Science.gov

Sample records for gas reactor simulator

  1. Simulation of Water Gas Shift Zeolite Membrane Reactor

    NASA Astrophysics Data System (ADS)

    Makertiharta, I. G. B. N.; Rizki, Z.; Zunita, Megawati; Dharmawijaya, P. T.

    2017-07-01

    The search of alternative energy sources keeps growing from time to time. Various alternatives have been introduced to reduce the use of fossil fuel, including hydrogen. Many pathways can be used to produce hydrogen. Among all of those, the Water Gas Shift (WGS) reaction is the most common pathway to produce high purity hydrogen. The WGS technique faces a downstream processing challenge due to the removal hydrogen from the product stream itself since it contains a mixture of hydrogen, carbon dioxide and also the excess reactants. An integrated process using zeolite membrane reactor has been introduced to improve the performance of the process by selectively separate the hydrogen whilst boosting the conversion. Furthermore, the zeolite membrane reactor can be further improved via optimizing the process condition. This paper discusses the simulation of Zeolite Membrane Water Gas Shift Reactor (ZMWGSR) with variation of process condition to achieve an optimum performance. The simulation can be simulated into two consecutive mechanisms, the reaction prior to the permeation of gases through the zeolite membrane. This paper is focused on the optimization of the process parameters (e.g. temperature, initial concentration) and also membrane properties (e.g. pore size) to achieve an optimum product specification (concentration, purity).

  2. GAS-PASS/H : a simulation code for gas reactor plant systems.

    SciTech Connect

    Vilim, R. B.; Mertyurek, U.; Cahalan, J. E.; Nuclear Engineering Division; Texas A&M Univ.

    2004-01-01

    A simulation code for gas reactor plant systems has been developed. The code is intended for use in safety analysis and control studies for Generation-IV reactor concepts. We developed it anticipating an immediate application to direct cycle gas reactors. By programming in flexibility as to how components can be configured, we believe the code can be adapted for the indirect-cycle gas reactors relatively easy. The use of modular components and a general purpose equation solver allows for this. There are several capabilities that are included for investigating issues associated with direct cycle gas reactors. Issues include the safety characteristics of single shaft plants during coastdown and transition to shutdown heat removal following unprotected accidents, including depressurization, and the need for safety grade control systems. Basic components provided include turbine, compressor, recuperator, cooler, bypass valve, leak, accumulator, containment, and flow junction. The code permits a more rapid assessment of design concepts than is achievable using RELAP. RELAP requires detail beyond what is necessary at the design scoping stage. This increases the time to assemble an input deck and tends to make the code slower running. The core neutronics and decay heat models of GAS-PASS/H are taken from the liquid-metal version of MINISAS. The ex-reactor component models were developed from first principles. The network-based method for assembling component models into a system uses a general nonlinear solver to find the solution to the steady-state equations. The transient time-differenced equations are solved implicitly using the same solver. A direct cycle gas reactor is modeled and a loss of generator load transient is simulated for this reactor. While normally the reactor is scrammed, the plant safety case will require analysis of this event with failure of various safety systems. Therefore, we simulated the loss of load transient with a combined failure of the

  3. 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.

  4. Accident Analysis Simulation in Modular 300MWt Gas Cooled Fast Reactor

    NASA Astrophysics Data System (ADS)

    Zaki, Su'ud

    2017-01-01

    Safety analysis of 300MWt helium gas cooled long-life fast reactors has been performed. The analysis of unprotected loss of flow(ULOF) and unprotected rod run-out transient overpower (UTOP) are discussed. Some simulations for 300 MWt He gas cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations. GCFR relatively has hard spectrum so it has relatively small Doppler coefficient. In the UTOP accident case the analysis has been performed against external reactivity up to 0.002dk/k. In addition the steam generator design has also consider excess power during severe UTOP case..

  5. 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.

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

    NASA Technical Reports Server (NTRS)

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

    2009-01-01

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

  7. 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.

  8. 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.

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

    SciTech Connect

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

    1980-01-01

    Variable width ribbon heating elements have been fabricated which provide a chopped cosine, variable heat flux profile for fuel rod 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 result in the derivation of an analytical expression for the ribbon contours. From this, the ribbons are machined and wound on numerically controlled equipment. Postprocessing and inspection results in a wound, variable width ribbon with the precise dimensional, electrical, and mechanical properties needed for use in fuel pin simulators.

  10. Performance of a catalytic reactor at simulated gas turbine combustor operating conditions

    NASA Technical Reports Server (NTRS)

    Anderson, D. N.; Tacina, R. R.; Mroz, T. S.

    1975-01-01

    The performance of a catalytic reactor 12 cm in diameter and 17 cm long was evaluated at simulated gas turbine combustor operating conditions using premixed propane and air. Inlet temperatures of 600 and 800 K, pressures of 3 and 6 atm, and reference velocities of 9 to 30 m/s were tested. Data were taken for equivalence ratios as high as 0.43. The operating range was limited on the low-temperature side by very poor efficiency; the minimum exit temperature for good performance ranged from 1400 to 1600 K depending on inlet conditions. As exit temperatures were raised above this minimum, emissions of unburned hydrocarbons decreased, carbon monoxide emissions became generally less than 1 g CO/kg fuel, and nitrogen oxides were less than about 0.1 g NO2/kg fuel.

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  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. 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.

  5. Zero Power Reactor Simulation

    SciTech Connect

    2010-01-01

    Ever wanted to see a nuclear reactor core in action? Here's a detailed simulation of the Zero Power Reactor experiment, run by Argonne's unique "UNIC" code. Here, we use VisIt to visualize a numerical model of the ZPR-6 Assembly 6a experiment simulated using the Argonne UNIC code. 0:00-0:06: The fuel and other plates are dropped into an example drawer, and the drawer is inserted into the matrix tube. 0:06-0:21: The two matrix halves are brought together to make a critical (self-sustaining) assembly. 0:21-0:35: The fission power is revealed to be centralized in the thin, enriched Uranium plates. 0:35-0:53: Returning to our example drawer, we show the detailed local plate powers and their relation to the drawer composition. 0:53-1:11: Our model shows that each plate can have widely-varying local changes in the space and energy neutron densities. Read more at http://www.anl.gov/Media_Center/News/2010/news100121.html

  6. 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.

