. Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

    SciTech Connect

    Not Available

    1986-10-01

    This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc. (C-E).

  5. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    SciTech Connect

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

  6. Functionally gradient material for membrane reactors to convert methane gas into value-added products

    DOEpatents

    Balachandran, Uthamalingam; Dusek, Joseph T.; Kleefisch, Mark S.; Kobylinski, Thadeus P.

    1996-01-01

    A functionally gradient material for a membrane reactor for converting methane gas into value-added-products includes an outer tube of perovskite, which contacts air; an inner tube which contacts methane gas, of zirconium oxide, and a bonding layer between the perovskite and zirconium oxide layers. The bonding layer has one or more layers of a mixture of perovskite and zirconium oxide, with the layers transitioning from an excess of perovskite to an excess of zirconium oxide. The transition layers match thermal expansion coefficients and other physical properties between the two different materials.

  7. Functionally gradient material for membrane reactors to convert methane gas into value-added products

    DOEpatents

    Balachandran, U.; Dusek, J.T.; Kleefisch, M.S.; Kobylinski, T.P.

    1996-11-12

    A functionally gradient material for a membrane reactor for converting methane gas into value-added-products includes an outer tube of perovskite, which contacts air; an inner tube which contacts methane gas, of zirconium oxide, and a bonding layer between the perovskite and zirconium oxide layers. The bonding layer has one or more layers of a mixture of perovskite and zirconium oxide, with the layers transitioning from an excess of perovskite to an excess of zirconium oxide. The transition layers match thermal expansion coefficients and other physical properties between the two different materials. 7 figs.

  8. Hot waste-to-energy flue gas treatment using an integrated fluidised bed reactor.

    PubMed

    Bianchini, A; Pellegrini, M; Saccani, C

    2009-04-01

    This paper describes an innovative process to increase superheated steam temperatures in waste-to-energy (WTE) plants. This solution is mainly characterised by a fluidised bed reactor in which hot flue gas is treated both chemically and mechanically. This approach, together with gas recirculation, increases the energy conversion efficiency, and raises the superheated steam temperature without decreasing the useful life of the superheater. This paper presents new experimental data obtained from the test facility installed at the Hera S.p.A. WTE plant in Forlì, Italy; discusses changes that can be implemented to increase the duration of experimental testing; offers suggestions for the design of an industrial solution.

  9. 3-D kinetics simulations of the NRU reactor using the DONJON code

    SciTech Connect

    Leung, T. C.; Atfield, M. D.; Koclas, J.

    2006-07-01

    The NRU reactor is highly heterogeneous, heavy-water cooled and moderated, with online refuelling capability. It is licensed to operate at a maximum power of 135 MW, with a peak thermal flux of approximately 4.0 x 10{sup 18} n.m{sup -2} . s{sup -1}. In support of the safe operation of NRU, three-dimensional kinetics calculations for reactor transients have been performed using the DONJON code. The code was initially designed to perform space-time kinetics calculations for the CANDU{sup R} power reactors. This paper describes how the DONJON code can be applied to perform neutronic simulations for the analysis of reactor transients in NRU, and presents calculation results for some transients. (authors)

  10. Dispersed bubble reactor for enhanced gas-liquid-solids contact and mass transfer

    DOEpatents

    Vimalchand, Pannalal; Liu, Guohai; Peng, WanWang; Bonsu, Alexander

    2016-01-26

    An apparatus to promote gas-liquid contact and facilitate enhanced mass transfer. The dispersed bubble reactor (DBR) operates in the dispersed bubble flow regime to selectively absorb gas phase constituents into the liquid phase. The dispersion is achieved by shearing the large inlet gas bubbles into fine bubbles with circulating liquid and additional pumped liquid solvent when necessary. The DBR is capable of handling precipitates that may form during absorption or fine catalysts that may be necessary to promote liquid phase reactions. The DBR can be configured with multistage counter current flow sections by inserting concentric cylindrical sections into the riser to facilitate annular flow. While the DBR can absorb CO.sub.2 in liquid solvents that may lead to precipitates at high loadings, it is equally capable of handling many different types of chemical processes involving solids (precipitates/catalysts) along with gas and liquid phases.

  11. WIPP panel simulations with gas generation

    SciTech Connect

    DeVries, K.L.; Callahan, G.D.

    1996-02-01

    An important issue in nuclear waste repository performance is the potential for fracture development resulting in pathways for release of radionuclides beyond the confines of the repository. A series of demonstration calculations using structural finite element analyses are presented here to examine the effect of internal gas generation on the response of a sealed repository. From the calculated stress fields, the most probable location for a fracture to develop was determined to be within the pillars interior to the repository for the range of parameter values considered. If a fracture interconnects the rooms and panels of the repository, fracture opening produces significant additional void volume to limit the excess gas pressure to less than 1.0 MPa above the overburden pressure. Consequently, the potential for additional fracture development into the barrier pillar is greatly reduced, which provides further confidence that the waste will be contained within the repository.

  12. Numerical Simulations of Low Pressure Inductively Coupled Plasmas in Geometrically Complex Reactors

    NASA Astrophysics Data System (ADS)

    Yu, Ben; Wu, Hanming; Krishnan, Anantha

    1996-10-01

    A two-dimensional fluid model has been developed for simulation of low pressure inductively coupled plasma (ICP) reactors. The model obtains solutions for the plasma density, electron temperature, and electric field for the given operating conditions. The physical phenomena and processes such as ambipolar diffusion, thermal diffusion, quasi-neutrality, ionization, inductive Joule heating, and excitations are considered in the model. A significant feature of the model is its capability of handling complex geometries that are often encountered in industrial reactors. Complex reactor geometries are modeled by a body-fitted-coordinate (BFC) formulation. A series of numerical experiments have been conducted using the model to study effects of various parameters such as chamber pressure, size of the wafer, position of the inductive coil, and the power input into the plasma. Different reactor geometries such as the GEC ICP reference cell and the belljar reactor have been simulated. The results of the parametric experiments are presented to show certain systematic trends in performance parameters such as uniformity and processing rates. The ICP model has been coupled to a computational fluid dynamics (CFD) code (capable of 3D simulations) that obtains the flow and pressure distribution inside the chamber. The ICP model will use pressure predictions (from the CFD model) to compute the local ionization rates. Chemical source/sink terms from the plasma dissociation model will be used by the CFD code to account for local reactant depletion effects.

  13. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  14. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  15. Lattice gas hydrodynamics: Theory and simulations

    SciTech Connect

    Hasslacher, B.

    1993-01-01

    The first successful application of a microscopic analogy to create a skeleton cellular automaton and analyze it with statistical mechanical tools, was the work of Frisch, Hasslacher and Pomeau on the Navier-Stokes equation in two and three dimensions. This has become a very large research area with lattice gas models and methods being used for both fundamental investigations into the foundations of statistical mechanics and a large number of diverse applications. This present research was devoted to enlarging the fundamental scope of lattice gas models and proved successful. Since the beginning of this proposal, cellular automata have been constructed for statistical mechanical models, fluids, diffusion and shock systems in fundamental investigations. In applied areas, there are now excellent lattice gas models for complex flows through porous media, chemical reaction and combustion dynamics, multiphase flow systems, and fluid mixtures with natural boundaries. With extended cellular fluid models, one can do problems with arbitrary pairwise potentials. Recently, these have been applied to such problems as non-newtonian or polymeric liquids and a mixture of immiscible fluids passing through fractal or spongelike media in two and three dimensions. This proposal has contributed to and enlarged the scope of this work.

  16. Plasma core reactor simulations using RF uranium seeded argon discharges

    NASA Technical Reports Server (NTRS)

    Roman, W. C.

    1976-01-01

    Experimental results are described in which pure uranium hexafluoride was injected into an argon-confined, steady-state, RF-heated plasma to investigate characteristics of plasma core nuclear reactors. The 80 kW (13.56 MHz) and 1.2 MW (5.51 MHz) rf induction heater facilities were used to determine a test chamber flow scheme which offered best uranium confinement with minimum wall coating. The cylindrical fused-silica test chamber walls were 5.7-cm-ID by 10-cm-long. Test conditions included RF powers of 2-85 kW, chamber pressures of 1-12 atm, and uranium hexafluoride mass-flow rates of 0.005-0.13 g/s. Successful techniques were developed for fluid-mechanical confinement of RF-heated plasmas with pure uranium hexafluoride injection.

  17. Plasma core reactor simulations using RF uranium seeded argon discharges

    NASA Technical Reports Server (NTRS)

    Roman, W. C.

    1975-01-01

    An experimental investigation was conducted using the United Technologies Research Center (UTRC) 80 kW and 1.2 MW RF induction heater systems to aid in developing the technology necessary for designing a self-critical fissioning uranium plasma core reactor (PCR). A nonfissioning, steady-state RF-heated argon plasma seeded with pure uranium hexafluoride (UF6) was used. An overall objective was to achieve maximum confinement of uranium vapor within the plasma while simultaneously minimizing the uranium compound wall deposition. Exploratory tests were conducted using the 80 kW RF induction heater with the test chamber at approximately atmospheric pressure and discharge power levels on the order of 10 kW. Four different test chamber flow configurations were tested to permit selection of the configuration offering the best confinement characteristics for subsequent tests at higher pressure and power in the 1.2 MW RF induction heater facility.

  18. Dynamic Comparison of Three- and Four-Equation Reactor Core Models in a Full-Scope Power Plant Training Simulator

    SciTech Connect

    Espinosa-Paredes, Gilberto; Alvarez-Ramirez, Jose; Nunez-Carrera, Alejandro; Garcia-Gutierrez, Alfonso; Martinez-Mendez, Elizabeth Jeannette

    2004-02-15

    A comparative analysis of the dynamic behavior of a boiling water reactor in a full-scope power plant simulator for operator training is presented. Three- and four-equation reactor core models were used to examine three transients following tests described in acceptance test procedures: scram, loss of feedwater flow, and closure of main isolation valves. The three-equation model consists of water and steam mixture momentum, including mass and energy balances. The four-equation model is based on liquid and gas phase mass balances, together with a drift-flux approach for the analysis of phase separation. Analysis of the models showed that the scram transient was slightly different for three- and four-equation models. The drift-flux effects can explain such differences. Regarding the loss-of-feedwater transient, the predicted steam flow after scram is larger for the three-equation model. Finally, for the transient related to the closure of main steam isolation valves, the three-equation model provides slightly different results for the pressure change, which affects reactor level behavior.

  19. Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, April 1, 1983-September 30, 1983

    SciTech Connect

    Not Available

    1983-12-01

    An assessment of the HTGR opportunities from the year 2000 through 2045 was the principal activity on the Market Definition Task (WBS 03). Within the Plant Technology (WBS 13) task, there were activities to develop analytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. The activities in support of the HTGR-SC/C Lead Plant (WBS 30 and 31) were the participation in the Lead Plant System Engineering (LPSE) effort and the plant simulation task. The efforts on the Advanced HTGR systems was performed under the Modular Reactor Systems (MRS) (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors.

  20. Simulation of the neutron flux in the irradiation facility at RA-3 reactor.

    PubMed

    Bortolussi, S; Pinto, J M; Thorp, S I; Farias, R O; Soto, M S; Sztejnberg, M; Pozzi, E C C; Gonzalez, S J; Gadan, M A; Bellino, A N; Quintana, J; Altieri, S; Miller, M

    2011-12-01

    A facility for the irradiation of a section of patients' explanted liver and lung was constructed at RA-3 reactor, Comisión Nacional de Energía Atómica, Argentina. The facility, located in the thermal column, is characterized by the possibility to insert and extract samples without the need to shutdown the reactor. In order to reach the best levels of security and efficacy of the treatment, it is necessary to perform an accurate dosimetry. The possibility to simulate neutron flux and absorbed dose in the explanted organs, together with the experimental dosimetry, allows setting more precise and effective treatment plans. To this end, a computational model of the entire reactor was set-up, and the simulations were validated with the experimental measurements performed in the facility.

  1. Sequential UASB and dual media packed-bed reactors for domestic wastewater treatment - experiment and simulation.

    PubMed

    Rodríguez-Gómez, Raúl; Renman, Gunno

    2016-01-01

    A wastewater treatment system composed of an upflow anaerobic sludge blanket (UASB) reactor followed by a packed-bed reactor (PBR) filled with Sorbulite(®) and Polonite(®) filter material was tested in a laboratory bench-scale experiment. The system was operated for 50 weeks and achieved very efficient total phosphorus (P) removal (99%), 7-day biochemical oxygen demand removal (99%) and pathogenic bacteria reduction (99%). However, total nitrogen was only moderately reduced in the system (40%). A model focusing on simulation of organic material, solids and size of granules was then implemented and validated for the UASB reactor. Good agreement between the simulated and measured results demonstrated the capacity of the model to predict the behaviour of solids and chemical oxygen demand, which is critical for successful P removal and recovery in the PBR. PMID:27332842

  2. Gas-grain simulation experiment module conceptual design and gas-grain simulation facility breadboard development

    NASA Technical Reports Server (NTRS)

    Zamel, James M.; Petach, Michael; Gat, Nahum; Kropp, Jack; Luong, Christina; Wolff, Michael

    1993-01-01

    This report delineates the Option portion of the Phase A Gas-Grain Simulation Facility study. The conceptual design of a Gas-Grain Simulation Experiment Module (GGSEM) for Space Shuttle Middeck is discussed. In addition, a laboratory breadboard was developed during this study to develop a key function for the GGSEM and the GGSF, specifically, a solid particle cloud generating device. The breadboard design and test results are discussed and recommendations for further studies are included. The GGSEM is intended to fly on board a low earth orbit (LEO), manned platform. It will be used to perform a subset of the experiments planned for the GGSF for Space Station Freedom, as it can partially accommodate a number of the science experiments. The outcome of the experiments performed will provide an increased understanding of the operational requirements for the GGSF. The GGSEM will also act as a platform to accomplish technology development and proof-of-principle experiments for GGSF hardware, and to verify concepts and designs of hardware for GGSF. The GGSEM will allow assembled subsystems to be tested to verify facility level operation. The technology development that can be accommodated by the GGSEM includes: GGSF sample generation techniques, GGSF on-line diagnostics techniques, sample collection techniques, performance of various types of sensors for environmental monitoring, and some off-line diagnostics. Advantages and disadvantages of several LEO platforms available for GGSEM applications are identified and discussed. Several of the anticipated GGSF experiments require the deagglomeration and dispensing of dry solid particles into an experiment chamber. During the GGSF Phase A study, various techniques and devices available for the solid particle aerosol generator were reviewed. As a result of this review, solid particle deagglomeration and dispensing were identified as key undeveloped technologies in the GGSF design. A laboratory breadboard version of a solid

  3. The Role of Bottom Simulating Reflectors in Gas Hydrate Assessment

    NASA Astrophysics Data System (ADS)

    Majumdar, U.; Shedd, W. W.; Cook, A.; Frye, M.

    2015-12-01

    In this research we test the viability of using a bottom simulating reflector (BSR) to detect gas hydrate. Bottom simulating reflectors (BSRs) occur at many gas hydrate sites near the thermodynamic base of the gas hydrate stability zone (GHSZ), and are frequently used to identify possible presence of gas hydrate on a regional scale. To find if drilling a BSR actually increases the chances of finding gas hydrate, we combine an updated dataset of BSR distribution from the Bureau of Ocean Energy Management with a comprehensive dataset of natural gas hydrate distribution as appraised from well logs, covering an area of around 200,000 square kilometers in the northern Gulf of Mexico. The BSR dataset compiles industry 3-D seismic data, and includes mostly good-quality and high-confidence traditional and non-traditional BSRs. Resistivity well logs were used to identify the presence of gas hydrate from over 700 existing industry wells and we have found over 110 wells with likely gas hydrate occurrences. By integrating the two datasets, our results show that the chances of encountering gas hydrate when drilling through a BSR is ~ 42%, while that when drilling outside the BSR is ~15%. Our preliminary analysis indicates that a positive relationship exists between BSRs and gas hydrate accumulations, and the chances of encountering gas hydrate increases almost three-fold when drilling through a BSR. One interesting observation is that ~ 58% of the wells intersecting a BSR show no apparent evidence of gas hydrate. In this case, a BSR may occur at sites with no gas hydrate accumulations due to the presence of very low concentration of free gas that is not detected on resistivity logs. On the other hand, in a few wells, accumulations of gas hydrate were observed where no BSR is present. For example in a well in Atwater Valley Block 92, two intervals of gas hydrate accumulation in fractures have been identified on resistivity logs, of which, the deeper interval has 230 feet thick

  4. LWR (Light Water Reactor) power plant simulations using the AD10 and AD100 systems

    SciTech Connect

    Wulff, W.; Cheng, H.S.; Chien, C.J.; Jang, J.Y.; Lin, H.C.; Mallen, A.N.; Wang, S.J.; Institute of Nuclear Energy Research, Lung-Tan; Tawian Power Co., Taipei; Brookhaven National Lab., Upton, NY; Institute of Nuclear Energy Research, Lung-Tan )

    1989-01-01

    Boiling (BWR) and Pressurized (PWR) Water Reactor Power Plants are being simulated at BNL with the AD10 and AD100 Peripheral Processor Systems. The AD10 system has been used for BWR simulations since 1984 for safety analyses, emergency training and optimization studies. BWR simulation capabilities have been implemented recently on the AD100 system and PWR simulation capabilities are currently being developed under the auspices of international cooperation. Modeling and simulation methods are presented with emphasis on the simulation of the Nuclear Steam Supply System. Results are presented for BWR simulation and performance characteristics are compared of the AD10 and AD100 systems. It will be shown that the AD100 simulates two times faster than two AD10 processors operating in parallel and that the computing capacity of one AD100 (with FMU processor) is twice as large as that of two AD10 processors. 9 refs., 5 figs., 1 tab.