  7. Reduction of dioxin emission by a multi-layer reactor with bead-shaped activated carbon in simulated gas stream and real flue gas of a sinter plant.

    PubMed

    Hung, Pao Chen; Lo, Wei Chiao; Chi, Kai Hsien; Chang, Shu Hao; Chang, Moo Been

    2011-01-01

    A laboratory-scale multi-layer system was developed for the adsorption of PCDD/Fs from gas streams at various operating conditions, including gas flow rate, operating temperature and water vapor content. Excellent PCDD/F removal efficiency (>99.99%) was achieved with the multi-layer design with bead-shaped activated carbons (BACs). The PCDD/F removal efficiency achieved with the first layer adsorption bed decreased as the gas flow rate was increased due to the decrease of the gas retention time. The PCDD/F concentrations measured at the outlet of the third layer adsorption bed were all lower than 0.1 ng I-TEQ Nm⁻³. The PCDD/Fs desorbed from BAC were mainly lowly chlorinated congeners and the PCDD/F outlet concentrations increased as the operating temperature was increased. In addition, the results of pilot-scale experiment (real flue gases of an iron ore sintering plant) indicated that as the gas flow rate was controlled at 15 slpm, the removal efficiencies of PCDD/F congeners achieved with the multi-layer reactor with BAC were better than that in higher gas flow rate condition (20 slpm). Overall, the lab-scale and pilot-scale experiments indicated that PCDD/F removal achieved by multi-layer reactor with BAC strongly depended on the flow rate of the gas stream to be treated.

  8. 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.

  9. Gas-cooled nuclear reactor

    DOEpatents

    Peinado, Charles O.; Koutz, Stanley L.

    1985-01-01

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

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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…

  15. Dry desulfurization of simulated flue gas in a fluidized-bed reactor for a broad range of SO{sub 2} concentration and temperature: A comparison of models

    SciTech Connect

    Suyadal, Y.; Oguz, H.

    1999-08-01

    In this work, dry desulfurization of simulated flue gas was investigated in a batchwise operated laboratory-scale stainless steel fluidized-bed reactor (46 x 500 mm{sup 2}) by using calcium-containing local Turkish limestone (Karaagach/MUS) which was calcined at 900 C with 5% H{sub 2}O vapor. The sulfation reaction was carried out in a broad range of temperature (200 {le} T({degree}C) {le} 900) and SO{sub 2} feedstock concentration (1000 {le} C(ppm SO{sub 2}) {le} 6000). The experimental sulfation conversion-time data were tested according to unreacted shrinking core model (SCM), changing the grain size model (GM) and random pore model (RPM). It was found that the random pore model with control of product layer (CaSO{sub 3}/CaSO{sub 4}) diffusion described the experimental data best.

  16. Hydrogen production by reforming of simulated hot coke oven gas over nickel catalysts promoted with lanthanum and cerium in a membrane reactor

    SciTech Connect

    Hongwei Cheng; Xionggang Lu; Yuwen Zhang; Weizhong Ding

    2009-05-15

    Catalysts of Ni/Mg(Al)O promoted with lanthanum and cerium were tested in a BaCo{sub 0.7}Fe{sub 0.2}Nb{sub 0.1}O{sub 3{delta}} (BCFNO) membrane reactor by catalytic partial oxidation of simulated hot coke oven gas (COG) with toluene as a model tar compound under atmospheric pressure. Analysis of the catalysts suggested that the hydrotalcite precursor after thermal treatment lead to a good dispersion of nickel forming the solid solution NiO-MgO and spinel (Ni,Mg)Al{sub 2}O{sub 4}. The promoted catalysts had higher oxygen permeation flux, better catalytic activity, and better resistance to carbon formation, which will be promising catalysts in the catalytic partial oxidation reforming of hot COG. 29 refs., 11 figs., 2 tabs.

  17. High-Temperature Gas-Cooled Test Reactor Point Design

    SciTech Connect

    Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville; Gougar, Hans David; Kinsey, James Carl; Strydom, Gerhard; Kumar, Akansha

    2016-04-01

    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 UCO 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 technological readiness level, licensing approach and costs.

  18. 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.

  19. 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.

  20. Transport simulation for EBT reactor

    SciTech Connect

    Uckan, T.; Uckan, N.A.; Jaeger, E.F.

    1983-08-01

    Transport simulation and modeling studies for the ELMO Bumpy Torus (EBT) reactor are carried out by using zero-dimensional (0-D) and one-and-one-half-dimensional (1 1/2-D) transport calculations. The time-dependent 0-D model is used for global analysis, whereas the 1 1/2-D radial transport code is used for accurate determination of density, temperature, and ambipolar potential profiles and of the role of these profiles in reactor plasma performance. Analysis with the 1 1/2-D transport code shows that profile effects near the outer edge of the hot electron ring lead to enhanced confinement by at least a factor of 2 to 5 beyond the simple scaling that is obtained from the global analysis. The radial profiles of core plasma density and temperatures (or core pressure) obtained from 1 1/2-D transport calculations are found to be similar to those theoretically required for stability.

  1. Programming for a nuclear reactor instrument simulator

    SciTech Connect

    Cohn, C.E.

    1989-01-01

    A new computerized control system for a transient test reactor incorporates a simulator for pre-operational testing of control programs. The part of the simulator pertinent to the discussion here consists of two microprocessors. An 8086/8087 reactor simulator calculates simulated reactor power by solving the reactor kinetics equations. An 8086 instrument simulator takes the most recent power value developed by the reactor simulator and simulates the appropriate reading on each of the eleven reactor instruments. Since the system is required to run on a one millisecond cycle, careful programming was required to take care of all eleven instruments in that short time. This note describes the special programming techniques used to attain the needed performance.