  5. The effects of neutral gas heating on H mode transition and maintenance currents in a 13.56 MHz planar coil inductively coupled plasma reactor

    SciTech Connect

    Jayapalan, Kanesh K.; Chin, Oi-Hoong

    2012-09-15

    The H mode transition and maintenance currents in a 13.56 MHz laboratory 6 turn planar coil inductively coupled plasma (ICP) reactor are simulated for low pressure argon discharge range of 0.02-0.3 mbar with neutral gas heating and at ambient temperature. An experimentally fitted 3D power evolution plot for 0.02 mbar argon pressure is also shown to visualize the effects of hysteresis in the system. Comparisons between simulation and experimental measurements show good agreement in the pressure range of 0.02-0.3 mbar for transition currents and 0.02-0.1 mbar for maintenance currents only when neutral gas heating is considered. This suggests that neutral gas heating plays a non-negligible role in determining the mode transition points of a rf ICP system.

  6. The effects of neutral gas heating on H mode transition and maintenance currents in a 13.56 MHz planar coil inductively coupled plasma reactor

    NASA Astrophysics Data System (ADS)

    Jayapalan, Kanesh K.; Chin, Oi-Hoong

    2012-09-01

    The H mode transition and maintenance currents in a 13.56 MHz laboratory 6 turn planar coil inductively coupled plasma (ICP) reactor are simulated for low pressure argon discharge range of 0.02-0.3 mbar with neutral gas heating and at ambient temperature. An experimentally fitted 3D power evolution plot for 0.02 mbar argon pressure is also shown to visualize the effects of hysteresis in the system. Comparisons between simulation and experimental measurements show good agreement in the pressure range of 0.02-0.3 mbar for transition currents and 0.02-0.1 mbar for maintenance currents only when neutral gas heating is considered. This suggests that neutral gas heating plays a non-negligible role in determining the mode transition points of a rf ICP system.

  7. Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit

    SciTech Connect

    Merzari, E.; Shemon, E. R.; Yu, Y. Q.; Thomas, J. W.; Obabko, A.; Jain, Rajeev; Mahadevan, Vijay; Tautges, Timothy; Solberg, Jerome; Ferencz, Robert Mark; Whitesides, R.

    2015-12-21

    This report describes to employ SHARP to perform a first-of-a-kind analysis of the core radial expansion phenomenon in an SFR. This effort required significant advances in the framework Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit used to drive the coupled simulations, manipulate the mesh in response to the deformation of the geometry, and generate the necessary modified mesh files. Furthermore, the model geometry is fairly complex, and consistent mesh generation for the three physics modules required significant effort. Fully-integrated simulations of a 7-assembly mini-core test problem have been performed, and the results are presented here. Physics models of a full-core model of the Advanced Burner Test Reactor have also been developed for each of the three physics modules. Standalone results of each of the three physics modules for the ABTR are presented here, which provides a demonstration of the feasibility of the fully-integrated simulation.

  8. Effect Of Gas Mixture Composition On Tar Removal Process In A Pulsed Corona Discharge Reactor

    NASA Astrophysics Data System (ADS)

    Filimonova E.; Naidis, G.

    2010-07-01

    The simulation of naphthalene (C10H8) removal from several gas mixtures (pure nitrogen, mixtures containing N2 with CO2, CO, H2, H2O, and biogas - the product of biomass gasification), has been investigated. The modeling is based on the experimental data obtained in the reactor with a pulsed positive corona discharge. The problem of simulation of the cleaning process includes description of two stages. The first, fast stage is generation of primary active species during streamer propagation. The second, slow stage is the chain of chemical transformations triggered by these species. The input parameters for the modeling of the second stage are G-values for generation of primary active species, obtained under consideration of streamer dynamics. Simulation of the second stage of the removal process takes into account the processes of chemical kinetics and diffusion outside and inside of streamer traces during multi-pulsed treatment. Besides neutral active species, streamer discharges produce electrons and ions. Primary positive ions (N2+, CO+, CO2+, H2+, H2O+) in a chain of fast ion-molecule reactions transform into more stable positive ions. The ions recombine with electrons. Both ion-molecule reactions and electron-ion recombination process are additional (to dissociation of gas molecules by electron impact in the streamer head) sources of neutral active species. The relative contribution of these sources to the G-values for H, OH and O is rather large. In our modeling two approaches have been used. At the first approach the contribution of ion-molecule reactions is estimated approximately assuming that the dominating stable ion is N4+ (in pure N2 and its mixtures with H2) or CO2+ (in mixtures including CO2). Other way is the calculations with kinetic scheme including the molecular ions, aquated ions such as H3O(H2O)m+, NO2(H2O)-, NO2(H2O)+ and other. The comparison of results of two approaches is presented. Only full kinetic scheme allowed describing the

  9. Physics-based multiscale coupling for full core nuclear reactor simulation

    SciTech Connect

    Gaston, Derek R.; Permann, Cody J.; Peterson, John W.; Slaughter, Andrew E.; Andrš, David; Wang, Yaqi; Short, Michael P.; Perez, Danielle M.; Tonks, Michael R.; Ortensi, Javier; Zou, Ling; Martineau, Richard C.

    2015-10-01

    Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different data exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle. 2014 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-SA license

  10. Physics-based multiscale coupling for full core nuclear reactor simulation

    DOE PAGESBeta

    Gaston, Derek R.; Permann, Cody J.; Peterson, John W.; Slaughter, Andrew E.; Andrš, David; Wang, Yaqi; Short, Michael P.; Perez, Danielle M.; Tonks, Michael R.; Ortensi, Javier; et al

    2015-10-01

    Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different datamore » exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle. 2014 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-SA license« less

  11. Robust Low-Cost Water-Gas Shift Membrane Reactor for High-Purity Hydrogen Production form Coal-Derived Syngas

    SciTech Connect

    James Torkelson; Neng Ye; Zhijiang Li; Decio Coutinho; Mark Fokema

    2008-05-31

    This report details work performed in an effort to develop a low-cost, robust water gas shift membrane reactor to convert coal-derived syngas into high purity hydrogen. A sulfur- and halide-tolerant water gas shift catalyst and a sulfur-tolerant dense metallic hydrogen-permeable membrane were developed. The materials were integrated into a water gas shift membrane reactor in order to demonstrate the production of >99.97% pure hydrogen from a simulated coal-derived syngas stream containing 2000 ppm hydrogen sulfide. The objectives of the program were to (1) develop a contaminant-tolerant water gas shift catalyst that is able to achieve equilibrium carbon monoxide conversion at high space velocity and low steam to carbon monoxide ratio, (2) develop a contaminant-tolerant hydrogen-permeable membrane with a higher permeability than palladium, (3) demonstrate 1 L/h purified hydrogen production from coal-derived syngas in an integrated catalytic membrane reactor, and (4) conduct a cost analysis of the developed technology.

  12. Reactor2D: A tool for simulation of shock deformation

    NASA Astrophysics Data System (ADS)

    Kraus, Eugeny I.; Shabalin, Ivan I.

    2016-10-01

    The basic steps for creating a numerical tool to simulate the deformation and failure processes of complex technical objects (CTO) are presented. Calculations of shock loading of CTO both at low and high speeds, showing the efficiency of the numerical tools created are carried out.

  13. Demonstration of landfill gas enhancement techniques in landfill simulators

    NASA Astrophysics Data System (ADS)

    Walsh, J. J.; Vogt, W. G.

    1982-02-01

    Various techniques to enhance gas production in sanitary landfills were applied to landfill simulators. These techniques include (1) accelerated moisture addition, (2) leachate recycling, (3) buffer addition, (4) nutrient addition, and (5) combinations of the above. Results are compiled through on-going operation and monitoring of sixteen landfill simulators. These test cells contain about 380 kg of municipal solid waste. Quantities of buffer and nutrient materials were placed in selected cells at the time of loading. Water is added to all test cells on a monthly basis; leachate is withdrawn from all cells (and recycled on selected cells) also on a monthly basis. Daily monitoring of gas volumes and refuse temperatures is performed. Gas and leachate samples are collected and analyzed on a monthly basis. Leachate and gas quality and quantity reslts are presented for the first 18 months of operation.

  14. Simulation of gas particle flow in a HVOF torch

    SciTech Connect

    Chang, C.H.; Moore, R.L.

    1995-12-31

    A transient two-dimensional numerical simulation of Inconel spraying in an HVOF torch barrel has been performed. The gas flow is treated as a continuum multicomponent chemically reacting flow, while particles are modeled using a stochastic particle spray model, fully coupled to the gas flow. The calculated results agree well with experimental data, and show important statistical aspects of particle flow in the torch.

  15. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  16. Atmospheric pressure plasma chemical vapor deposition reactor for 100 mm wafers, optimized for minimum contamination at low gas flow rates

    NASA Astrophysics Data System (ADS)

    Anand, Venu; Nair, Aswathi R.; Shivashankar, S. A.; Mohan Rao, G.

    2015-08-01

    Gas discharge plasmas used for thinfilm deposition by plasma-enhanced chemical vapor deposition (PECVD) must be devoid of contaminants, like dust or active species which disturb the intended chemical reaction. In atmospheric pressure plasma systems employing an inert gas, the main source of such contamination is the residual air inside the system. To enable the construction of an atmospheric pressure plasma (APP) system with minimal contamination, we have carried out fluid dynamic simulation of the APP chamber into which an inert gas is injected at different mass flow rates. On the basis of the simulation results, we have designed and built a simple, scaled APP system, which is capable of holding a 100 mm substrate wafer, so that the presence of air (contamination) in the APP chamber is minimized with as low a flow rate of argon as possible. This is examined systematically by examining optical emission from the plasma as a function of inert gas flow rate. It is found that optical emission from the plasma shows the presence of atmospheric air, if the inlet argon flow rate is lowered below 300 sccm. That there is minimal contamination of the APP reactor built here, was verified by conducting an atmospheric pressure PECVD process under acetylene flow, combined with argon flow at 100 sccm and 500 sccm. The deposition of a polymer coating is confirmed by infrared spectroscopy. X-ray photoelectron spectroscopy shows that the polymer coating contains only 5% of oxygen, which is comparable to the oxygen content in polymer deposits obtained in low-pressure PECVD systems.

  17. Real-time dynamic simulator for the Topaz II reactor power system

    SciTech Connect

    Kwok, K.S.

    1994-10-01

    A dynamic simulator of the TOPAZ II reactor system has been developed for the Nuclear Electric Propulsion Space Test Program. The simulator is a self-contained IBM-PC compatible based system that executes at a speed faster than real-time. The simulator combines first-principle modeling and empirical correlations in its algorithm to attain the modeling accuracy and computational through-put that are required for real-time execution. The overall execution time of the simulator for each time step is 15 ms when no data is written to the disk, and 18 ms when nine double precision data points are written to the disk once in every time step. The simulation program has been tested and it is able to handle a step decrease of $8 worth of reactivity. It also provides simulation of fuel, emitter, collector, stainless steel, and ZrH moderator failures. Presented in this paper are the models used in the calculations, a sample simulation session, and a discussion of the performance and limitations of the simulator. The simulator has been found to provide realistic real-time dynamic response of the TOPAZ II reactor system under both normal and causality conditions.

  18. Detailed Reaction Kinetics for CFD Modeling of Nuclear Fuel Pellet Coating for High Temperature Gas-Cooled Reactors

    SciTech Connect

    Battaglia, Francine

    2008-11-29

    The research project was related to the Advanced Fuel Cycle Initiative and was in direct alignment with advancing knowledge in the area of Nuclear Fuel Development related to the use of TRISO fuels for high-temperature reactors. The importance of properly coating nuclear fuel pellets received a renewed interest for the safe production of nuclear power to help meet the energy requirements of the United States. High-temperature gas-cooled nuclear reactors use fuel in the form of coated uranium particles, and it is the coating process that was of importance to this project. The coating process requires four coating layers to retain radioactive fission products from escaping into the environment. The first layer consists of porous carbon and serves as a buffer layer to attenuate the fission and accommodate the fuel kernel swelling. The second (inner) layer is of pyrocarbon and provides protection from fission products and supports the third layer, which is silicon carbide. The final (outer) layer is also pyrocarbon and provides a bonding surface and protective barrier for the entire pellet. The coating procedures for the silicon carbide and the outer pyrocarbon layers require knowledge of the detailed kinetics of the reaction processes in the gas phase and at the surfaces where the particles interact with the reactor walls. The intent of this project was to acquire detailed information on the reaction kinetics for the chemical vapor deposition (CVD) of carbon and silicon carbine on uranium fuel pellets, including the location of transition state structures, evaluation of the associated activation energies, and the use of these activation energies in the prediction of reaction rate constants. After the detailed reaction kinetics were determined, the reactions were implemented and tested in a computational fluid dynamics model, MFIX. The intention was to find a reduced mechanism set to reduce the computational time for a simulation, while still providing accurate results

  19. HYBRID SULFUR CYCLE FLOWSHEETS FOR HYDROGEN PRODUCTION USING HIGH-TEMPERATURE GAS-COOLED REACTORS

    SciTech Connect

    Gorensek, M.

    2011-07-06

    Two hybrid sulfur (HyS) cycle process flowsheets intended for use with high-temperature gas-cooled reactors (HTGRs) are presented. The flowsheets were developed for the Next Generation Nuclear Plant (NGNP) program, and couple a proton exchange membrane (PEM) electrolyzer for the SO2-depolarized electrolysis step with a silicon carbide bayonet reactor for the high-temperature decomposition step. One presumes an HTGR reactor outlet temperature (ROT) of 950 C, the other 750 C. Performance was improved (over earlier flowsheets) by assuming that use of a more acid-tolerant PEM, like acid-doped poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI), instead of Nafion{reg_sign}, would allow higher anolyte acid concentrations. Lower ROT was accommodated by adding a direct contact exchange/quench column upstream from the bayonet reactor and dropping the decomposition pressure. Aspen Plus was used to develop material and energy balances. A net thermal efficiency of 44.0% to 47.6%, higher heating value basis is projected for the 950 C case, dropping to 39.9% for the 750 C case.

  20. 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.

  1. Continuous biological waste gas treatment in stirred trickle-bed reactor with discontinuous removal of biomass

    SciTech Connect

    Laurenzis, A.; Heits, H.; Wuebker, S.M.; Heinze, U.; Friedrich, C.; Werner, U.

    1998-02-20

    A new reactor for biological waste gas treatment was developed to eliminate continuous solvents from waste gases. A trickle-bed reactor was chosen with discontinuous movement of the packed bed and intermittent percolation. The reactor was operated with toluene as the solvent and an optimum average biomass concentration of between 5 and 30 kg dry cell weight per cubic meter packed bed (m{sub pb}{sup 3}). This biomass concentration resulted in a high volumetric degradation rate. Reduction of surplus biomass by stirring and trickling caused a prolonged service life and prevented clogging of the trickle bed and a pressure drop increase. The pressure drop after biomass reduction was almost identical to the theoretical pressure drop as calculated for the irregular packed bed without biomass. The reduction in biomass and intermittent percolation of mineral medium resulted in high volumetric degradation rates of about 100 g of toluene m{sub pb}{sup {minus}3} h{sup {minus}1} at a load of 150 g of toluene m{sub pb}{sup {minus}3} h{sup {minus}1}. Such a removal rate with a trickle-bed reactor was not reported before.