  2. 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.

  3. 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.

  4. Effect of temperature on reduction of CaSO{sub 4} oxygen carrier in chemical-looping combustion of simulated coal gas in a fluidized bed reactor

    SciTech Connect

    Song, Q.L.; Xiao, R.; Deng, Z.Y.; Shen, L.H.; Xiao, J.; Zhang, M.Y.

    2008-12-15

    Chemical-looping combustion (CLC) is a promising combustion technology for gaseous and solid fuel with efficient use of energy and inherent separation of CO{sub 2}. The concept of a coal-fueled CLC system using, calcium sulfate (CaSO{sub 4}) as oxygen carrier is proposed in this study. Reduction tests of CaSO{sub 4} oxygen carrier with simulated coal gas were performed in a laboratory-scale fluidized bed reactor in the temperature range of 890-950{degree}C. A high concentration of CO{sub 2} was obtained at the initial reduction period. CaSO{sub 4} oxygen carrier exhibited high reactivity initially and decreased gradually at the late period of reduction. The sulfur release during the reduction of CaSO{sub 4} as oxygen carrier was also observed and analyzed. H{sub 2} and CO{sub 2} conversions were greatly influenced by reduction temperature. The oxygen carrier conversion and mass-based reaction rates during the reduction at typical temperatures were compared. Higher temperatures would enhance reaction rates and result in high conversion of oxygen carrier. An XRD patterns study indicated that CaS was the dominant product of reduction and the variation of relative intensity with temperature is in agreement with the solid conversion. ESEM analysis indicated that the surface structure of oxygen carrier particles changed significantly from impervious to porous after reduction. EDS analysis also demonstrated the transfer of oxygen from the oxygen carrier to the fuel gas and a certain amount of sulfur loss and CaO formation on the surface at higher temperatures. The reduction kinetics of CaSO{sub 4} oxygen carrier was explored with the shrinking unreacted-core model. The apparent kinetic parameters were obtained, and the kinetic equation well predicted the experimental data. Finally, some basic considerations on the use of CaSO{sub 4} oxygen carrier in a CLC system for solid fuels were discussed.

  5. Gas-cooled fast reactor fuel fabrication

    NASA Astrophysics Data System (ADS)

    Fielding, Randall; Meyer, Mitch; Jue, Jan-Fong; Gan, Jian

    2007-09-01

    The gas-cooled fast reactor is a high temperature helium-cooled Generation IV reactor concept. Operating parameters for this type of reactor are well beyond those of current fuels so a novel fuel must be developed. One fuel concept calls for UC particles dispersed throughout an SiC matrix. This study examines a hybrid reaction bonding process as a possible fabrication route for this fuel. Processing parameters are also optimized. The process combines carbon and SiC powders and a carbon yielding polymer. In order to obtain dense reaction bonded SiC samples the porosity to carbon ratio in the preform must be large enough to accommodate SiC formation from the carbon present in the sample, however too much porosity reduces mechanical integrity which leads to poor infiltration properties. The porosity must also be of a suitable size to allow silicon transport throughout the sample but keep residual silicon to a minimum.

  6. Gas-Fast Reactor Fuel Fabrication

    SciTech Connect

    Randall Fielding; Mitchell Meyer; Ramprashad Prabhakaran; Jim Miller; Sean McDeavitt

    2005-11-01

    The gas-cooled fast reactor is a high temperature helium cooled Generation IV reactor concept. Operating parameters for this type of reactor are well beyond those of current fuels so a novel fuel must be developed. One fuel concept calls for UC particles dispersed throughout a SiC matrix. This study examines a hybrid reaction bonding process as a possible fabrication route for this fuel. Processing parameters are also optimized. The process combines carbon and SiC powders and a carbon yielding polymer. In order to obtain dense reaction bonded SiC samples the porosity to carbon ratio in the preform must be large enough to accommodate SiC formation from the carbon present in the sample, however too much porosity reduces mechanical integrity which leads to poor infiltration properties . The porosity must also be of a suitable size to allow silicon transport throughout the sample but keep residual silicon to a minimum.

  7. 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.

  8. Metabolic modeling of synthesis gas fermentation in bubble column reactors.

    PubMed

    Chen, Jin; Gomez, Jose A; Höffner, Kai; Barton, Paul I; Henson, Michael A

    2015-01-01

    A promising route to renewable liquid fuels and chemicals is the fermentation of synthesis gas (syngas) streams to synthesize desired products such as ethanol and 2,3-butanediol. While commercial development of syngas fermentation technology is underway, an unmet need is the development of integrated metabolic and transport models for industrially relevant syngas bubble column reactors. We developed and evaluated a spatiotemporal metabolic model for bubble column reactors with the syngas fermenting bacterium Clostridium ljungdahlii as the microbial catalyst. Our modeling approach involved combining a genome-scale reconstruction of C. ljungdahlii metabolism with multiphase transport equations that govern convective and dispersive processes within the spatially varying column. The reactor model was spatially discretized to yield a large set of ordinary differential equations (ODEs) in time with embedded linear programs (LPs) and solved using the MATLAB based code DFBAlab. Simulations were performed to analyze the effects of important process and cellular parameters on key measures of reactor performance including ethanol titer, ethanol-to-acetate ratio, and CO and H2 conversions. Our computational study demonstrated that mathematical modeling provides a complementary tool to experimentation for understanding, predicting, and optimizing syngas fermentation reactors. These model predictions could guide future cellular and process engineering efforts aimed at alleviating bottlenecks to biochemical production in syngas bubble column reactors.

  9. Multicycle study on chemical-looping combustion of simulated coal gas with a CaSO{sub 4} oxygen carrier in a fluidized bed reactor

    SciTech Connect

    Qilei Song; Rui Xiao; Zhongyi Deng; Wenguang Zheng; Laihong Shen; Jun Xiao

    2008-11-15

    The cyclic test of a CaSO{sub 4}-based oxygen carrier (natural anhydrite) in alternating reducing simulated coal gas and oxidizing conditions was performed at 950{degree}C in a fluidized bed reactor at atmospheric pressure. A high concentration of CO{sub 2} was obtained in the reduction. The H{sub 2} and CO conversions and CO{sub 2} yield increased initially and final decreased significantly. The release of SO{sub 2} and H{sub 2}S during the cyclic test was found to be responsible for the decrease of reactivity of a CaSO{sub 4} oxygen carrier. The oxygen carrier conversion after the reduction reaction decreased gradually in the cyclic test. Through the comparison of mass-based reaction rates as a function of mass conversion at typical cycles, it was also evident that the reactivity of a CaSO{sub 4} oxygen carrier increased for the initial cycles but finally decreased after around 15 cycles. X-ray diffraction analysis revealed that the presence and intensity of the reduction sulfur species was in accordance with the results of gas conversion. The content of CaO was higher than expected, suggesting the formation of SO{sub 2} and H{sub 2}S during the cycles. Surface morphology analysis demonstrates that the natural anhydrite particle surface varied from impervious to porous after the cyclic test. It was also observed that the small grains on the surface of the oxygen carrier sintered in the cyclic tests. Energy-dispersive spectrum analysis also demonstrated the decrease of oxygen intensity after reduction, and CaO became the main component after the 20th oxidation. Pore structure analysis suggested that the particles agglomerated or sintered in the cyclic tests. The possible method for sulfur mitigation is proposed. Finally, some basic consideration on the design criteria of a CLC system for solid fuels using a CaSO{sub 4} oxygen carrier is discussed by the references and provides direction for future work. 49 refs., 10 figs., 5 tabs.

  10. 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.

  11. 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.

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

    NASA Technical Reports Server (NTRS)

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

    1973-01-01

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

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

    NASA Technical Reports Server (NTRS)

    Kunze, J. F.; Suckling, D. H.; Copper, C. G.

    1972-01-01

    Flow tests were conducted on models of the gas core (cavity) reactor. Variations in cavity wall and injection configurations 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. All tests were run in the down-firing direction so that gravitational effects simulated the acceleration effect of a rocket. Results show that acceptable flow patterns with high 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 along the cavity wall, using louvered or oblique-angle-honeycomb injection schemes.

  14. 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.

  15. 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.

  16. 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.

  17. Simulating industrial plasma reactors - A fresh perspective

    NASA Astrophysics Data System (ADS)

    Mohr, Sebastian; Rahimi, Sara; Tennyson, Jonathan; Ansell, Oliver; Patel, Jash

    2016-09-01

    A key goal of the presented research project PowerBase is to produce new integration schemes which enable the manufacturability of 3D integrated power smart systems with high precision TSV etched features. The necessary high aspect ratio etch is performed via the BOSCH process. Investigations in industrial research are often use trial and improvement experimental methods. Simulations provide an alternative way to study the influence of external parameters on the final product, whilst also giving insights into the physical processes. This presentation investigates the process of simulating an industrial ICP reactor used over high power (up to 2x5 kW) and pressure (up to 200 mTorr) ranges, analysing the specific procedures to achieve a compromise between physical correctness and computational speed, while testing commonly made assumptions. This includes, for example, the effect of different physical models and the inclusion of different gas phase and surface reactions with the aim of accurately predicting the dependence of surface rates and profiles on external parameters in SF6 and C4F8 discharges. This project has received funding from the Electronic Component Systems for European Leadership Joint Undertaking under Grant Agreement No. 662133 PowerBase.

  18. 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.

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

    SciTech Connect

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

    2016-03-01

    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.

  20. 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.

  1. 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.

  2. Two-dimensional simulation of an oxy-acetylene torch diamond reactor with a detailed gas-phase and surface mechanism

    NASA Astrophysics Data System (ADS)

    Okkerse, M.; Kleijn, C. R.; van den Akker, H. E. A.; de Croon, M. H. J. M.; Marin, G. B.

    2000-10-01

    A two-dimensional model is presented for the hydrodynamics and chemistry of an oxy-acetylene torch reactor for chemical vapor deposition of diamond, and it is validated against spectroscopy and growth rate data from the literature. The model combines the laminar equations for flow, heat, and mass transfer with combustion and deposition chemistries, and includes multicomponent diffusion and thermodiffusion. A two-step solution approach is used. In the first step, a lumped chemistry model is used to calculate the flame shape, temperatures and hydrodynamics. In the second step, a detailed, 27 species / 119 elementary reactions gas phase chemistry model and a 41 species / 67 elementary reactions surface chemistry model are used to calculate radicals and intermediates concentrations in the gas phase and at the surface, as well as growth rates. Important experimental trends are predicted correctly, but there are some discrepancies. The main problem lies in the use of the Miller-Melius hydrocarbon combustion mechanism for rich oxy-acetylene flames. [J. A. Miller and C. F. Melius, Combustion and Flame 91, 21 (1992)]. Despite this problem, some aspects of the diamond growth process are clarified. It is demonstrated that gas-phase diffusion limitations play a minor role in the diamond growth process, which is determined by surface kinetics. Except for atomic hydrogen, gas phase diffusion is also of minor importance for the transport of species in and behind the flame front. Finally, it is shown that penetration of nitrogen from the ambient air into the flame cannot explain the observed changes at the center of the diamond films as reported in the literature.

  3. Computational Flow Predictions for the Lower Plenum of a High-Temperature, Gas-Cooled Reactor

    SciTech Connect

    Donna Post Guillen

    2006-11-01

    Advanced gas-cooled reactors offer the potential advantage of higher efficiency and enhanced safety over present day nuclear reactors. Accurate simulation models of these Generation IV reactors are necessary for design and licensing. One design under consideration by the Very High Temperature Reactor (VHTR) program is a modular, prismatic gas-cooled reactor. In this reactor, the lower plenum region may experience locally high temperatures that can adversely impact the plant’s structural integrity. Since existing system analysis codes cannot capture the complex flow effects occurring in the lower plenum, computational fluid dynamics (CFD) codes are being employed to model these flows [1]. The goal of the present study is to validate the CFD calculations using experimental data.