  2. 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)

  3. High-performance simulations for atmospheric pressure plasma reactor

    NASA Astrophysics Data System (ADS)

    Chugunov, Svyatoslav

    Plasma-assisted processing and deposition of materials is an important component of modern industrial applications, with plasma reactors sharing 30% to 40% of manufacturing steps in microelectronics production. Development of new flexible electronics increases demands for efficient high-throughput deposition methods and roll-to-roll processing of materials. The current work represents an attempt of practical design and numerical modeling of a plasma enhanced chemical vapor deposition system. The system utilizes plasma at standard pressure and temperature to activate a chemical precursor for protective coatings. A specially designed linear plasma head, that consists of two parallel plates with electrodes placed in the parallel arrangement, is used to resolve clogging issues of currently available commercial plasma heads, as well as to increase the flow-rate of the processed chemicals and to enhance the uniformity of the deposition. A test system is build and discussed in this work. In order to improve operating conditions of the setup and quality of the deposited material, we perform numerical modeling of the plasma system. The theoretical and numerical models presented in this work comprehensively describe plasma generation, recombination, and advection in a channel of arbitrary geometry. Number density of plasma species, their energy content, electric field, and rate parameters are accurately calculated and analyzed in this work. Some interesting engineering outcomes are discussed with a connection to the proposed setup. The numerical model is implemented with the help of high-performance parallel technique and evaluated at a cluster for parallel calculations. A typical performance increase, calculation speed-up, parallel fraction of the code and overall efficiency of the parallel implementation are discussed in details.

  4. Simulation and testing of a vertical organometallic vapor phase epitaxy reactor

    NASA Astrophysics Data System (ADS)

    Sani, R. A.; Barmawi, M.; Mindara, J. Y.

    1998-02-01

    The purpose of the study is to design a single wafer vertical organo-metallic vapor phase epitaxy (OMVPE) reactor which gives a uniform deposition around the symmetry axis. The vertical reactor under the consideration consist of a diffuser and a system of coaxial cylinders to laminarize the flow which may lead to a uniform deposition without rotating the susceptor. The simulation shows that for a susceptor with a radius of 2.5 cm, a uniformity can be achieved in a region of a radius of 2 cm within 1% for certain operating condition. The result is compared with the experimental measurement of TiO2 deposition from TTIP.

  5. Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

    SciTech Connect

    Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.; Mitchell, Mark R.; Gore, Bryan F.; Faris, Drury K.

    2009-10-09

    The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected to come from increasingly diverse educational and experiential backgrounds.

  6. Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

    NASA Astrophysics Data System (ADS)

    Bragg-Sitton, Shannon M.

    The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and

  7. Characteristics and control response of the TOPAZ II Reactor System Real-time Dynamic Simulator

    SciTech Connect

    Kwok, K.S.

    1993-11-12

    A dynamic simulator of the TOPAZ II reactor system has been developed for the Nuclear Electric Propulsion Space Test Program. The simulator combines first-principle modeling and empirical correlations in its algorithm to attain the modeling accuracy and computational through-put that are required for real-time execution. The overall execution time of the simulator for each time step is 15 ms when no data is written to the disk, and 18 ms when nine double precision data points are written to the disk once in every time step. The simulation program has been tested and it is able to handle a step decrease of $8 worth of reactivity. It also provides simulations of fuel, emitter, collector, stainless steel, and ZrH moderator failures. Presented in this paper are the models used in the calculations, a sample simulation session, and a discussion of the performance and limitations of the simulator. The simulator has been found to provide realistic real-time dynamic response of the TOPAZ II reactor system under both normal and casualty conditions.

  8. Flow chemistry: intelligent processing of gas-liquid transformations using a tube-in-tube reactor.

    PubMed

    Brzozowski, Martin; O'Brien, Matthew; Ley, Steven V; Polyzos, Anastasios

    2015-02-17

    reactive gas in a given reaction mixture. We have developed a tube-in-tube reactor device consisting of a pair of concentric capillaries in which pressurized gas permeates through an inner Teflon AF-2400 tube and reacts with dissolved substrate within a liquid phase that flows within a second gas impermeable tube. This Account examines our efforts toward the development of a simple, unified methodology for the processing of gaseous reagents in flow by way of development of a tube-in-tube reactor device and applications to key C-C, C-N, and C-O bond forming and hydrogenation reactions. We further describe the application to multistep reactions using solid-supported reagents and extend the technology to processes utilizing multiple gas reagents. A key feature of our work is the development of computer-aided imaging techniques to allow automated in-line monitoring of gas concentration and stoichiometry in real time. We anticipate that this Account will illustrate the convenience and benefits of membrane tube-in-tube reactor technology to improve and concomitantly broaden the scope of gas/liquid/solid reactions in organic synthesis.

  9. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-01

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  10. A catalytic reactor for the trapping of free radicals from gas phase oxidation reactions

    NASA Astrophysics Data System (ADS)

    Conte, Marco; Wilson, Karen; Chechik, Victor

    2010-10-01

    A catalytic reactor for the trapping of free radicals originating from gas phase catalytic reactions is described and discussed. Radical trapping and identification were initially carried out using a known radical generator such as dicumyl peroxide. The trapping of radicals was further demonstrated by investigating genuine radical oxidation processes, e.g., benzaldehyde oxidation over manganese and cobalt salts. The efficiency of the reactor was finally proven by the partial oxidation of cyclohexane over MoO3, Cr2O3, and WO3, which allowed the identification of all the radical intermediates responsible for the formation of the products cyclohexanol and cyclohexanone. Assignment of the trapped radicals was carried out using spin trapping technique and X-band electron paramagnetic resonance spectroscopy.

  11. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    SciTech Connect

    Mcwilliams, A. J.

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  12. 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.

  13. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    SciTech Connect

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-30

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  14. Development of a coolant channel helium and nitrogen gas ratio sensor for a high temperature gas reactor

    SciTech Connect

    Cadell, S. R.; Woods, B. G.

    2012-07-01

    To measure the changing gas composition of the coolant during a postulated High Temperature Gas Reactor (HTGR) accident, an instrument is needed. This instrument must be compact enough to measure the ratio of the coolant versus the break gas in an individual coolant channel. This instrument must minimally impact the fluid flow and provide for non-direct signal routing to allow minimal disturbance to adjacent channels. The instrument must have a flexible geometry to allow for the measurement of larger volumes such as in the upper or lower plenum of a HTGR. The instrument must be capable of accurately functioning through the full operating temperature and pressure of a HTGR. This instrument is not commercially available, but a literature survey has shown that building off of the present work on Capacitance Sensors and Cross-Capacitors will provide a basis for the development of the desired instrument. One difficulty in developing and instrument to operate at HTGR temperatures is acquiring an electrical conductor that will not melt at 1600 deg. C. This requirement limits the material selection to high temperature ceramics, graphite, and exotic metals. An additional concern for the instrument is properly accounting for the thermal expansion of both the sensing components and the gas being measured. This work covers the basic instrument overview with a thorough discussion of the associated uncertainty in making these measurements. (authors)

  15. Environmental aspects of MHTGR (Modular High-Temperature Gas-Cooled Reactor) operation

    SciTech Connect

    Neylan, A.J.; Dilling, D.A.; Cardito, J.M.

    1988-09-01

    The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept being developed under a cooperative program involving the US Government, the utilities and the nuclear industry. This plant design utilizes basic High Temperature Gas-Cooled Reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The MHTGR design approach leading to exceptional safety performance also leads to plant operation which is characterized by extremely low radiological emissions even for very low probability accidents. Coated fuel particles retain radionuclides within the fuel, thus minimizing material contamination and personnel exposure. The objective of this paper is to characterize radioactive effluents expected from the normal operation of an MHTGR. In addition, other nonradioactive effluents associated with a power generating facility are discussed. Nuclear power plants produce radioactive effluents during normal operation in gaseous, liquid and solid forms. Principal sources of radioactive waste within the MHTGR are identified. The manner in which it is planned to treat these wastes is described. Like other reactors, the MHTGR produces nonradioactive effluents associated with heat generation and chemical usage. However, due to the MHTGR's higher efficiency, water usage requirements and chemical discharges for the MHTGR are minimized relative to other types of nuclear power plants. Based upon prior operating HTGR experience and analysis, effluents are quantified in terms of radioactivity levels and/or emission volume. Results, quantified within the paper, demonstrate that effluents from the MHTGR are well below regulatory limits and that the MHTGR has a minimal impact upon the public and the environment. 14 refs., 2 figs., 4 tabs.

  16. The effect of carrier gas flow rate and source cell temperature on low pressure organic vapor phase deposition simulation by direct simulation Monte Carlo method

    NASA Astrophysics Data System (ADS)

    Wada, Takao; Ueda, Noriaki

    2013-04-01

    The process of low pressure organic vapor phase deposition (LP-OVPD) controls the growth of amorphous organic thin films, where the source gases (Alq3 molecule, etc.) are introduced into a hot wall reactor via an injection barrel using an inert carrier gas (N2 molecule). It is possible to control well the following substrate properties such as dopant concentration, deposition rate, and thickness uniformity of the thin film. In this paper, we present LP-OVPD simulation results using direct simulation Monte Carlo-Neutrals (Particle-PLUS neutral module) which is commercial software adopting direct simulation Monte Carlo method. By estimating properly the evaporation rate with experimental vaporization enthalpies, the calculated deposition rates on the substrate agree well with the experimental results that depend on carrier gas flow rate and source cell temperature.

  17. Gas loss in simulated galaxies as they fall into clusters.

    PubMed

    Cen, Renyue; Pop, Ana Roxana; Bahcall, Neta A

    2014-06-01

    We use high-resolution cosmological hydrodynamic galaxy formation simulations to gain insights into how galaxies lose their cold gas at low redshift as they migrate from the field to the high-density regions of clusters of galaxies. We find that beyond three cluster virial radii, the fraction of gas-rich galaxies is constant, representing the field. Within three cluster-centric radii, the fraction of gas-rich galaxies declines steadily with decreasing radius, reaching <10% near the cluster center. Our results suggest galaxies start to feel the effect of the cluster environment on their gas content well beyond the cluster virial radius. We show that almost all gas-rich galaxies at the cluster virial radius are falling in for the first time at nearly radial orbits. Furthermore, we find that almost no galaxy moving outward at the cluster virial radius is gas-rich (with a gas-to-baryon ratio greater than 1%). These results suggest that galaxies that fall into clusters lose their cold gas within a single radial round-trip.

  18. Gas loss in simulated galaxies as they fall into clusters.

    PubMed

    Cen, Renyue; Pop, Ana Roxana; Bahcall, Neta A

    2014-06-01

    We use high-resolution cosmological hydrodynamic galaxy formation simulations to gain insights into how galaxies lose their cold gas at low redshift as they migrate from the field to the high-density regions of clusters of galaxies. We find that beyond three cluster virial radii, the fraction of gas-rich galaxies is constant, representing the field. Within three cluster-centric radii, the fraction of gas-rich galaxies declines steadily with decreasing radius, reaching <10% near the cluster center. Our results suggest galaxies start to feel the effect of the cluster environment on their gas content well beyond the cluster virial radius. We show that almost all gas-rich galaxies at the cluster virial radius are falling in for the first time at nearly radial orbits. Furthermore, we find that almost no galaxy moving outward at the cluster virial radius is gas-rich (with a gas-to-baryon ratio greater than 1%). These results suggest that galaxies that fall into clusters lose their cold gas within a single radial round-trip. PMID:24843167

  19. Gas loss in simulated galaxies as they fall into clusters

    PubMed Central

    Cen, Renyue; Pop, Ana Roxana; Bahcall, Neta A.

    2014-01-01

    We use high-resolution cosmological hydrodynamic galaxy formation simulations to gain insights into how galaxies lose their cold gas at low redshift as they migrate from the field to the high-density regions of clusters of galaxies. We find that beyond three cluster virial radii, the fraction of gas-rich galaxies is constant, representing the field. Within three cluster-centric radii, the fraction of gas-rich galaxies declines steadily with decreasing radius, reaching <10% near the cluster center. Our results suggest galaxies start to feel the effect of the cluster environment on their gas content well beyond the cluster virial radius. We show that almost all gas-rich galaxies at the cluster virial radius are falling in for the first time at nearly radial orbits. Furthermore, we find that almost no galaxy moving outward at the cluster virial radius is gas-rich (with a gas-to-baryon ratio greater than 1%). These results suggest that galaxies that fall into clusters lose their cold gas within a single radial round-trip. PMID:24843167

  20. 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.

  1. Extension of the CENTAR system simulation code to thermionic space nuclear reactors

    NASA Astrophysics Data System (ADS)

    Nassersharif, Bahram; Gaeta, Michael J.; Berge, Francoise; Guffee, Laura; Williams, Ken

    The Code for Extended Nonlinear Transient Analysis of Extraterrestial Reactors (CENTAR) is a general-purpose reactor system simulation code capable of modeling coupled heat transfer, fluid flow, neutronic, and control in an arbitrary reactor system (or subsystem) configuration. CENTAR 4.0 has been enhanced to support thermionic solid-core systems as well as liquid-metal systems. Several new models have been added. The fuel model has been enhanced to support two-dimensional heat transfer with multiple material gap interfaces. Gaps may contain thermionic converters. A quasi-steady-state heat-pipe component model has been added to predict operating limits. The thermoelectric/electromagnetic (TEM) pump model in CENTAR has been extended to also model electromagnetic pumps. A thermionic energy conversion model has been added. Numerous other enhancements to code architecture, user interface, and I/O have also been added in version 4.0.

  2. Numerical simulation of vortex pyrolysis reactors for condensable tar production from biomass

    SciTech Connect

    Miller, R.S.; Bellan, J.

    1998-08-01

    A numerical study is performed in order to evaluate the performance and optimal operating conditions of vortex pyrolysis reactors used for condensable tar production from biomass. A detailed mathematical model of porous biomass particle pyrolysis is coupled with a compressible Reynolds stress transport model for the turbulent reactor swirling flow. An initial evaluation of particle dimensionality effects is made through comparisons of single- (1D) and multi-dimensional particle simulations and reveals that the 1D particle model results in conservative estimates for total pyrolysis conversion times and tar collection. The observed deviations are due predominantly to geometry effects while directional effects from thermal conductivity and permeability variations are relatively small. Rapid ablative particle heating rates are attributed to a mechanical fragmentation of the biomass particles that is modeled using a critical porosity for matrix breakup. Optimal thermal conditions for tar production are observed for 900 K. Effects of biomass identity, particle size distribution, and reactor geometry and scale are discussed.

  3. New approach to creation of geometrical module for nuclear reactor neutron transport computer simulation analysis

    SciTech Connect

    Poveschenko, T.; Poveschenko, O.

    2012-07-01

    This paper presents the new approach to creation of geometrical module for nuclear reactor neutron transport computer simulation analysis so called the differential cross method. It is elaborated for detecting boards between physical zones. It is proposed to use GMSH open source mesh editor extended by some features: a special option and a special kind of mesh (cubic background mesh).This method is aimed into Monte Carlo Method as well as for deterministic neutron transport methods. Special attention is attended for reactor core composed of a set of material zones with complicate geometrical boundaries. The idea of this approach is described. In general case method works for 3-D space. Algorithm of creation of the geometrical module is given. 2-D neutron transport benchmark-test for RBMK reactor cluster cell is described. It demonstrates the ability of this approach to provide flexible definition of geometrical meshing with preservation of curved surface or any level of heterogeneity. (authors)

  4. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    SciTech Connect

    Raussi, P.; Kainulainen, S.; Kouhia, J.

    1995-09-01

    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

  5. Design of small gas cooled fast reactor with two region of natural Uranium fuel fraction

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Monado, Fiber; Sekimoto, Hiroshi; Nakayama, Sinsuke

    2012-06-01

    A design study of small Gas Cooled Fast Reactor with two region fuel has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region fuel i.e. 60% fuel fraction of Natural Uranium as inner core and 65% fuel fraction of Natural Uranium as outer core. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 filled by fresh Natural Uranium. This concept is basically applied to all regions in both cores area, i.e. shifted the core of ith region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium on each region-1. The burn-up calculation is performed using collision probability method PIJ (cell burn-up calculation) in SRAC code which then given eight energy group macroscopic cross section data to be used in two dimensional R-Z geometry multi groups diffusion calculation in CITATION code. This reactor can results power thermal 600 MWth with average power density i.e. 80 watt/cc. After reactor start-up the operation, furthermore reactor only needs Natural Uranium supply for continue operation along 100 years. This calculation result then compared with one region fuel design i.e. 60% and 65% fuel fraction. This core design with two region fuel fraction can be an option for fuel optimization.

  6. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    SciTech Connect

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

  7. Pockels-effect cell for gas-flow simulation

    NASA Technical Reports Server (NTRS)

    Weimer, D.

    1982-01-01

    A Pockels effect cell using a 75 cu cm DK*P crystal was developed and used as a gas flow simulator. Index of refraction gradients were produced in the cell by the fringing fields of parallel plate electrodes. Calibration curves for the device were obtained for index of refraction gradients in excess of .00025 m.

  8. Fueling galaxy growth through gas accretion in cosmological simulations

    NASA Astrophysics Data System (ADS)

    Nelson, Dylan Rubaloff

    Despite significant advances in the numerical modeling of galaxy formation and evolution, it is clear that a satisfactory theoretical picture of how galaxies acquire their baryons across cosmic time remains elusive. In this thesis we present a computational study which seeks to address the question of how galaxies get their gas. We make use of new, more robust simulation techniques and describe the first investigations of cosmological gas accretion using a moving-mesh approach for solving the equations of continuum hydrodynamics. We focus first on a re-examination of past theoretical conclusions as to the relative importance of different accretion modes for galaxy growth. We study the rates and nature of gas accretion at z=2, comparing our new simulations run with the Arepo code to otherwise identical realizations run with the smoothed particle hydrodynamics code Gadget. We find significant physical differences in the thermodynamic history of accreted gas, explained in terms of numerical inaccuracies in SPH. In contrast to previous results, we conclude that hot mode accretion generally dominates galaxy growth, while cold gas filaments experience increased heating and disruption. Next, we consider the impact of feedback on our results, including models for galactic-scale outflows driven by stars as well as the energy released from supermassive black holes. We find that feedback strongly suppresses the inflow of "smooth" mode gas at all redshifts, regardless of its temperature history. Although the geometry of accretion at the virial radius is largely unmodified, strong galactic-fountain recycling motions dominate the inner halo. We measure a shift in the characteristic timescale of accretion, and discuss implications for semi-analytical models of hot halo gas cooling. To overcome the resolution limitations of cosmological volumes, we simulate a suite of eight individual 1012 solar mass halos down to z=2. We quantify the thermal and dynamical structure of the gas in

  9. Implications of Results from the Advanced Gas Reactor Fuel Development and Qualification Program on Licensing of Modular HTGRs

    SciTech Connect

    David Petti

    2001-10-01

    The high level of safety of modular high temperature gas-cooled reactor (HTGR) designs is achieved by passively maintaining core temperatures below fission product release thresholds under all envisioned accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to other reactor types but is predicated on exceptionally high coated-particle fuel fabrication quality and excellent fuel performance under normal operation and accident conditions. The Advanced Gas Reactor Fuel Development and Qualification (AGR) Program decided to qualify for uranium oxide/uranium carbide (UCO) TRISO coated-particle fuel in an operating envelope that would bound both pebble bed and prismatic modular HTGR options. By using a mixture of uranium oxide and uranium carbide, the kernel composition is engineered to minimize CO formation and fuel kernel migration, which is key to maintain to fuel integrity at the higher burnups, temperatures, and temperature gradients anticipated in prismatic HTGRs. Fuel fabrication conducted at both laboratory and engineering scale has demonstrated the ability to fabricate high quality UCO TRISO fuel with very low defects. The first irradiation (AGR 1) exposed about 300,000 TRISO fuel particles to a peak burnup of 19.6% FIMA, a peak fast-neutron fluence of about 4.3 × 1025 n/m2, and a maximum time-averaged fuel temperature of about 1,200°C without a single particle failure. The very low release of key metallic fission products (except silver) measured in post-irradiation examination (PIE) confirms the excellent performance measured under irradiation. Very low releases have been measured in accident simulation heatup testing (''safety testing'') after hundreds of hours at 1600 and 1700°C and no particle failures (no noble gas release measured) have been observed. Even after hundreds of hours at 1800°C, the releases are still very low

  10. 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;

  11. Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

    SciTech Connect

    Behafarid, F.; Shaver, D. R.; Bolotnov, I. A.; Jansen, K. E.; Antal, S. P.; Podowski, M. Z.

    2012-07-01

    The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

  12. Toward the improved simulation of microscale gas flow

    NASA Astrophysics Data System (ADS)

    McNenly, Matthew James

    2007-12-01

    Recent interest in fluidic micro-electro-mechanical systems (MEMS) in gaseous environments has increased the need for accurate simulation techniques to aid in their design process. Many fluidic MEMS operate in a low-speed non-equilibrium gas flow regime that is challenging to simulate both accurately and efficiently. Classic computational fluid dynamics techniques (e.g. Navier-Stokes simulation) are based on the assumption that the fluid behaves as a continuum. This assumption, however, becomes increasingly less accurate as the local flow conditions deviate further from local thermodynamic equilibrium. Alternatively, it is possible to achieve an accurate approximation of non-equilibrium gas flows using particle-based methods (e.g. DSMC), but the resulting simulations are much more computationally expensive than the continuum-based method. In fact, for the very low speeds commonly found in fluidic MEMS, the slow convergence of the DSMC solution can lead to intractably long computation times on all but the largest supercomputers. Two different approaches are pursued in this investigation, in an effort to design a physically accurate and computationally efficient simulation of low-speed, non-equilibrium flows. The first approach constructs new empirical models to correct the error in the Navier-Stokes simulation in the transition regime due to the appreciable deviation from local thermodynamic equilibrium. The empirically corrected Navier-Stokes simulation is not actually predicting strongly non-equilibrium gas flows; however, it is shown to be a useful analysis tool in certain design situations. The second more novel approach develops an original quasi-Monte Carlo (QMC) particle simulation that retains the physical accuracy of the DSMC method while at the same time achieving a faster (near-linear) convergence rate. The design of a QMC method is far more complex in general than a Monte Carlo method for the same problem. Further, no known QMC particle simulation has

  13. 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.

  14. DynMo: Dynamic Simulation Model for Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed; Tournier, Jean-Michel

    2005-02-01

    A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK® platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

  15. DynMo: Dynamic Simulation Model for Space Reactor Power Systems

    SciTech Connect

    El-Genk, Mohamed; Tournier, Jean-Michel

    2005-02-06

    A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK registered platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

  16. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    SciTech Connect

    Durbin, Samuel; Lindgren, Eric R.

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  17. 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.

  18. 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.

  19. MHTGR (Modular High-Temperature Gas-Cooled Reactor) technology development plan

    SciTech Connect

    Homan, F.J.; Neylan, A.J.

    1988-01-01

    This paper presents the approach used to define the technology program needed to support design and licensing of a Modular High-Temperature Gas-Cooled Reactor (MHTGR). The MHTGR design depends heavily on data and information developed during the past 25 years to support large HTGR (LHTGR) designs. The technology program focuses on MHTGR-specific operating and accident conditions, and on validation of models and assumptions developed using LHTGR data. The technology program is briefly outlined, and a schedule is presented for completion of technology work which is consistent with completion of a Final Safety Summary Analysis Report (FSSAR) by 1992.

  20. 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.

  1. 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.

  2. Hypothetical air ingress scenarios in advanced modular high temperature gas cooled reactors

    SciTech Connect

    Kroeger, P.G.

    1988-01-01

    Considering an extremely hypothetical scenario of complete cross duct failure and unlimited air supply into the reactor vessel of a modular high temperature gas cooled ractor, it is found that the potential air inflow remains limited due to the high friction pressure drop through the active core. All incoming air will be oxidized to CO and some local external burning would be temporarily possible in such a scenario. The accident would have to continue with unlimited air supply for hundreds of hours before the core structural integrity would be jeopardized.

  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. NUMERICAL SIMULATION FOR MECHANICAL BEHAVIOR OF U10MO MONOLITHIC MINIPLATES FOR RESEARCH AND TEST REACTORS

    SciTech Connect

    Hakan Ozaltun & Herman Shen

    2011-11-01

    This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stresses (2) thermal cycling of fabricated plates; and finally (3) transient mechanical behavior under actual operating conditions. Because the temperatures approach the melting temperature of the cladding during the fabrication and thermal cycling, high temperature material properties were incorporated to improve the accuracy. Once residual stress fields due to fabrication process were identified, solution was used as initial state for the subsequent simulations. For thermal cycling simulation, elasto-plastic material model with thermal creep was constructed and residual stresses caused by the fabrication process were included. For in-service simulation, coupled fluid-thermal-structural interaction was considered. First, temperature field on the plates was calculated and this field was used to compute the thermal stresses. For time dependent mechanical behavior, thermal creep of cladding, volumetric swelling and fission induced creep of the fuel foil were considered. The analysis showed that the stresses evolve very rapidly in the reactor. While swelling of the foil increases the stress of the foil, irradiation induced creep causes stress relaxation.

  5. 3D Simulation of Missing Pellet Surface Defects in Light Water Reactor Fuel Rods

    SciTech Connect

    B.W. Spencer; J.D. Hales; S.R. Novascone; R.L. Williamson

    2012-09-01

    The cladding on light water reactor (LWR) fuel rods provides a stable enclosure for fuel pellets and serves as a first barrier against fission product release. Consequently, it is important to design fuel to prevent cladding failure due to mechanical interactions with fuel pellets. Cladding stresses can be effectively limited by controlling power increase rates. However, it has been shown that local geometric irregularities caused by manufacturing defects known as missing pellet surfaces (MPS) in fuel pellets can lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. Nuclear fuel performance codes commonly use a 1.5D (axisymmetric, axially-stacked, one-dimensional radial) or 2D axisymmetric representation of the fuel rod. To study the effects of MPS defects, results from 1.5D or 2D fuel performance analyses are typically mapped to thermo-mechanical models that consist of a 2D plane-strain slice or a full 3D representation of the geometry of the pellet and clad in the region of the defect. The BISON fuel performance code developed at Idaho National Laboratory employs either a 2D axisymmetric or 3D representation of the full fuel rod. This allows for a computational model of the full fuel rod to include local defects. A 3D thermo-mechanical model is used to simulate the global fuel rod behavior, and includes effects on the thermal and mechanical behavior of the fuel due to accumulation of fission products, fission gas production and release, and the effects of fission gas accumulation on thermal conductivity across the fuel-clad gap. Local defects can be modeled simply by including them in the 3D fuel rod model, without the need for mapping between two separate models. This allows for the complete set of physics used in a fuel performance analysis to be included naturally in the computational representation of the local defect, and for the effects of the

  6. Simulation of ion beam scattering in a gas stripper

    NASA Astrophysics Data System (ADS)

    Maxeiner, Sascha; Suter, Martin; Christl, Marcus; Synal, Hans-Arno

    2015-10-01

    Ion beam scattering in the gas stripper of an accelerator mass spectrometer (AMS) enlarges the beam phase space and broadens its energy distribution. As the size of the injected beam depends on the acceleration voltage through phase space compression, the stripper becomes a limiting factor of the overall system transmission especially for low energy AMS system in the sub MV region. The spatial beam broadening and collisions with the accelerator tube walls are a possible source for machine background and energy loss fluctuations influence the mass resolution and thus isotope separation. To investigate the physical processes responsible for these effects, a computer simulation approach was chosen. Monte Carlo simulation methods are applied to simulate elastic two body scattering processes in screened Coulomb potentials in a (gas) stripper and formulas are derived to correctly determine random collision parameters and free path lengths for arbitrary (and non-homogeneous) gas densities. A simple parametric form for the underlying scattering cross sections is discussed which features important scaling behaviors. An implementation of the simulation was able to correctly model the data gained with the TANDY AMS system at ETH Zurich. The experiment covered transmission measurements of uranium ions in helium and beam profile measurements after the ion beam passed through the He-stripper. Beam profiles measured up to very high stripper densities could be understood in full system simulations including the relevant ion optics. The presented model therefore simulates the fundamental physics of the interaction between an ion beam and a gas stripper reliably. It provides a powerful and flexible tool for optimizing existing AMS stripper geometries and for designing new, state of the art low energy AMS systems.

  7. Off-gas Adsorption Model and Simulation - OSPREY

    SciTech Connect

    Veronica J Rutledge

    2013-10-01

    The absence of industrial scale nuclear fuel reprocessing in the U.S. has precluded the necessary driver for developing the advanced simulation capability now prevalent in so many other countries. Thus, it is essential to model complex series of unit operations to simulate, understand, and predict inherent transient behavior. A capability of accurately simulating the dynamic behavior of advanced fuel cycle separation processes is expected to provide substantial cost savings and many technical benefits. To support this capability, a modeling effort focused on the off-gas treatment system of a used nuclear fuel recycling facility is in progress. The off-gas separation consists of a series of scrubbers and adsorption beds to capture constituents of interest. Dynamic models are being developed to simulate each unit operation involved so each unit operation can be used as a stand-alone model and in series with multiple others. Currently, an adsorption model has been developed within Multi-physics Object Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). Off-gas Separation and REcoverY (OSPREY) models the adsorption of offgas constituents for dispersed plug flow in a packed bed under non-isothermal and non-isobaric conditions. Inputs to the model include gas composition, sorbent and column properties, equilibrium and kinetic data, and inlet conditions. The simulation outputs component concentrations along the column length as a function of time from which breakthrough data can be obtained. The breakthrough data can be used to determine bed capacity, which in turn can be used to size columns. In addition to concentration data, the model predicts temperature along the column length as a function of time and pressure drop along the column length. A description of the OSPREY model, results from krypton adsorption modeling and plans for modeling the behavior of iodine, xenon, and tritium will be discussed.

  8. Modeling and simulation of an enzymatic reactor for hydrolysis of palm oil.

    PubMed

    Bhatia, S; Naidu, A D; Kamaruddin, A H

    1999-01-01

    Hydrolysis of palm oil has become an important process in Oleochemical industries. Therefore, an investigation was carried out for hydrolysis of palm oil to fatty acid and glycerol using immobilized lipase in packed bed reactor. The conversion vs. residence time data were used in Michaelis-Menten rate equation to evaluate the kinetic parameters. A mathematical model for the rate of palm oil hydrolysis was proposed incorporating role of external mass transfer and pore diffusion. The model was simulated for steady-state isothermal operation of immobilized lipase packed bed reactor. The experimental data were compared with the simulated results. External mass transfer was found to affect the rate of palm oil hydrolysis at higher residence time.

  9. Transient analysis and startup simulation of a thermionic space nuclear reactor system

    SciTech Connect

    El-Genk, M.S.; Xue, Huimin; Paramonov, D. . Dept. of Chemical and Nuclear Engineering)

    1994-01-01

    The thermionic transient analysis model is used to simulate the startup of the TOPAZ-2 space nuclear power system in orbit. The simulated startup procedures are assumed for the purpose of demonstrating the capabilities of the model and may not represent an accurate account of the actual startup procedures of the TOPAZ-2 system. The temperature reactivity feedback effects of the moderator, UO[sub 2] fuel, electrodes, coolant, and other components in the core are calculated, and their effects on the thermal and criticality conditions of the reactor are investigated. Also, estimates of the time constants of the temperature reactivity feedback for the UO[sub 2] fuel and the ZrH moderator during startup, as well as of the total temperature reactivity feedback as a function of the reactor steady-state thermal power, are obtained.

  10. Start-up simulation of a thermionic space nuclear reactor system

    SciTech Connect

    El-Genk, M.S.; Xue, H.; Paramonov, D. )

    1993-01-15

    The Thermionic Transient Analysis Model (TITAM) is used in this paper to simulate the start-up of the TOPAZ-II space nuclear power system in orbit. The start-up procedures simulated herein are assumed for the purpose of demonstrating the capabilities of the model and may not represent an accurate account of the actual start-up procedures of the TOPAZ-II system. The temperature reactivity feedback effects of the moderator, UO[sub 2] fuel, electrodes, coolant, and other components in the core are calculated and their effects on the thermal and criticality conditions of the reactor are investigated. Also, estimates of the time constants of the temperature reactivity feedback for the UO[sub 2] fuel and the ZrH moderator during start-up, as well as of the total temperature reactivity feedback as a function of the reactor steady-state thermal power, are obtained.

  11. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    DOEpatents

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  12. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    DOEpatents

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  13. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    DOEpatents

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2010-02-23

    Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

  14. Design of a Gas Test Loop Facility for the Advanced Test Reactor

    SciTech Connect

    C. A. Wemple

    2005-09-01

    The Office of Nuclear Energy within the U.S. Department of Energy (DOE-NE) has identified the need for irradiation testing of nuclear fuels and materials, primarily in support of the Generation IV (Gen-IV) and Advanced Fuel Cycle Initiative (AFCI) programs. These fuel development programs require a unique environment to test and qualify potential reactor fuel forms. This environment should combine a high fast neutron flux with a hard neutron spectrum and high irradiation temperature. An effort is presently underway at the Idaho National Laboratory (INL) to modify a large flux trap in the Advanced Test Reactor (ATR) to accommodate such a test facility [1,2]. The Gas Test Loop (GTL) Project Conceptual Design was initiated to determine basic feasibility of designing, constructing, and installing in a host irradiation facility, an experimental vehicle that can replicate with reasonable fidelity the fast-flux test environment needed for fuels and materials irradiation testing for advanced reactor concepts. Such a capability will be needed if programs such as the AFCI, Gen-IV, the Next Generation Nuclear Plant (NGNP), and space nuclear propulsion are to meet development objectives and schedules. These programs are beginning some irradiations now, but many call for fast flux testing within this decade.