  4. Computational Flow Predictions for the Lower Plenum of a High-Temperature, Gas-Cooled Reactor

    SciTech Connect

    Not Available

    2006-11-01

    Advanced gas-cooled reactors offer the potential advantage of higher efficiency and enhanced safety over present day nuclear reactors. Accurate simulation models of these Generation IV reactors are necessary for design and licensing. One design under consideration by the Very High Temperature Reactor (VHTR) program is a modular, prismatic gas-cooled reactor. In this reactor, the lower plenum region may experience locally high temperatures that can adversely impact the plant's structural integrity. Since existing system analysis codes cannot capture the complex flow effects occurring in the lower plenum, computational fluid dynamics (CFD) codes are being employed to model these flows [1]. The goal of the present study is to validate the CFD calculations using experimental data.

  5. High-temperature Gas Reactor (HTGR)

    NASA Astrophysics Data System (ADS)

    Abedi, Sajad

    2011-05-01

    General Atomics (GA) has over 35 years experience in prismatic block High-temperature Gas Reactor (HTGR) technology design. During this period, the design has recently involved into a modular have been performed to demonstrate its versatility. This versatility is directly related to refractory TRISO coated - particle fuel that can contain any type of fuel. This paper summarized GA's fuel cycle studies individually and compares each based upon its cycle sustainability, proliferation-resistance capabilities, and other performance data against pressurized water reactor (PWR) fuel cycle data. Fuel cycle studies LEU-NV;commercial HEU-Th;commercial LEU-Th;weapons-grade plutonium consumption; and burning of LWR waste including plutonium and minor actinides in the MHR. results show that all commercial MHR options, with the exception of HEU-TH, are more sustainable than a PWR fuel cycle. With LEU-NV being the most sustainable commercial options. In addition, all commercial MHR options out perform the PWR with regards to its proliferation-resistance, with thorium fuel cycle having the best proliferation-resistance characteristics.

  6. 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.

  7. Fuel development for gas-cooled fast reactors

    NASA Astrophysics Data System (ADS)

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

    2007-09-01

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

  8. 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.

  9. 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.

  10. 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.

  11. 89. ARAIII. Petrochem oilfired gas heater installed in reactor building ...

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

    89. ARA-III. Petro-chem oil-fired gas heater installed in reactor building (ARA-608). View is at floor level. Shows hand rails around heater pit and top of pit extending upwards through ceiling. January 20, 1959. Ineel photo no. 59-321. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. Status report on high fidelity reactor simulation.

    SciTech Connect

    Palmiotti, G.; Smith, M.; Rabiti, C.; Lewis, E.; Yang, W.; Leclere,M.; Siegel, A.; Fischer, P.; Kaushik, D.; Ragusa, J.; Lottes, J.; Smith, B.

    2006-12-11

    This report presents the effort under way at Argonne National Laboratory toward a comprehensive, integrated computational tool intended mainly for the high-fidelity simulation of sodium-cooled fast reactors. The main activities carried out involved neutronics, thermal hydraulics, coupling strategies, software architecture, and high-performance computing. A new neutronics code, UNIC, is being developed. The first phase involves the application of a spherical harmonics method to a general, unstructured three-dimensional mesh. The method also has been interfaced with a method of characteristics. The spherical harmonics equations were implemented in a stand-alone code that was then used to solve several benchmark problems. For thermal hydraulics, a computational fluid dynamics code called Nek5000, developed in the Mathematics and Computer Science Division for coupled hydrodynamics and heat transfer, has been applied to a single-pin, periodic cell in the wire-wrap geometry typical of advanced burner reactors. Numerical strategies for multiphysics coupling have been considered and higher-accuracy efficient methods proposed to finely simulate coupled neutronic/thermal-hydraulic reactor transients. Initial steps have been taken in order to couple UNIC and Nek5000, and simplified problems have been defined and solved for testing. Furthermore, we have begun developing a lightweight computational framework, based in part on carefully selected open source tools, to nonobtrusively and efficiently integrate the individual physics modules into a unified simulation tool.

  17. 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.

  18. 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.

  19. 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

  20. Hot gas cleanup for molten carbonate fuel cells: A zinc reactor model

    NASA Astrophysics Data System (ADS)

    Steinfeld, G.

    1980-09-01

    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 F, and the gasifier effluent could be between 1200 to 1900 F, a hot gas cleanup system will require little or no change in process gas temperature, thereby contributing to a high overall system efficiency. Simulated operating characteristics to aid in system design and system simulations of gasifier/MCFC systems are described. The modeling of the ZnO reactor is presented.

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

    PubMed

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

    2014-09-02

    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.

  2. 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.

  3. 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.

  4. 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.

  5. Solar coal gasification reactor with pyrolysis gas recycle

    SciTech Connect

    Aiman, W.R.; Gregg, D.W.

    1983-11-15

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

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

    NASA Technical Reports Server (NTRS)

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

    1982-01-01

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

  7. Development of hollow-fiber catalytic-membrane reactors for high-temperature gas cleanup

    SciTech Connect

    Ma, Yi H.; Moser, M.R.; Pien, S.M.

    1992-12-01

    The project consist of the following main activities: (1) Design of catalytic hollow fiber membrane reactors. Single and multiple hollow-fiber membranes were studied in reactor/permeation cells made from stainless steel or quartz tubes. Modification of the hollow fiber membrane with catalysts was performed by aqueous impregnation, vapor deposition, and utilization of packed-bed reactors. (2) Investigation of gas separations and catalytic reactions in membrane reactors. Permeation of pure gases and gas mixtures was studied as a function of temperature. Pure component catalytic studies on the decomposition of H{sub 2}S was typically studied using 10% H{sub 2}S diluted in He. The H{sub 2}S and H{sub 2} concentrations were measured in both the tube and shell sides of the membrane reactor to determine the degree of chemical equilibrium shift. (3) Process development of the cleanup system using a simulated gas stream with a composition similar to that from an IGCC system. Catalytic studies using the IGCC gas composition will be performed according to the procedure used in the H{sub 2}S experiments. The conditions for optimum conversion in a gas mixture will be investigated.

  8. Development of hollow-fiber catalytic-membrane reactors for high-temperature gas cleanup

    SciTech Connect

    Ma, Yi H.; Moser, M.R.; Pien, S.M.

    1992-01-01

    The project consist of the following main activities: (1) Design of catalytic hollow fiber membrane reactors. Single and multiple hollow-fiber membranes were studied in reactor/permeation cells made from stainless steel or quartz tubes. Modification of the hollow fiber membrane with catalysts was performed by aqueous impregnation, vapor deposition, and utilization of packed-bed reactors. (2) Investigation of gas separations and catalytic reactions in membrane reactors. Permeation of pure gases and gas mixtures was studied as a function of temperature. Pure component catalytic studies on the decomposition of H[sub 2]S was typically studied using 10% H[sub 2]S diluted in He. The H[sub 2]S and H[sub 2] concentrations were measured in both the tube and shell sides of the membrane reactor to determine the degree of chemical equilibrium shift. (3) Process development of the cleanup system using a simulated gas stream with a composition similar to that from an IGCC system. Catalytic studies using the IGCC gas composition will be performed according to the procedure used in the H[sub 2]S experiments. The conditions for optimum conversion in a gas mixture will be investigated.

  9. 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.

  10. Novel reactor configuration for synthesis gas conversion to alcohols

    SciTech Connect

    Akgerman, A.; Anthony, R.G. . Dept. of Chemical Engineering)

    1989-01-01

    Research continued on the conversion of synthesis gas to alcohols and reactor configuration. Objectives for this quarter: the project stated on October 1, 1989 and according to the Task Schedule provided in the original work breakdown schedule, Task I was to be completed in the first quarter and Task II to be started. Task I consisted of construction of the slurry reactor set-up to be used in Task IV for determination of the reactor kinetics and procurement of the parts for automation equipment, separators, computer activated parts etc. for automation of the trickle bed rector and GC equipment. Task II consisted of standardization and automation of GC analysis protocols. 1 fig.

  11. Flow Simulation and Optimization of Plasma Reactors for Coal Gasification

    NASA Astrophysics Data System (ADS)

    Ji, Chunjun; Zhang, Yingzi; Ma, Tengcai

    2003-10-01

    This paper reports a 3-d numerical simulation system to analyze the complicated flow in plasma reactors for coal gasification, which involve complex chemical reaction, two-phase flow and plasma effect. On the basis of analytic results, the distribution of the density, temperature and components' concentration are obtained and a different plasma reactor configuration is proposed to optimize the flow parameters. The numerical simulation results show an improved conversion ratio of the coal gasification. Different kinds of chemical reaction models are used to simulate the complex flow inside the reactor. It can be concluded that the numerical simulation system can be very useful for the design and optimization of the plasma reactor.

  12. 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.

  13. VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

    DOEpatents

    Furgerson, W.T.

    1963-12-17

    A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

  14. G2 Gas Cloud Simulation

    NASA Image and Video Library

    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...

  15. 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.

  16. 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.

  17. 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.

  18. Chemical Characterization of Simulated Boiling Water Reactor Coolant

    DTIC Science & Technology

    1990-05-01

    industry to reduce personnel radiation exposure and down-time associated with the operation, mainte- nance and refueling of Light Water Reactor (LWR...AD-A226 654 t t-FILL UIY C CHEMICAL CHARACTERIZATION OF SIMULATED , .BOILING WATER REACTOR COOLANt by Li . . , . , - VERRDON HOLBROOK MASON f ; B.S...CHARACTERIZATION OF SIMULATED BOILING WATER REACTOR COOLANT by VERRDON HOLBROOK MASON Submitted to the Department of Nuclear Engineering on May 9, 1988 in

  19. Modeling and simulation of silicon epitaxial growth in Siemens CVD reactor

    NASA Astrophysics Data System (ADS)

    Ni, Haoyin; Lu, Shijie; Chen, Caixia

    2014-10-01

    Siemens CVD reactor is an important chemical device for the production of polysilicon. The chemical and physical phenomenon involved in the reactor is very complex. Understanding the multispecies thermal fluid transport and its interaction with the gas/surface reactions is crucial for an optimal design and operation of the reactor. In the present paper, a mathematical model was constructed to describe the fluid dynamics, the heat and mass transfer and the reaction kinetics of the epitaxial growth process in industrial CVD reactors. A modified reaction kinetics model was used to represent the gas phase and surface reactions. The kinetics model was validated using the published experimental data obtained in a temperature range similar to the industrial CVD processes of silicon productions. The epitaxial growth of silicon in a Siemens reactor was simulated using commercial Computational Fluid Dynamics (CFD) software ANSYS FLUENT. The distributions of gas velocity, temperature and species concentrations in the reactor were predicted numerically. Based on the numerical simulation results, a sensitivity analysis was carried out to determine the key factors influencing the growth rate in industrial CVD reactors. Under the conditions of fixed heating power applied to three different rod diameters of 50 mm, 80 mm and 100 mm, the simulated results show, when the rods' diameter is 50 mm, the surface temperature is high and the gas temperature is low, the growth rate of silicon is determined by the transport of gas species. When the rods' diameter increases to 80 mm, the averaged surface temperature decreases to 1361 K, the surface reaction rate and transport of gas species control the growth rate of Si together. When the rods' diameter is 100 mm, the surface temperature decreases further, the rates of surface reactions become the control factor of deposition rate of Si.