  15. TEMPEST code simulations of hydrogen distribution in reactor containment structures. Final report

    SciTech Connect

    Trent, D.S.; Eyler, L.L.

    1985-03-01

    The mass transport version of the TEMPEST computer code was used to simulate hydrogen distribution in geometric configurations relevant to reactor containment structures. Predicted results of Battelle-Frankfurt hydrogen distribution tests 1 to 6, and 12 are presented. Agreement between predictions and experimental data is good. Best agreement is obtained using the k-epsilon turbulence model in TEMPEST in flow cases where turbulent diffusion and stable stratification are dominant mechanisms affecting transport. The code's general analysis capabilities are summarized.

  16. Multiscale Simulation of Thermo-mechanical Processes in Irradiated Fission-reactor Materials

    SciTech Connect

    El-Azab, Anter

    2012-05-28

    This report contains a summary of progress made on the subtask area on phase field model development for microstructure evolution in irradiated materials, which was a part of the Computational Materials Science Network (CMSN) project entitled: Multiscale Simulation of Thermo-mechanical Processes in Irradiated Fission-reactor Materials. The model problem chosen has been that of void nucleation and growth under irradiation conditions in single component systems.

  17. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    NASA Astrophysics Data System (ADS)

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  18. 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

  19. Modeling and computational simulation of dilution and biochemical materials balance equations for partially emptied batch reactors.

    PubMed

    Ndon, Udeme J; Zhao, Jin C; Siguan, Marta Larroya

    2007-01-01

    Sequencing batch reactors (SBRs) including aerobic SBRs and anaerobic SBRs (ASBRs) are partially emptied batch reactors that are widely used as bioprocesses in pollution control. We present dilution and biochemical materials balance modeling equations and simulation results for the partially emptied batch reactors, especially for ASBR treatment of low-strength wastewater. The simulated substrate and microbial concentrations for both dilution and materials balance equations follow the same pattern during both feeding and reaction times. However, the results of the materials balance equations show microbial activities during feeding as well as during reaction times and were found to be more appropriate for the biologic system in which substrate removal is associated with microbial growth. Furthermore, the simulation results point to the need to foster high microbial accumulation in the system during startup to optimize the process performance and the need to operate the system at a short reaction time, especially for low substrate concentrations. The results were found to be in agreement with the results of prior laboratory studies.

  20. Specification of requirements for the virtual environment for reactor applications simulation environment

    SciTech Connect

    Hess, S. M.; Pytel, M.

    2012-07-01

    In 2010, the United States Dept. of Energy initiated a research and development effort to develop modern modeling and simulation methods that could utilize high performance computing capabilities to address issues important to nuclear power plant operation, safety and sustainability. To respond to this need, a consortium of national laboratories, academic institutions and industry partners (the Consortium for Advanced Simulation of Light Water Reactors - CASL) was formed to develop an integrated Virtual Environment for Reactor Applications (VERA) modeling and simulation capability. A critical element for the success of the CASL research and development effort was the development of an integrated set of overarching requirements that provides guidance in the planning, development, and management of the VERA modeling and simulation software. These requirements also provide a mechanism from which the needs of a broad array of external CASL stakeholders (e.g. reactor / fuel vendors, plant owner / operators, regulatory personnel, etc.) can be identified and integrated into the VERA development plans. This paper presents an overview of the initial set of requirements contained within the VERA Requirements Document (VRD) that currently is being used to govern development of the VERA software within the CASL program. The complex interdisciplinary nature of these requirements together with a multi-physics coupling approach to realize a core simulator capability pose a challenge to how the VRD should be derived and subsequently revised to accommodate the needs of different stakeholders. Thus, the VRD is viewed as an evolving document that will be updated periodically to reflect the changing needs of identified CASL stakeholders and lessons learned during the progress of the CASL modeling and simulation program. (authors)

  1. Statistical Simulation of Gas Flows through Short Rough Microchannels

    NASA Astrophysics Data System (ADS)

    Markelov, G.; Stefanov, S.

    2011-11-01

    Rarefied gas flow though the rough microchannel is modeled using the direct simulation Monte Carlo method (DSMC). Deterministic and statistical methods of wall roughness deception have been applied. The well-known DSMC-based software, SMILE, allows one to model a flow through the channel with an arbitrary shape, here series of triangular obstacles has been modeled. A statistical approach is based on porous layer model of dusty gas and the model has been implemented in SMILE. Results obtained with two methods are compared and channel performance is presented for different pressure ratio and size of obstacles.

  2. Gas hydrate, fluid flow and free gas: Formation of the bottom-simulating reflector

    NASA Astrophysics Data System (ADS)

    Haacke, R. Ross; Westbrook, Graham K.; Hyndman, Roy D.

    2007-09-01

    Gas hydrate in continental margins is commonly indicated by a prominent bottom-simulating seismic reflector (BSR) that occurs a few hundred metres below the seabed. The BSR marks the boundary between sediments containing gas hydrate above and free gas below. Most of the reflection amplitude is caused by the underlying free gas. Gas hydrate can occur without a BSR, however, and the controls on its formation are not well understood. Here we describe two complementary mechanisms for free gas accumulation beneath the gas hydrate stability zone (GHSZ). The first is the well-recognised hydrate recycling mechanism that generates gas from dissociating hydrate when the base of the GHSZ moves upward relative to hydrate-bearing sediment. The second is a recently identified mechanism in which the relationship between the advection and diffusion of dissolved gas with the local solubility curve allows the liquid phase to become saturated in a thick layer beneath the GHSZ when hydrate is present near its base. This mechanism for gas production (called the solubility-curvature mechanism) is possible in systems where the influence of diffusion becomes important relative to the influence of advection and where the gas-water solubility decreases to a minimum several hundred metres below the GHSZ. We investigate a number of areas in which gas hydrate occurs to determine where gas formation is dominated by the solubility-curvature mechanism and where it is dominated by hydrate recycling. We show that the former is dominant in areas with low rates of upward fluid flow (such as old, rifted continental margins), low rates of seafloor uplift, and high geothermal gradient and/or pressure. Conversely, free-gas formation is dominated by hydrate recycling where there are rapid rates of upward fluid flow and seabed uplift (such as in subduction zone accretionary wedges). Using these two mechanisms to investigate the formation of free gas beneath gas hydrate in continental margins, we are able

  3. 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.

  4. 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)

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

    DOE PAGESBeta

    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

  6. Removal of acetaldehyde and skatole in gas by a corona-discharge reactor

    SciTech Connect

    Sano, Noriaki; Nagamoto, Toshiki; Hamon, Hajime; Suzuki, Tetsuo; Okazaki, Morio

    1997-09-01

    Recently, ultrahigh gas purification has been important in many cases, such as, for example, (1) removal of dioxin from incineration plants, (2) complete removal of radioactive iodine compounds from nuclear fuel recycling, (3) simultaneous removal of NO{sub x} and SO{sub x} in exhaust gases from cogeneration plants, (4) removal or decomposition of chlorofluorocarbons, and (5) supply of purified gas for semiconductor industries. A corona-discharge reactor, called a deposition-type reactor, was applied to remove acetaldehyde and skatole from nitrogen and an oxygen-nitrogen mixture. In the removal from nitrogen, acetaldehyde and skatole are negatively ionized and removed by depositing at the anode surface. In simultaneous removals of acetaldehyde and skatole, it is found that skatole has a higher reactivity of electron attachment than acetaldehyde. In the removal of acetaldehyde from an oxygen-nitrogen mixture, 40 molecules of acetaldehyde were removed by one electron. The reason for the extremely high removal efficiency is considered to be based on the ozone reaction and the formation of negative-ion clusters. Stabilization energies of the negative-ion clusters were estimated by ab initio molecular orbital calculation. Skatole was removed from a nitrogen-oxygen mixture perfectly with extremely low discharge current by the ozone reaction. Simultaneous removals of acetaldehyde and skatole from a nitrogen-oxygen mixture suggest that coexisting skatole inhibits the removal of acetaldehyde.

  7. 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.

  8. 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.

  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. Lunar Regolith Simulant Feed System for a Hydrogen Reduction Reactor System

    NASA Technical Reports Server (NTRS)

    Mueller, R. P.; Townsend, Ivan I., III

    2009-01-01

    One of the goals of In-Situ Resource Utilization (ISRU) on the moon is to produce oxygen from the lunar regolith which is present in the form of Ilmenite (FeTi03) and other compounds. A reliable and attainable method of extracting some of the oxygen from the lunar regolith is to use the hydrogen reduction process in a hot reactor to create water vapor which is then condensed and electrolyzed to obtain oxygen for use as a consumable. One challenge for a production system is to reliably acquire the regolith with an excavator hauler mobility platform and then introduce it into the reactor inlet tube which is raised from the surface and above the reactor itself. After the reaction, the hot regolith (-1000 C) must be expelled from the reactor for disposal by the excavator hauler mobility system. In addition, the reactor regolith inlet and outlet tubes must be sealed by valves during the reaction in order to allow collection of the water vapor by the chemical processing sub-system. These valves must be able to handle abrasive regolith passing through them as well as the heat conduction from the hot reactor. In 2008, NASA has designed and field tested a hydrogen reduction system called ROxygen in order to demonstrate the feasibility of extracting oxygen from lunar regolith. The field test was performed with volcanic ash known as Tephra on Mauna Kea volcano on the Big Island of Hawai'i. The tephra has similar properties to lunar regolith, so that it is regarded as a good simulant for the hydrogen reduction process. This paper will discuss the design, fabrication, operation, test results and lessons learned with the ROxygen regolith feed system as tested on Mauna Kea in November 2008.

  11. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    SciTech Connect

    Slater, C.O.; Reed, D.A.; Cramer, S.N.; Emmett, M.B.; Tomlinson, E.T.

    1981-01-01

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design.

  12. 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.

  13. Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

    SciTech Connect

    Sonat Sen; Gilles Youinou

    2013-02-01

    Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

  14. Beneficial synergetic effect on gas production during co-pyrolysis of sewage sludge and biomass in a vacuum reactor.

    PubMed

    Zhang, Weijiang; Yuan, Chengyong; Xu, Jiao; Yang, Xiao

    2015-05-01

    A vacuum fixed bed reactor was used to pyrolyze sewage sludge, biomass (rice husk) and their blend under high temperature (900°C). Pyrolytic products were kept in the vacuum reactor during the whole pyrolysis process, guaranteeing a long contact time (more than 2h) for their interactions. Remarkable synergetic effect on gas production was observed. Gas yield of blend fuel was evidently higher than that of both parent fuels. The syngas (CO and H2) content and gas lower heating value (LHV) were obviously improved as well. It was highly possible that sewage sludge provided more CO2 and H2O during co-pyrolysis, promoting intense CO2-char and H2O-char gasification, which benefited the increase of gas yield and lower heating value. The beneficial synergetic effect, as a result, made this method a feasible one for gas production.

  15. Beneficial synergetic effect on gas production during co-pyrolysis of sewage sludge and biomass in a vacuum reactor.

    PubMed

    Zhang, Weijiang; Yuan, Chengyong; Xu, Jiao; Yang, Xiao

    2015-05-01

    A vacuum fixed bed reactor was used to pyrolyze sewage sludge, biomass (rice husk) and their blend under high temperature (900°C). Pyrolytic products were kept in the vacuum reactor during the whole pyrolysis process, guaranteeing a long contact time (more than 2h) for their interactions. Remarkable synergetic effect on gas production was observed. Gas yield of blend fuel was evidently higher than that of both parent fuels. The syngas (CO and H2) content and gas lower heating value (LHV) were obviously improved as well. It was highly possible that sewage sludge provided more CO2 and H2O during co-pyrolysis, promoting intense CO2-char and H2O-char gasification, which benefited the increase of gas yield and lower heating value. The beneficial synergetic effect, as a result, made this method a feasible one for gas production. PMID:25728344

  16. Gas temperature measurements inside a hot wall chemical vapor synthesis reactor.

    PubMed

    Notthoff, Christian; Schilling, Carolin; Winterer, Markus

    2012-11-01

    One key but complex parameter in the chemical vapor synthesis (CVS) of nanoparticles is the time temperature profile of the gas phase, which determines particle characteristics such as size (distribution), morphology, microstructure, crystal, and local structure. Relevant for the CVS process and for the corresponding particle characteristics is, however, not the T(t)-profile generated by an external energy source such as a hot wall or microwave reactor but the temperature of the gas carrying reactants and products (particles). Due to a complex feedback of the thermodynamic and chemical processes in the reaction volume with the external energy source, it is very difficult to predict the real gas phase temperature field from the externally applied T(t)-profile. Therefore, a measurement technique capable to determine the temperature distribution of the gas phase under process conditions is needed. In this contribution, we demonstrate with three proof of principle experiments the use of laser induced fluorescence thermometry to investigate the CVS process under realistic conditions.

  17. Simulation for Supporting Scale-Up of a Fluidized Bed Reactor for Advanced Water Oxidation

    PubMed Central

    Abdul Raman, Abdul Aziz; Daud, Wan Mohd Ashri Wan

    2014-01-01

    Simulation of fluidized bed reactor (FBR) was accomplished for treating wastewater using Fenton reaction, which is an advanced oxidation process (AOP). The simulation was performed to determine characteristics of FBR performance, concentration profile of the contaminants, and various prominent hydrodynamic properties (e.g., Reynolds number, velocity, and pressure) in the reactor. Simulation was implemented for 2.8 L working volume using hydrodynamic correlations, continuous equation, and simplified kinetic information for phenols degradation as a model. The simulation shows that, by using Fe3+ and Fe2+ mixtures as catalyst, TOC degradation up to 45% was achieved for contaminant range of 40–90 mg/L within 60 min. The concentration profiles and hydrodynamic characteristics were also generated. A subsequent scale-up study was also conducted using similitude method. The analysis shows that up to 10 L working volume, the models developed are applicable. The study proves that, using appropriate modeling and simulation, data can be predicted for designing and operating FBR for wastewater treatment. PMID:25309949

  18. High-Temperature Gas-Cooled Reactor Technology Development Program: Annual progress report for period ending December 31, 1987

    SciTech Connect

    Jones, J.E.,Jr.; Kasten, P.R.; Rittenhouse, P.L.; Sanders, J.P.

    1989-03-01

    The High-Temperature Gas-Cooled Reactor (HTGR) Program being carried out under the US Department of Energy (DOE) continues to emphasize the development of modular high-temperature gas-cooled reactors (MHTGRs) possessing a high degree of inherent safety. The emphasis at this time is to develop the preliminary design of the reference MHTGR and to develop the associated technology base and licensing infrastructure in support of future reactor deployment. A longer-term objective is to realize the full high-temperature potential of HTGRs in gas turbine and high-temperature, process-heat applications. This document summarizes the activities of the HTGR Technology Development Program for the period ending December 31, 1987.

  19. Simulation Of Plasma Processes--On The Integration Of Feature Simulation With Reactor Scale Models

    SciTech Connect

    Soukane, S.

    2008-09-23

    Advances in both reactor and feature scale modeling have been achieved in plasma processing. Models are being used to address several practical issues in industry. However, accurate predictions of physical phenomena that occur during the process need an integrated multiscale approach that couples reactor scale to feature scale. The implementation of multiscale codes is still a challenge where robustness plays a critical role. This paper addresses the weakest points of such codes especially at the feature scale and provides proper choices for successful implementations. It focuses on the level set methods and fast marching algorithms implemented on quadtree (2D) and octree (3D) data structures and possible extensions of semi-lagrangian methods.

  20. Improving Gas Storage Development Planning Through Simulation-Optimization

    SciTech Connect

    Johnson, V.M.; Ammer, J.; Trick, M.D.

    2000-07-25

    This is the first of two papers describing the application of simulator-optimization methods to a natural gas storage field development planning problem. The results presented here illustrate the large gains in cost-effectiveness that can be made by employing the reservoir simulator as the foundation for a wide-ranging search for solutions to management problems. The current paper illustrates the application of these techniques given a deterministic view of the reservoir. A companion paper will illustrate adaptations needed to accommodate uncertainties regarding reservoir properties.