  20. 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).

  1. Dynamic simulation platform to verify the performance of the reactor regulating system for a research reactor

    SciTech Connect

    2015-07-01

    Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of research reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)

  2. Passive modular gas safety system for a reactor

    SciTech Connect

    Abalin, S.S.; Isaev, I.F.; Kulakov, A.A.; Sivokon, V.P.; Udovenko, A.N.; Ionaitis, R.R.

    1994-01-01

    Reactor safety systems have developed gradually. Today in particular, auxiliary systems are being developed which are based on nontraditional operational concepts, by using gaseous neutron absorbers. The Scientific-Research and Design Institute of Power Technology (NIKIET) and the Institute of Nuclear Reactors, Kurchatov Institute Reactor Science Center (RNTs), have done preliminary development and experimental verification of separate elements of this system, in which helium is used as the absorber. This article presents a rapid passive safety system based on gaseous absorber, which is made as autonomous modules as the final stage of reactor safety. Its effectiveness is discussed by using an RBMK reactor as an example. As opposed to traditional active, systems, it does not require a functioning power supply and information signals from outside the reactors system, which makes it stable against unsanctioned actions by personnel, the influence of other systems, and also outside actions (sabotage and natural calamities which could destroy the the nuclear power plant structure). Because the gas safety system can operate instantaneously (0.1-0.3 sec), in principle, it can shut down the reactor even with fast-neutron runaway, where traditional safety systems are ineffective.

  3. Computer simulation of the NASA water vapor electrolysis reactor

    NASA Technical Reports Server (NTRS)

    Bloom, A. M.

    1974-01-01

    The water vapor electrolysis (WVE) reactor is a spacecraft waste reclamation system for extended-mission manned spacecraft. The WVE reactor's raw material is water, its product oxygen. A computer simulation of the WVE operational processes provided the data required for an optimal design of the WVE unit. The simulation process was implemented with the aid of a FORTRAN IV routine.

  4. 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.

  5. 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

  6. 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.

  7. 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.

  8. Internal transport barrier simulation with pellet injection in tokamak and helical reactor plasmas

    NASA Astrophysics Data System (ADS)

    Higashiyama, Y.; Yamazaki, K.; Garcia, J.; Arimoto, H.; Shoji, T.

    2008-07-01

    In the future fusion reactor, the plasma density peaking is important for the increase in the fusion power gain. The density control and the internal transport barrier (ITB) formation due to the pellet injection have been simulated in tokamak and helical reactors using the toroidal transport linkage code TOTAL. Firstly, the pellet injection simulation is carried out including the neutral gas shielding model and the mass relocation model in the TOTAL code, and the effectiveness of the high field side (HFS) pellet injection is clarified. Secondly, the ITB simulation with the pellet injection is carried out with the confinement improvement model based on the E×B shear effects, and it is found that the deep pellet penetration is helpful for the ITB formation as well as the plasma core fuelling in the reversed shear tokamak reactor, but the deep pellet penetration is not effective in the helical reactor.

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

    NASA Astrophysics Data System (ADS)

    Kedlaya, Aditya

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

  10. Bypass Flow Resistance in Prismatic Gas-Cooled Nuclear Reactors

    SciTech Connect

    McEligot, Donald M.; Johnson, Richard W.

    2016-12-20

    Available computational fluid dynamics (CFD) predictions of pressure distributions in the vertical bypass flow between blocks in a prismatic gas-cooled reactor (GCR) have been analyzed to deduce apparent friction factors and loss coefficients for systems and network codes. We performed calculations for vertical gap spacings "s" of 2, 6 and 10 mm, horizontal gaps between the blocks of two mm and two flow rates, giving a range of gap Reynolds numbers ReDh of about 40 to 5300. Laminar predictions of the fully-developed friction factor ffd were about three to ten per cent lower than the classical infinitely-wide channel In the entry region, the local apparent friction factor was slightly higher than the classic idealized case but the hydraulic entry length Lhy was approximately the same. The per cent reduction in flow resistance was greater than the per cent increase in flow area at the vertical corners of the blocks. The standard k-ϵ model was employed for flows expected to be turbulent. Its predictions of ffd and flow resistance were significantly higher than direct numerical simulations for the classic case; the value of Lhy was about thirty gap spacings. Initial quantitative information for entry coefficients and loss coefficients for the expansion-contraction junctions between blocks is also presented. Our study demonstrates how CFD predictions can be employed to provide integral quantities needed in systems and network codes.

  11. Bypass Flow Resistance in Prismatic Gas-Cooled Nuclear Reactors

    DOE PAGES

    McEligot, Donald M.; Johnson, Richard W.

    2016-12-20

    Available computational fluid dynamics (CFD) predictions of pressure distributions in the vertical bypass flow between blocks in a prismatic gas-cooled reactor (GCR) have been analyzed to deduce apparent friction factors and loss coefficients for systems and network codes. We performed calculations for vertical gap spacings "s" of 2, 6 and 10 mm, horizontal gaps between the blocks of two mm and two flow rates, giving a range of gap Reynolds numbers ReDh of about 40 to 5300. Laminar predictions of the fully-developed friction factor ffd were about three to ten per cent lower than the classical infinitely-wide channel In themore » entry region, the local apparent friction factor was slightly higher than the classic idealized case but the hydraulic entry length Lhy was approximately the same. The per cent reduction in flow resistance was greater than the per cent increase in flow area at the vertical corners of the blocks. The standard k-ϵ model was employed for flows expected to be turbulent. Its predictions of ffd and flow resistance were significantly higher than direct numerical simulations for the classic case; the value of Lhy was about thirty gap spacings. Initial quantitative information for entry coefficients and loss coefficients for the expansion-contraction junctions between blocks is also presented. Our study demonstrates how CFD predictions can be employed to provide integral quantities needed in systems and network codes.« less

  12. 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.