  1. Numerical simulation of a fully baffled biological reactor: the differential circumferential averaging mixing plane approach.

    PubMed

    Dubey, Hitesh; Das, Sarit Kumar; Panda, Tapobrata

    2006-11-01

    A modified mixing plane approach for steady state simulation of flow field in fully baffled biological reactor is presented and discussed. Without requiring any experimental input, this approach of dividing the vessel into suitable number of connected and disconnected zones; solving steady state equation separately in each zone and then transferring information between them, provides a computationally less intensive alternative for simulating the flow in the whole vessel. Impeller used is the standard Rushton Turbine positioned at mid-height of the reactor and simulations are carried out using standard k-epsilon turbulence model implemented in CFD code FLUENT. Meshing is done using tetrahedral elements such that mesh size gradually increases from the center to the periphery. Most of the previous simulation works present only a few aspects of the flow field with scant importance to the energy balance in the tank and near tip turbulence. In this work, complete model prediction for velocity field and turbulence parameters (near tip and in the bulk region) are validated by comparison with experimental data. As compared to previous simulation works, the results predicted by this "Differential circumferential averaging mixing plane approach" show a better qualitative and quantitative agreement with the published experimental data. A distribution of energy dissipation in various zones of vessel is presented. Also a qualitative picture of flow field and stagnant zones inside the reactor is presented and discussed. Comparison of flow characteristics for different number of baffles shows that for the present dimension of the vessel, five baffles gives maximum enhanced mixing. PMID:16767780

  2. Destruction of chemical agent simulants in a supercritical water oxidation bench-scale reactor.

    PubMed

    Veriansyah, Bambang; Kim, Jae-Duck; Lee, Jong-Chol

    2007-08-17

    A new design of supercritical water oxidation (SCWO) bench-scale reactor has been developed to handle high-risk wastes resulting from munitions demilitarization. The reactor consists of a concentric vertical double wall in which SCWO reaction takes place inside an inner tube (titanium grade 2, non-porous) whereas pressure resistance is ensured by a Hastelloy C-276 external vessel. The performances of this reactor were investigated with two different kinds of chemical warfare agent simulants: OPA (a mixture of isopropyl amine and isopropyl alcohol) as the binary precursor for nerve agent of sarin and thiodiglycol [TDG, (HOC(2)H(4))2S] as the model organic sulfur heteroatom. High destruction rates based on total organic carbon (TOC) were achieved (>99.99%) without production of chars or undesired gases such as carbon monoxide and methane. The carbon-containing product was carbon dioxide whereas the nitrogen-containing products were nitrogen and nitrous oxide. Sulfur was totally recovered in the aqueous effluent as sulfuric acid. No corrosion was noticed in the reactor after a cumulative operation time of more than 250 h. The titanium tube shielded successfully the pressure vessel from corrosion.

  3. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  4. Numerical simulation of the countercurrent flow in a gas centrifuge

    SciTech Connect

    Cloutman, L.D.; Gentry, R.A.

    1981-01-01

    A finite difference method is presented for the numerical simulation of the axisymmetric countercurrent flows in gas centrifuge. A time-marching technique is used to relax an arbitrary initial condition to the desired steady-state solution. All boundary layers may be resolved, and nonlinear effects may be included. Numerical examples are presented. It is concluded that this technique is capable of accurately predicting the performance of a wide variety of machines under all operating conditions of interest.

  5. Off-gas adsorption model and simulation - OSPREY

    SciTech Connect

    Rutledge, V.J.

    2013-07-01

    A capability of accurately simulating the dynamic behavior of advanced fuel cycle separation processes is expected to provide substantial cost savings and many technical benefits. To support this capability, a modeling effort focused on the off-gas treatment system of a used nuclear fuel recycling facility is in progress. The off-gas separation consists of a series of scrubbers and adsorption beds to capture constituents of interest. Dynamic models are being developed to simulate each unit operation involved so each unit operation can be used as a stand-alone model and in series with multiple others. Currently, an adsorption model has been developed within Multi-physics Object Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). Off-gas Separation and Recovery (OSPREY) models the adsorption of offgas constituents for dispersed plug flow in a packed bed under non-isothermal and non-isobaric conditions. Inputs to the model include gas composition, sorbent and column properties, equilibrium and kinetic data, and inlet conditions. The simulation outputs component concentrations along the column length as a function of time from which breakthrough data can be obtained. The breakthrough data can be used to determine bed capacity, which in turn can be used to size columns. In addition to concentration data, the model predicts temperature along the column length as a function of time and pressure drop along the column length. A description of the OSPREY model, results from krypton adsorption modeling and plans for modeling the behavior of iodine, xenon, and tritium will be discussed. (author)

  6. Innovative method for increased methane recovery from two-phase anaerobic digestion of food waste through reutilization of acidogenic off-gas in methanogenic reactor.

    PubMed

    Yan, Bing Hua; Selvam, Ammaiyappan; Wong, Jonathan W C

    2016-10-01

    In this study, the performance of a two-phase anaerobic digestion reactor treating food waste with the reutilization of acidogenic off-gas was investigated with the objective to improve the hydrogen availability for the methanogenic reactor. As a comparison a treatment without off-gas reutilization was also set up. Results showed that acidogenic off-gas utilization in the upflow anaerobic sludge blanket (UASB) reactor increased the methane recovery up to 38.6%. In addition, a 27% increase in the production of cumulative chemical oxygen demand (COD) together with an improved soluble microbial products recovery dominated by butyrate was observed in the acidogenic leach bed reactor (LBR) with off-gas reutilization. Of the increased methane recovery, ∼8% was contributed by the utilization of acidogenic off-gas in UASB. Results indicated that utilization of acidogenic off-gas in methanogenic reactor is a viable technique for improving overall methane recovery.

  7. Thermal striping in nuclear reactors: POD analysis of LES simulations and experiment

    NASA Astrophysics Data System (ADS)

    Merzari, Elia; Alvarez, Andres; Marin, Oana; Obabko, Aleksandr; Lomperski, Steve; Aithal, Shashi

    2015-11-01

    Thermal fatigue caused due to thermal striping impacts design and analyses of a wide-range of industrial apparatus. This phenomena is of particular significance in nuclear reactor applications, primarily in sodium cooled fast reactors. In order to conduct systematic analyses of the thermal striping phenomena a simplified experimental set-up was designed and built at Argonne National Laboratory. In this set-up two turbulent jets with a temperature difference of about 20K were mixed in a rectangular tank. The jets entered the tank via 2 hexagonal inlets. Two different inlet geometries were studied, both experimentally and via high-fidelity LES simulations. Proper Orthogonal Decomposition (POD) was performed on the turbulent velocity field in the tank to identify the most dominant energetic modes. The POD analyses of the experimental data in both inlet geometrical configurations were compared with LES simulations. Detailed POD analyses are presented to highlight the impact of geometry on the velocity and thermal fields. These can be correlated with experimental and numerical data to assess the impact of thermal striping on the design of the upper plenum of sodium-cooled nuclear reactors. ALCF.

  8. Update on ORNL TRANSFORM Tool: Simulating Multi-Module Advanced Reactor with End-to-End I&C

    SciTech Connect

    Hale, Richard Edward; Fugate, David L.; Cetiner, Sacit M.; Qualls, A. L.

    2015-05-01

    The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the fourth year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled reactor) concepts, including the use of multiple coupled reactors at a single site. The focus of this report is the development of a steam generator and drum system model that includes the complex dynamics of typical steam drum systems, the development of instrumentation and controls for the steam generator with drum system model, and the development of multi-reactor module models that reflect the full power reactor innovative small module design concept. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor models; ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface technical area; and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the TRANSFORM tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the Advanced Reactors Technology program; (2) developing a library of baseline component modules that can be assembled into full plant models using available geometry, design, and thermal-hydraulic data; (3) defining modeling conventions for interconnecting component models; and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

  9. Regional scale atmospheric dispersion simulation of accidental releases of radionuclides from Fukushima Dai-ichi reactor

    NASA Astrophysics Data System (ADS)

    Srinivas, C. V.; Venkatesan, R.; Baskaran, R.; Rajagopal, V.; Venkatraman, B.

    2012-12-01

    This paper presents the results of regional scale atmospheric dispersion simulation of accidental emission of radionuclides from the Fukushima Daiichi Reactor, Japan following the Tohoku earthquake and tsunami event on 11 March 2011. The objective was to study the temporal behaviour of plume trajectory, concentration, deposition and radiation dose pattern over an 80 km range around the reactor. The time-varying meteorological parameters during the release period were simulated with a multi-scale nested atmospheric model WRF ARW and the trajectory, plume dispersion were computed with Lagrangian Particle Dispersion models HYSPLIT, FLEXPART using the available information on accidental source term. The simulations indicated that the wind flow over Japan during the release period was driven by the large scale extra-tropical westerly waves and associated low pressure systems. In the lower levels, the flow was influenced by the local topography/sea breeze causing occasional landward wind shift on the east coast of Japan. Simulated airflow trajectories revealed that the plume stayed over the ocean by westerly winds on most days and the radioactivity dispersed over sea surface. Landward trajectories were found on a few days due to southeasterly, easterly and northeasterly flow (15-17, 19-21 March 2011) during which much of the radionuclides deposited over the land region. The hotspot of depositions occurred over east Pacific Ocean near to Japan. Over the land relatively high depositions were simulated in a narrow zone of 20 km width and 80 km length in the northwest sector in agreement with monitor data. Simulations showed wet depositions over the land to be higher than the dry depositions during 12-30 March due to occurrence of rainfall on some days. Comparison of activity deposition and air dose values with available observations confirmed that the plume pattern in a finer length scale around the site could be simulated realistically and agree with the measurements

  10. Sensitivity of rarefied gas simulations of ground tests to gas surface models

    NASA Astrophysics Data System (ADS)

    Cline, Jason A.; Deschenes, Timothy R.; Quenneville, Jason; Taylor, Ramona S.

    2014-09-01

    The space environment produces a number of performance challenges to satellite and spacecraft manufacturers that require measurements, including effects from hyperthermal atomic oxygen, charged particles, magnetic fields, spacecraft charging, ultraviolet radiation, micrometeoroids, and cryogenic temperatures. Ground tests involving a simulated space environment help explore these challenges, but also benefit from simulations that predict the anticipated physical phenomena, or help reconcile the measured observations to physical parameters. We present an update and application of a flexible multi-physics software simulation framework intended for predicting space environment performance and ground-test simulations of spacecraft. In this specific application we show how the energy dependent erosion yield may be applied with a rarefied gas dynamics simulation to aid comparison of terrestrial erosion rate measurements and on-orbit materials degradation. For the considered fluoropolymer material, we found that explicit consideration of the atomic oxygen energy distribution could potentially modify the expected correspondence between ground tests and space by 67%.

  11. Features of temperature control of fuel element cladding for pressurized water nuclear reactor ``WWER-1000'' while simulating reactor accidents

    NASA Astrophysics Data System (ADS)

    Zaytsev, P. A.; Priymak, S. V.; Usachev, V. B.; Oleynikov, P. P.; Soldatkin, D. M.

    2013-09-01

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones "cladding-junction" and "junction-coolant". The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ˜5% of maximum value by the final moment of the stage of linear increase in the temperature.

  12. The Gamma Ray Spectra Measurement in Reactor Pressure Vessel Simulator in WWER-1000 Engineering Benchmark

    NASA Astrophysics Data System (ADS)

    Ošmera, B.; Kyncl, J.; Mařík, M.; Cvachovec, F.; Smutný, V.; Králík, M.

    2003-06-01

    The space - energy distribution of the mixed neutron - photon radiation field has been measured over the Reactor Pressure Vessel (RPV) simulator thickness in the WWER-1000 engineering benchmark assembly in the LR-0 experimental reactor (in Nuclear Research Institute Řež plc) with a scintillation spectrometer. The spectra have been measured before the RPV in one quarter, one half, and three quarters of its thickness and behind the RPV in the energy range 0.5 - 10 MeV. The evaluated integral fluxes above 1 MeV and their ratio are compared with the MCNP and DORT calculation, the measured spectra are presented graphically. The measurements are being performed in the frame of the project REDOS [1], 5th Framwork Programme of the European Community 1998 - 2002.

  13. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding. [PWR

    SciTech Connect

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed.

  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. Fast simulations of gas sloshing and cold front formation

    NASA Astrophysics Data System (ADS)

    Roediger, E.; Zuhone, J. A.

    2012-01-01

    We present a simplified and fast method for simulating minor mergers between galaxy clusters. Instead of following the evolution of the dark matter haloes directly by the N-body method, we employ a rigid potential approximation for both clusters. The simulations are run in the rest frame of the more massive cluster and account for the resulting inertial accelerations in an optimized way. We test the reliability of this method for studies of minor merger induced gas sloshing by performing a one-to-one comparison between our simulations and hydro+N-body ones. We find that the rigid potential approximation reproduces the sloshing-related features well except for two artefacts: the temperature just outside the cold fronts is slightly overpredicted, and the outward motion of the cold fronts is delayed by typically 200 Myr. We discuss reasons for both artefacts.

  16. Process Simulation of Gas Metal Arc Welding Software

    2005-09-06

    ARCWELDER is a Windows-based application that simulates gas metal arc welding (GMAW) of steel and aluminum. The software simulates the welding process in an accurate and efficient manner, provides menu items for process parameter selection, and includes a graphical user interface with the option to animate the process. The user enters the base and electrode material, open circuit voltage, wire diameter, wire feed speed, welding speed, and standoff distance. The program computes the size andmore » shape of a square-groove or V-groove weld in the flat position. The program also computes the current, arc voltage, arc length, electrode extension, transfer of droplets, heat input, filler metal deposition, base metal dilution, and centerline cooling rate, in English or SI units. The simulation may be used to select welding parameters that lead to desired operation conditions.« less

  17. Fast Simulations of Gas Sloshing and Cold Front Formation

    NASA Technical Reports Server (NTRS)

    Roediger, E.; ZuHone, J. A.

    2012-01-01

    We present a simplified and fast method for simulating minor mergers between galaxy clusters. Instead of following the evolution of the dark matter halos directly by the N-body method, we employ a rigid potential approximation for both clusters. The simulations are run in the rest frame of the more massive cluster and account for the resulting inertial accelerations in an optimised way. We test the reliability of this method for studies of minor merger induced gas sloshing by performing a one-to-one comparison between our simulations and hydro+N-body ones. We find that the rigid potential approximation reproduces the sloshing-related features well except for two artifacts: the temperature just outside the cold fronts is slightly over-predicted, and the outward motion of the cold fronts is delayed by typically 200 Myr. We discuss reasons for both artifacts.

  18. Fast Simulations of Gas Sloshing and Cold Front Formation

    NASA Technical Reports Server (NTRS)

    Roediger, E.; ZuHone, J. A.

    2011-01-01

    We present a simplified and fast method for simulating minor mergers between galaxy clusters. Instead of following the evolution of the dark matter halos directly by the N-body method, we employ a rigid potential approximation for both clusters. The simulations are run in the rest frame of the more massive cluster and account for the resulting inertial accelerations in an optimised way. We test the reliability of this method for studies of minor merger induced gas sloshing by performing a one-to-one comparison between our simulations and hydro+N-body ones. We find that the rigid potential approximation reproduces the sloshing-related features well except for two artefacts: the temperature just outside the cold fronts is slightly over-predicted, and the outward motion of the cold fronts is delayed by typically 200 Myr. We discuss reasons for both artefacts.

  19. Process Simulation of Gas Metal Arc Welding Software

    SciTech Connect

    Murray, Paul E.

    2005-09-06

    ARCWELDER is a Windows-based application that simulates gas metal arc welding (GMAW) of steel and aluminum. The software simulates the welding process in an accurate and efficient manner, provides menu items for process parameter selection, and includes a graphical user interface with the option to animate the process. The user enters the base and electrode material, open circuit voltage, wire diameter, wire feed speed, welding speed, and standoff distance. The program computes the size and shape of a square-groove or V-groove weld in the flat position. The program also computes the current, arc voltage, arc length, electrode extension, transfer of droplets, heat input, filler metal deposition, base metal dilution, and centerline cooling rate, in English or SI units. The simulation may be used to select welding parameters that lead to desired operation conditions.

  20. Two critical issues in Langevin simulation of gas flows

    SciTech Connect

    Zhang, Jun; Fan, Jing

    2014-12-09

    A stochastic algorithm based on the Langevin equation has been recently proposed to simulate rarefied gas flows. Compared with the direct simulation Monte Carlo (DSMC) method, the Langevin method is more efficient in simulating small Knudsen number flows. While it is well-known that the cell sizes and time steps should be smaller than the mean free path and the mean collision time, respectively, in DSMC simulations, the Langevin equation uses a drift term and a diffusion term to describe molecule movements, so no direct molecular collisions have to be modeled. This enables the Langevin simulation to proceed with a much larger time step than that in the DSMC method. Two critical issues in Langevin simulation are addressed in this paper. The first issue is how to reproduce the transport properties as that described by kinetic theory. Transport coefficients predicted by Langevin equation are obtained by using Green-Kubo formulae. The second issue is numerical scheme with boundary conditions. We present two schemes corresponding to small time step and large time step, respectively. For small time step, the scheme is similar to DSMC method as the update of positions and velocities are uncoupled; for large time step, we present an analytical solution of the hitting time, which is the crucial factor for accurate simulation. Velocity-Couette flow, thermal-Couette flow, Rayleigh-Bénard flow and wall-confined problem are simulated by using these two schemes. Our study shows that Langevin simulation is a promising tool to investigate small Knudsen number flows.

  1. Numerical analysis of an impinging jet reactor for the CVD and gas-phase nucleation of titania

    NASA Astrophysics Data System (ADS)

    Gokoglu, Suleyman A.; Stewart, Gregory D.; Collins, Joshua; Rosner, Daniel E.

    1994-06-01

    We model a cold-wall atmospheric pressure impinging jet reactor to study the CVD and gas-phase nucleation of TiO2 from a titanium tetra-iso-propoxide (TTIP)/oxygen dilute source gas mixture in nitrogen. The mathematical model uses the computational code FIDAP and complements our recent asymptotic theory for high activation energy gas-phase reactions in thin chemically reacting sublayers. The numerical predictions highlight deviations from ideality in various regions inside the experimental reactor. Model predictions of deposition rates and the onset of gas-phase nucleation compare favorably with experiments. Although variable property effects on deposition rates are not significant (approximately 11 percent at 1000 K), the reduction rates due to Soret transport is substantial (approximately 75 percent at 1000 K).

  2. Numerical Analysis of an Impinging Jet Reactor for the CVD and Gas-Phase Nucleation of Titania

    NASA Technical Reports Server (NTRS)

    Gokoglu, Suleyman A.; Stewart, Gregory D.; Collins, Joshua; Rosner, Daniel E.

    1994-01-01

    We model a cold-wall atmospheric pressure impinging jet reactor to study the CVD and gas-phase nucleation of TiO2 from a titanium tetra-iso-propoxide (TTIP)/oxygen dilute source gas mixture in nitrogen. The mathematical model uses the computational code FIDAP and complements our recent asymptotic theory for high activation energy gas-phase reactions in thin chemically reacting sublayers. The numerical predictions highlight deviations from ideality in various regions inside the experimental reactor. Model predictions of deposition rates and the onset of gas-phase nucleation compare favorably with experiments. Although variable property effects on deposition rates are not significant (approximately 11 percent at 1000 K), the reduction rates due to Soret transport is substantial (approximately 75 percent at 1000 K).

  3. An assessment of coupling algorithms for nuclear reactor core physics simulations

    DOE PAGESBeta

    Hamilton, Steven; Berrill, Mark; Clarno, Kevin; Pawlowski, Roger; Toth, Alex; Kelley, C. T.; Evans, Thomas; Philip, Bobby

    2016-04-01

    Here we evaluate the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product was developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK andmore » Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Finally, both criticality (k-eigenvalue) and critical boron search problems are considered.« less

  4. Optimization of pressurized water reactor shuffling by simulated annealing with heuristics

    SciTech Connect

    Stevens, J.G.; Smith, K.S.; Rempe, K.R.; Downar, T.J.

    1995-09-01

    Simulated-annealing optimization of reactor core loading patterns is implemented with support for design heuristics during candidate pattern generation. The SIMAN optimization module uses the advanced nodal method of SIMULATE-3 and the full cross-section detail of CASMO-3 to evaluate accurately the neutronic performance of each candidate, resulting in high-quality patterns. The use of heuristics within simulated annealing is explored. Heuristics improve the consistency of optimization results for both fast- and slow-annealing runs with no penalty from the exclusion of unusual candidates. Thus, the heuristic application of designer judgment during automated pattern generation is shown to be effective. The capability of the SIMAN module to find and evaluate families of loading patterns that satisfy design constraints and have good objective performance within practical run times is demonstrated. The use of automated evaluations of successive cycles to explore multicycle effects of design decisions is discussed.

  5. An assessment of coupling algorithms for nuclear reactor core physics simulations

    NASA Astrophysics Data System (ADS)

    Hamilton, Steven; Berrill, Mark; Clarno, Kevin; Pawlowski, Roger; Toth, Alex; Kelley, C. T.; Evans, Thomas; Philip, Bobby

    2016-04-01

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss-Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton-Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

  6. Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072

    SciTech Connect

    Poncet, Bernard

    2013-07-01

    About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

  7. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    SciTech Connect

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  8. 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

    power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%.

  9. Effects of a range of machined and ground surface finishes on the simulated reactor helium corrosion of several candidate structural materials. [Inconel MA 754

    SciTech Connect

    Thompson, L.D.

    1981-02-01

    This report discusses the corrosion behavior of several candidate reactor structural alloys in a simulated advanced high-temperature gas-cooled reactor (HTGR) environment over a range of lathe-machined and centerless-ground surface finishes. The helium environment contained 50 Pa H/sub 2//5 Pa CO/5 Pa CH/sub 4//<0.05 Pa H/sub 2/O (500 ..mu..atm H/sub 2//50 ..mu..atm CO/50 ..mu..atm CH/sub 4//<0.5 ..mu..atm H/sub 2/O) at 900/sup 0/C for a total exposure of 3000 h. The test alloys included two vacuum-cast superalloys (IN 100 and IN 713LC); a centrifugally cast austenitic alloy (HK 40); three wrought high-temperature alloys (Alloy 800H, Hastelloy X, and Inconel 617); and a nickel-base oxide-dispersion-strengthened alloy (Inconel MA 754). Surface finish variations did not affect the simulated advanced-HTGR corrosion behavior of these materials. Under these conditions, the availability of reactant gaseous impurities controls the kinetics of the observed gas-metal interactions. Variations in the near-surface activities and mobilities of reactive solute elements, such as chromium, which might be expected to be affected by changes in surface finish, do not seem to greatly influence corrosion in this simulated advanced HTGR environment. 18 figures, 4 tables.

  10. Analysis of a gas-liquid film plasma reactor for organic compound oxidation.

    PubMed

    Hsieh, Kevin; Wang, Huijuan; Locke, Bruce R

    2016-11-01

    A pulsed electrical discharge plasma formed in a tubular reactor with flowing argon carrier gas and a liquid water film was analyzed using methylene blue as a liquid phase hydroxyl radical scavenger and simultaneous measurements of hydrogen peroxide formation. The effects of liquid flow rate, liquid conductivity, concentration of dye, and the addition of ferrous ion on dye decoloration and degradation were determined. Higher liquid flow rates and concentrations of dye resulted in less decoloration percentages and hydrogen peroxide formation due to initial liquid conductivity effects and lower residence times in the reactor. The highest decoloration energy yield of dye found in these studies was 5.2g/kWh when using the higher liquid flow rate and adding the catalyst. The non-homogeneous nature of the plasma discharge favors the production of hydrogen peroxide in the plasma-liquid interface over the chemical oxidation of the organic in the bulk liquid phase and post-plasma reactions with the Fenton catalyst lead to complete utilization of the plasma-formed hydrogen peroxide. PMID:27267693

  11. 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.

  12. Analysis of a gas-liquid film plasma reactor for organic compound oxidation.

    PubMed

    Hsieh, Kevin; Wang, Huijuan; Locke, Bruce R

    2016-11-01

    A pulsed electrical discharge plasma formed in a tubular reactor with flowing argon carrier gas and a liquid water film was analyzed using methylene blue as a liquid phase hydroxyl radical scavenger and simultaneous measurements of hydrogen peroxide formation. The effects of liquid flow rate, liquid conductivity, concentration of dye, and the addition of ferrous ion on dye decoloration and degradation were determined. Higher liquid flow rates and concentrations of dye resulted in less decoloration percentages and hydrogen peroxide formation due to initial liquid conductivity effects and lower residence times in the reactor. The highest decoloration energy yield of dye found in these studies was 5.2g/kWh when using the higher liquid flow rate and adding the catalyst. The non-homogeneous nature of the plasma discharge favors the production of hydrogen peroxide in the plasma-liquid interface over the chemical oxidation of the organic in the bulk liquid phase and post-plasma reactions with the Fenton catalyst lead to complete utilization of the plasma-formed hydrogen peroxide.

  13. Characteristics of gas-liquid pulsed discharge plasma reactor and dye decoloration efficiency.

    PubMed

    Sun, Bing; Aye, Nyein Nyein; Gao, Zhiying; Lv, Dan; Zhu, Xiaomei; Sato, Masayuki

    2012-01-01

    The pulsed high-voltage discharge is a new advanced oxidation technology for water treatment. Methyl Orange (MO) dye wastewater was chosen as the target object. Some investigations were conducted on MO decoloration including the discharge characteristics of the multi-needle reactor, parameter optimization, and the degradation mechanism. The following results were obtained. The color group of the azo dye MO was effectively decomposed by water surface plasma. The decoloration rate was promoted with the increase of treatment time, peak voltage, and pulse frequency. When the initial conductivity was 1700 microS/cm, the decoloration rate was the highest. The optimum distance between the needle electrodes and the water surface was 1 mm, the distance between the grounding electrode and the water surface was 28 mm, and the number of needle electrodes and spacing between needles were 24 and 7.5 mm, respectively. The decoloration rate of MO was affected by the gas in the reactor and varied in the order oxygen > air> argon > nitrogen, and the energy yield obtained in this investigation was 0.45 g/kWh.

  14. Thermal-Mechanical Studies for Gas-Cooled Space Reactor Designs

    SciTech Connect

    Kapernick, Richard J.; Creamer, William C.

    2006-01-20

    Los Alamos National Laboratory has been involved in the development of reactor concepts to be used as a power source for nuclear electric propulsion and/or for surface power sources. As part of this effort, a high fidelity thermal-mechanical analysis method has been developed for rapid performance assessments of these designs. This method has been used to study several concept alternatives, including both annular and multi-hole monolithic block designs. This paper presents the analysis method developed and results of analyses performed for a gas-cooled reactor. Key results are 1) the annular block design is lower mass than the multi-hole block design, 2) fuel temperatures are effectively controlled by adjusting the number of fuel pins in the core, 3) large thermal-hydraulic performance enhancements are produced by increasing coolant pressure and/or helium mole fraction, and 4) manufacturing and assembly parameters have relatively small effects on thermal-hydraulic performance and care should be taken to balance mechanical design complexity and reliability issues with thermal-hydraulic performance.

  15. Upgrading light olefin fuel gas and catalytic reformate in a turbulent fluidized bed catalyst reactor

    SciTech Connect

    Kushnerick, J.D.; Harandi, M.N.; Owen, H.

    1989-05-02

    A fluidized bed catalyst process is described for conversion of C/sub 4/- light olefinic gas feedstock to C/sub 5/+ hydrocarbons and for conversion of C/sub 6/ to C/sub 8/ aromatic feedstock to C/sub 7/ to C/sub 11/ aromatics. It consists of: maintaining a fluidized bed of zeolite catalyst particles in a turbulent reactor bed at a temperature of about 500/sup 0/ to 800/sup 0/F, the catalyst having an apparent particle density of about 0.9 to 1.6 g/cm/sup 3/ and a size range of about 1 to 150 microns, an average catalyst particle size of about 20 to 100 microns containing about 10 to 25 weight percent of fine particles having a particle size less than 32 microns; contacting the feedstocks and passing the feedstocks upwardly through the fluidized catalyst bed under turbulent flow conditions at reaction conditions sufficient to convert at least about 70% of the olefin feedstock and to convert at least 5% of the C/sub 6/ to C/sub 8/ aromatic feedstock; maintaining turbulent fluidized bed conditions through the reactor bed between transition velocity and transport velocity at a superficial fluid velocity of about 0.3 to 2 meters per second; and recovering hydrocarbon product containing C/sub 5/+ hydrocarbons C/sub 7/ to C/sub 11/ aromatic hydrocarbons.

  16. 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.

  17. Technical basis for extending storage of the UK's advanced gas-cooled reactor fuel

    SciTech Connect

    Hambley, D.I.

    2013-07-01

    The UK Nuclear Decommissioning Agency has recently declared a date for cessation of reprocessing of oxide fuel from the UK's Advanced Gas-cooled Reactors (AGRs). This will fundamentally change the management of AGR fuel: from short term storage followed by reprocessing to long term fuel storage followed, in all likelihood, by geological disposal. In terms of infrastructure, the UK has an existing, modern wet storage asset that can be adapted for centralised long term storage of dismantled AGR fuel under the required pond water chemistry. No AGR dry stores exist, although small quantities of fuel have been stored dry as part of experimental programmes in the past. These experimental programmes have shown concerns about corrosion rates.

  18. A demonstration of a whole core neutron transport method in a gas cooled reactor

    SciTech Connect

    Connolly, K. J.; Rahnema, F.

    2013-07-01

    This paper illustrates a capability of the whole core transport method COMET. Building on previous works which demonstrated the accuracy of the method, this work serves to emphasize the robust capability of the method while also accentuating its efficiency. A set of core configurations is presented based on an operating gas-cooled thermal reactor, Japan's HTTR, and COMET determines the eigenvalue and fission density profile throughout each core configuration. Results for core multiplication factors are compared to MCNP for accuracy and also to compare runtimes. In all cases, the values given by COMET differ by those given by MCNP by less than the uncertainty inherent in the stochastic solution procedure, however, COMET requires runtimes shorter on the order of a few hundred. Figures are provided illustrating the whole core fission density profile, with segments of pins explicitly modeled individually, so that pin-level neutron flux behavior can be seen without any approximation due to simplification strategies such as homogenization. (authors)

  19. High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics

    SciTech Connect

    Larry Demick

    2011-08-01

    This paper summarizes the potential market for process heat produced by a high temperature gas-cooled reactor (HTGR), the environmental benefits reduced CO2 emissions will have on these markets, and the typical economics of projects using these applications. It gives examples of HTGR technological applications to industrial processes in the typical co-generation supply of process heat and electricity, the conversion of coal to transportation fuels and chemical process feedstock, and the production of ammonia as a feedstock for the production of ammonia derivatives, including fertilizer. It also demonstrates how uncertainties in capital costs and financial factors affect the economics of HTGR technology by analyzing the use of HTGR technology in the application of HTGR and high temperature steam electrolysis processes to produce hydrogen.

  20. NGNP: High Temperature Gas-Cooled Reactor Key Definitions, Plant Capabilities, and Assumptions

    SciTech Connect

    Phillip Mills

    2012-02-01

    This document is intended to provide a Next Generation Nuclear Plant (NGNP) Project tool in which to collect and identify key definitions, plant capabilities, and inputs and assumptions to be used in ongoing efforts related to the licensing and deployment of a high temperature gas-cooled reactor (HTGR). These definitions, capabilities, and assumptions are extracted from a number of sources, including NGNP Project documents such as licensing related white papers [References 1-11] and previously issued requirement documents [References 13-15]. Also included is information agreed upon by the NGNP Regulatory Affairs group's Licensing Working Group and Configuration Council. The NGNP Project approach to licensing an HTGR plant via a combined license (COL) is defined within the referenced white papers and reference [12], and is not duplicated here.

  1. Lattice gas simulations of dynamical geometry in two dimensions

    NASA Astrophysics Data System (ADS)

    Klales, Anna; Cianci, Donato; Needell, Zachary; Meyer, David A.; Love, Peter J.

    2010-10-01

    We present a hydrodynamic lattice gas model for two-dimensional flows on curved surfaces with dynamical geometry. This model is an extension to two dimensions of the dynamical geometry lattice gas model previously studied in one dimension. We expand upon a variation of the two-dimensional flat space Frisch-Hasslacher-Pomeau (FHP) model created by Frisch [Phys. Rev. Lett.PRLTAO0031-9007 56, 1505 (1986)]10.1103/PhysRevLett.56.1505 and independently by Wolfram, and modified by Boghosian [Philos. Trans. R. Soc. London, Ser. A 360, 333 (2002)]10.1098/rsta.2001.0933. We define a hydrodynamic lattice gas model on an arbitrary triangulation whose flat space limit is the FHP model. Rules that change the geometry are constructed using the Pachner moves, which alter the triangulation but not the topology. We present results on the growth of the number of triangles as a function of time. Simulations show that the number of triangles grows with time as t1/3 , in agreement with a mean-field prediction. We also present preliminary results on the distribution of curvature for a typical triangulation in these simulations.

  2. Monte Carlo simulations of microgap gas-filled proportional counters

    NASA Astrophysics Data System (ADS)

    Kundu, Ashoke; Morton, Edward J.; Key, Martyn J.; Luggar, Russell D.

    1999-09-01

    Monte Carlo calculations have been widely employed to model the interactions of electrons and photons as they travel through and collide with matter. This approach has been applied with some success to the problem of simulating the response of gas-filled proportional counters, mapping out electron transport through the electric field on an interaction-by-interaction basis. These studies focus on the multiplication of electrons as they drift into the high electric field region of the detector and subsequently avalanche. We are using this technique in our new simulation code to depict avalanching in microgap gas-filled proportional counters, in order to investigate the variation of two principle detector properties with the anode pitch used in the detector. Spatial resolution information can be obtained by measuring the lateral diffusion distance of an electron from the point where it is liberated to the point in the detector where it initiates an avalanche. By also modeling the motion of the positive ions that are left behind from the initial avalanche, we are able to gauge the effect of space charge distortion on subsequent avalanches. This effect is particularly important at the high X-ray count rates that we are interested in for our ultimate aim, which is to use the detectors as part of a high-speed tomography system for imaging multiphase oil/water/gas flows.

  3. Lattice gas simulations of dynamical geometry in two dimensions.

    PubMed

    Klales, Anna; Cianci, Donato; Needell, Zachary; Meyer, David A; Love, Peter J

    2010-10-01

    We present a hydrodynamic lattice gas model for two-dimensional flows on curved surfaces with dynamical geometry. This model is an extension to two dimensions of the dynamical geometry lattice gas model previously studied in one dimension. We expand upon a variation of the two-dimensional flat space Frisch-Hasslacher-Pomeau (FHP) model created by Frisch [Phys. Rev. Lett. 56, 1505 (1986)] and independently by Wolfram, and modified by Boghosian [Philos. Trans. R. Soc. London, Ser. A 360, 333 (2002)]. We define a hydrodynamic lattice gas model on an arbitrary triangulation whose flat space limit is the FHP model. Rules that change the geometry are constructed using the Pachner moves, which alter the triangulation but not the topology. We present results on the growth of the number of triangles as a function of time. Simulations show that the number of triangles grows with time as t(1/3), in agreement with a mean-field prediction. We also present preliminary results on the distribution of curvature for a typical triangulation in these simulations.

  4. Simulation of natural gas production from submarine gas hydrate deposits combined with carbon dioxide storage

    NASA Astrophysics Data System (ADS)

    Janicki, Georg; Schlüter, Stefan; Hennig, Torsten; Deerberg, Görge

    2013-04-01

    The recovery of methane from gas hydrate layers that have been detected in several submarine sediments and permafrost regions around the world so far is considered to be a promising measure to overcome future shortages in natural gas as fuel or raw material for chemical syntheses. Being aware that natural gas resources that can be exploited with conventional technologies are limited, research is going on to open up new sources and develop technologies to produce methane and other energy carriers. Thus various research programs have started since the early 1990s in Japan, USA, Canada, South Korea, India, China and Germany to investigate hydrate deposits and develop technologies to destabilize the hydrates and obtain the pure gas. In recent years, intensive research has focussed on the capture and storage of carbon dioxide from combustion processes to reduce climate change. While different natural or manmade reservoirs like deep aquifers, exhausted oil and gas deposits or other geological formations are considered to store gaseous or liquid carbon dioxide, the storage of carbon dioxide as hydrate in former methane hydrate fields is another promising alternative. Due to beneficial stability conditions, methane recovery may be well combined with CO2 storage in form of hydrates. This has been shown in several laboratory tests and simulations - technical field tests are still in preparation. Within the scope of the German research project »SUGAR«, different technological approaches are evaluated and compared by means of dynamic system simulations and analysis. Detailed mathematical models for the most relevant chemical and physical effects are developed. The basic mechanisms of gas hydrate formation/dissociation and heat and mass transport in porous media are considered and implemented into simulation programs like CMG STARS and COMSOL Multiphysics. New simulations based on field data have been carried out. The studies focus on the evaluation of the gas production

  5. 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.

  6. Simulated Verification of Fuel Element Inventory in a Small Reactor Core Using the Nuclear Materials Identification System (NMIS)

    SciTech Connect

    Grogan, Brandon R; Mihalczo, John T

    2009-01-01

    The International Panel on Climate Change projects that by 2050 the world energy demand may double. Although the primary focus for new nuclear power plants in industrialized nations is on large plants in the 1000-1600 MWe range, there is an increasing demand for small and medium reactors (SMRs). About half of the innovative SMR concepts are for small (<300 MWe) reactors with a 5-30 year life without on-site refueling. This type of reactor is also known as a battery-type reactor. These reactors are particularly attractive to countries with small power grids and for non-electrical purposes such as heating, hydrogen production, and seawater desalination. Traditionally, this type of reactor has been used in a nautical propulsion role. This type of reactor is designed as a permanently sealed unit to prevent the diversion of the uranium in the core by the user. However, after initial fabrication it will be necessary to verify that the newly fabricated reactor core contains the quantity of uranium that initially entered the fuel fabrication plant. In most instances, traditional inspection techniques can be used to perform this verification, but in certain situations the core design will be considered sensitive. Non-intrusive verification techniques must be utilized in these situations. The Nuclear Materials Identification System (NMIS) with imaging uses active interrogation and a fast time correlation processor to characterize fissile material. The MCNP-PoliMi computer code was used to simulate NMIS measurements of a small, sealed reactor core. Because most battery-type reactor designs are still in the early design phase, a more traditional design based on a Russian icebreaker core was used in the simulations. These simulations show how the radiography capabilities of the NMIS could be used to detect the diversion of fissile material by detecting void areas in the assembled core where fuel elements have been removed.

  7. 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

  8. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

    ERIC Educational Resources Information Center

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

  9. Simulation study of nanoparticle coating in a low pressure plasma reactor

    SciTech Connect

    Pourali, N.; Foroutan, G.

    2015-02-15

    A self-consistent combination of plasma fluid model, nanoparticle heating model, and surface deposition model is used to investigate the coating of nanosize particles by amorphous carbon layers in a low pressure plasma reactor. The numerical results show that, owing to the net heat release in the surface reactions, the particle temperature increases and its equilibrium value remains always 50 K above the background gas temperature. The deposition rate decreases with increasing of the particle temperature and the corresponding time scale is of the order of 10 ms. The deposition rate is also strongly affected by the change in plasma parameters. When the electron temperature is increased, the deposition rate first increases due to the enhanced ion and radical generation, shows a maximum and then declines as the particle temperature rises above the gas temperature. An enhancement in the background gas pressure and/or temperature leads to a reduction in the deposition rate, which can be explained in terms of the enhanced etching by atomic hydrogen and particle heating by the background gas.

  10. Optimizing Noble Gas-Water Interactions via Monte Carlo Simulations.

    PubMed

    Warr, Oliver; Ballentine, Chris J; Mu, Junju; Masters, Andrew

    2015-11-12

    In this work we present optimized noble gas-water Lennard-Jones 6-12 pair potentials for each noble gas. Given the significantly different atomic nature of water and the noble gases, the standard Lorentz-Berthelot mixing rules produce inaccurate unlike molecular interactions between these two species. Consequently, we find simulated Henry's coefficients deviate significantly from their experimental counterparts for the investigated thermodynamic range (293-353 K at 1 and 10 atm), due to a poor unlike potential well term (εij). Where εij is too high or low, so too is the strength of the resultant noble gas-water interaction. This observed inadequacy in using the Lorentz-Berthelot mixing rules is countered in this work by scaling εij for helium, neon, argon, and krypton by factors of 0.91, 0.8, 1.1, and 1.05, respectively, to reach a much improved agreement with experimental Henry's coefficients. Due to the highly sensitive nature of the xenon εij term, coupled with the reasonable agreement of the initial values, no scaling factor is applied for this noble gas. These resulting optimized pair potentials also accurately predict partitioning within a CO2-H2O binary phase system as well as diffusion coefficients in ambient water. This further supports the quality of these interaction potentials. Consequently, they can now form a well-grounded basis for the future molecular modeling of multiphase geological systems. PMID:26452070

  11. Object-oriented approach for gas turbine engine simulation

    NASA Technical Reports Server (NTRS)

    Curlett, Brian P.; Felder, James L.

    1995-01-01

    An object-oriented gas turbine engine simulation program was developed. This program is a prototype for a more complete, commercial grade engine performance program now being proposed as part of the Numerical Propulsion System Simulator (NPSS). This report discusses architectural issues of this complex software system and the lessons learned from developing the prototype code. The prototype code is a fully functional, general purpose engine simulation program, however, only the component models necessary to model a transient compressor test rig have been written. The production system will be capable of steady state and transient modeling of almost any turbine engine configuration. Chief among the architectural considerations for this code was the framework in which the various software modules will interact. These modules include the equation solver, simulation code, data model, event handler, and user interface. Also documented in this report is the component based design of the simulation module and the inter-component communication paradigm. Object class hierarchies for some of the code modules are given.

  12. Simulation of chemical-vapor-deposited silicon carbide for a cold wall vertical reactor

    NASA Astrophysics Data System (ADS)

    Lee, Y. L.; Sanchez, J. M.

    1997-07-01

    The growth rate of silicon carbide obtained by low-pressure chemical vapor deposition from tetramethylsilane is numerically simulated for a cold wall vertical reactor. The transport equations for momentum, heat, and mass transfer are simultaneously solved by employing the finite volume method. A model for reaction rate is also proposed in order to predict the measured growth rates [A. Figueras, S. Garelik, J. Santiso, R. Rodroguez-Clemente, B. Armas, C. Combescure, R. Berjoan, J.M. Saurel and R. Caplain, Mater. Sci. Eng. B 11 (1992) 83]. Finally, the effects of thermal diffusion on the growth rate are investigated.

  13. Parametric study of spallation targets for the MYRRHA reactor using MCNPX simulations

    NASA Astrophysics Data System (ADS)

    Rebello, A. L. P.; Martinez, A. S.; Gonçalves, A. C.

    2014-06-01

    The present work aims to evaluate the behavior of neutron multiplicity in a spallation target using MCNPX simulations, focusing on its application in the MYRRHA reactor. It was studied the two types of spallation target proposed for the MYRRHA project, windowless and windows target, in order to compare them and find saturation boundaries. Some saturation boundaries were found and the windowless target proved to be as viable as the windows one. Each one produced nearly the same number of neutrons per incident proton. Using the concept of neutron cost, it was also observed that the optimum conditions on neutron production occur at about 1GeV, for both target designs.

  14. A syncopated leap-frog algorithm for orbit consistent plasma simulation of materials processing reactors

    SciTech Connect

    Cobb, J.W.; Leboeuf, J.N.

    1994-10-01

    The authors present a particle algorithm to extend simulation capabilities for plasma based materials processing reactors. The orbit integrator uses a syncopated leap-frog algorithm in cylindrical coordinates, which maintains second order accuracy, and minimizes computational complexity. Plasma source terms are accumulated orbit consistently directly in the frequency and azimuthal mode domains. Finally they discuss the numerical analysis of this algorithm. Orbit consistency greatly reduces the computational cost for a given level of precision. The computational cost is independent of the degree of time scale separation.

  15. Comparison of calculated and measured fluid jet impingement velocities in a simulated reactor geometry

    SciTech Connect

    Hertrick, A.K.; Bryan, W.J.; Dorogy, G.M.; Hopkins, R.J.; Riddell, R.A.; Schwirian, E.R.

    1984-11-01

    Comparisons of analytical and experimental results are presented for the fluid jetting resulting from the existence of small gaps between parallel flow regions with dissimilar hydraulic characteristics. The experiment simulates the baffle gaps between a nuclear reactor core and the peripheral region around it, called the barrel-baffle region. Baffle gap fluid velocities are measured by a technique in which the only disturbance to the gap flow is a small pressure tap in the gap wall. The analysis uses an iterative, hydraulic network approach and is shown to yield good results when compared to the measured gap jet velocity and pressure drop distributions.

  16. 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

  17. 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

  18. Evaluation of dense-gas simulation models. Final report

    SciTech Connect

    Zapert, J.G.; Londergan, R.J.; Thistle, H.

    1991-05-01

    The report describes the approach and presents the results of an evaluation study of seven dense gas simulation models using data from three experimental programs. The models evaluated are two in the public domain (DEGADIS and SLAB) and five that are proprietary (AIRTOX, CHARM, FOCUS, SAFEMODE, and TRACE). The data bases used in the evaluation are the Desert Tortoise Pressurized Ammonia Releases, Burro Liquefied Natural Gas Spill Tests and the Goldfish Anhydrous Hydroflouric Acid Spill Experiments. A uniform set of performance statistics are calculated and tabulated to compare maximum observed concentrations and cloud half-width to those predicted by each model. None of the models demonstrated good performance consistently for all three experimental programs.

  19. Data Collection Methods for Validation of Advanced Multi-Resolution Fast Reactor Simulations

    SciTech Connect

    Tokuhiro, Akiro; Ruggles, Art; Pointer, David

    2015-01-22

    In pool-type Sodium Fast Reactors (SFR) the regions most susceptible to thermal striping are the upper instrumentation structure (UIS) and the intermediate heat exchanger (IHX). This project experimentally and computationally (CFD) investigated the thermal mixing in the region exiting the reactor core to the UIS. The thermal mixing phenomenon was simulated using two vertical jets at different velocities and temperatures as prototypic of two adjacent channels out of the core. Thermal jet mixing of anticipated flows at different temperatures and velocities were investigated. Velocity profiles are measured throughout the flow region using Ultrasonic Doppler Velocimetry (UDV), and temperatures along the geometric centerline between the jets were recorded using a thermocouple array. CFD simulations, using COMSOL, were used to initially understand the flow, then to design the experimental apparatus and finally to compare simulation results and measurements characterizing the flows. The experimental results and CFD simulations show that the flow field is characterized into three regions with respective transitions, namely, convective mixing, (flow direction) transitional, and post-mixing. Both experiments and CFD simulations support this observation. For the anticipated SFR conditions the flow is momentum dominated and thus thermal mixing is limited due to the short flow length associated from the exit of the core to the bottom of the UIS. This means that there will be thermal striping at any surface where poorly mixed streams impinge; rather unless lateral mixing is ‘actively promoted out of the core, thermal striping will prevail. Furthermore we note that CFD can be considered a ‘separate effects (computational) test’ and is recommended as part of any integral analysis. To this effect, poorly mixed streams then have potential impact on the rest of the SFR design and scaling, especially placement of internal components, such as the IHX that may see poorly mixed

  20. Large-eddy simulations of turbulent flow for grid-to-rod fretting in nuclear reactors

    SciTech Connect

    Bakosi, J.; Christon, M. A.; Lowrie, R. B.; Pritchett-Sheats, L. A.; Nourgaliev, R. R.

    2013-07-12

    The grid-to-rod fretting (GTRF) problem in pressurized water reactors is a flow-induced vibration problem that results in wear and failure of the fuel rods in nuclear assemblies. In order to understand the fluid dynamics of GTRF and to build an archival database of turbulence statistics for various configurations, implicit large-eddy simulations of time-dependent single-phase turbulent flow have been performed in 3 × 3 and 5 × 5 rod bundles with a single grid spacer. To assess the computational mesh and resolution requirements, a method for quantitative assessment of unstructured meshes with no-slip walls is described. The calculations have been carried out using Hydra-TH, a thermal-hydraulics code developed at Los Alamos for the Consortium for Advanced Simulation of Light water reactors, a United States Department of Energy Innovation Hub. Hydra-TH uses a second-order implicit incremental projection method to solve the singlephase incompressible Navier-Stokes equations. The simulations explicitly resolve the large scale motions of the turbulent flow field using first principles and rely on a monotonicity-preserving numerical technique to represent the unresolved scales. Each series of simulations for the 3 × 3 and 5 × 5 rod-bundle geometries is an analysis of the flow field statistics combined with a mesh-refinement study and validation with available experimental data. Our primary focus is the time history and statistics of the forces loading the fuel rods. These hydrodynamic forces are believed to be the key player resulting in rod vibration and GTRF wear, one of the leading causes for leaking nuclear fuel which costs power utilities millions of dollars in preventive measures. As a result, we demonstrate that implicit large-eddy simulation of rod-bundle flows is a viable way to calculate the excitation forces for the GTRF problem.