  13. Effect of accidental steam entry on gas-cooled fast reactor integral neutronics parameters

    SciTech Connect

    Bhattacharyya, S.K.; Morman, J.A.; Bucher, R.G.; Smith, D.M.; Robinson, W.R.; Bennett, E.F.

    1980-10-01

    A possible accident scenario in a gas-cooled fast reactor (GCFR) is the leakage of secondary steam into the core. A full-scale experimental study of the physics effects of such an accidental condition has been performed on the zero power reactor (ZPR)-9 critical facility at Argonne National Laboratory. The results of integral neutronics measurements performed on the simulated steam-flooded GCFR core are reported, and comparisons with corresponding results for the reference GCFR core presented. Results of calculations of these parameters with ENDF/B-IV nuclear data and standard design methods are also presented. 26 refs.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Long life catalytic membrane reactors for spontaneous conversion of natural gas to synthesis gas

    SciTech Connect

    Schwartz, M., White, J., Deych, S., Millard, J., Myers, M., Sammells, A.

    1997-10-01

    This program is focusing on the development of mixed ionic and electronic conducting materials based on the brown millerite structure for use in catalytic membrane reactors (CMRs). These CMRs are being evaluated for promoting the spontaneous and highly selective oxidative reforming of carbon dioxide / natural gas mixtures to synthesis gas.

  20. Medium-size high-temperature gas-cooled reactor

    SciTech Connect

    Peinado, C.O.; Koutz, S.L.

    1980-08-01

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760/sup 0/C (1400/sup 0/F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics (a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant) and engineered safety features (core auxiliary cooling, relief valve, and steam generator dump systems).

  1. 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)

  2. 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.

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

    SciTech Connect

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

    2014-08-06

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

  4. Study of carbon dioxide gas treatment based on equations of kinetics in plasma discharge reactor

    NASA Astrophysics Data System (ADS)

    Abedi-Varaki, Mehdi

    2017-08-01

    Carbon dioxide (CO2) as the primary greenhouse gas, is the main pollutant that is warming earth. CO2 is widely emitted through the cars, planes, power plants and other human activities that involve the burning of fossil fuels (coal, natural gas and oil). Thus, there is a need to develop some method to reduce CO2 emission. To this end, this study investigates the behavior of CO2 in dielectric barrier discharge (DBD) plasma reactor. The behavior of different species and their reaction rates are studied using a zero-dimensional model based on equations of kinetics inside plasma reactor. The results show that the plasma reactor has an effective reduction on the CO2 density inside the reactor. As a result of reduction in the temporal variations of reaction rate, the speed of chemical reactions for CO2 decreases and very low concentration of CO2 molecules inside the plasma reactor is generated. The obtained results are compared with the existing experimental and simulation findings in the literature.

  5. 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.

  6. 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.

  7. 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.

  8. Development of hollow fiber catalytic membrane reactors for high temperature gas cleanup

    SciTech Connect

    Ma, Y.H.; Moser, W.R.; Pien, S.; Shelekhin, A.B.

    1994-10-01

    The technology employed in the Integrated Gasification Combined Cycle (IGCC) permits burning coals with a wide range of sulfur concentrations. Emissions from the process should be reduced by an order of magnitude below stringent federal air quality regulations for coal-fired plants. The maximum thermal efficiency of this type of process can be achieved by removing sulfur and particulates from the high temperature gas. The objective of this project was to develop economically and technically viable catalytic membrane reactors for high temperature, high pressure gaseous contaminant control in IGCC systems. These catalytic membrane reactors were used to decompose H{sub 2}S and separate the reaction products. The reactors were designed to operate in the hostile process environment of the IGCC systems, and at temperatures ranging from 500 to 1,000. Feasibility of the membrane reactor process for decomposition of hydrogen sulfide was demonstrated; permeability and selectivity of molecular-sieve and Vycor glass membranes were studied at temperatures up to 1,000 C; experimental study of hydrogen sulfide in the membrane reactor was completed; and a generalized mathematical model was developed for the simulation of the high temperature membrane reactor.

  9. Improvement and modelling of hexenal transfer in liquid-gas reactor.

    PubMed

    Ben Akacha, Najla; Guizani, Om Elkhir; Gargouri, Mohamed

    2007-12-01

    The aim of the present work is to improve the extraction of an aromatic compound, 2E-hexenal, from a continuous liquid-gas reactor. Having an improved process to recover hexenal could be of interest to obtain this chemical if produced by any biotechnological process. The experimental program proposed on the basis of a full 23 factorial design demonstrated that the conditions optima for the extraction are 40 degrees C as a reactor temperature and -10 degrees C as a temperature of trap for a nitrogen flow of 45 mL s-1. These conditions allowed to recover more than 75% of hexenal, when this compound was stripping by a nitrogen stream from the liquid phase. The mass transfer was theoretically and experimentally studied. The model predictions were validated against experimental results obtained for a reactor, and good correlation was observed for a simulation.

  10. 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.

  11. 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.

  12. 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.}

  13. 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.

  14. 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.

  15. 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.

  16. Moving bed reactor setup to study complex gas-solid reactions.

    PubMed

    Gupta, Puneet; Velazquez-Vargas, Luis G; Valentine, Charles; Fan, Liang-Shih

    2007-08-01

    A moving bed scale reactor setup for studying complex gas-solid reactions has been designed in order to obtain kinetic data for scale-up purpose. In this bench scale reactor setup, gas and solid reactants can be contacted in a cocurrent and countercurrent manner at high temperatures. Gas and solid sampling can be performed through the reactor bed with their composition profiles determined at steady state. The reactor setup can be used to evaluate and corroborate model parameters accounting for intrinsic reaction rates in both simple and complex gas-solid reaction systems. The moving bed design allows experimentation over a variety of gas and solid compositions in a single experiment unlike differential bed reactors where the gas composition is usually fixed. The data obtained from the reactor can also be used for direct scale-up of designs for moving bed reactors.

  17. 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.

  18. High-Temperature Gas-cooled Reactor steam-cycle/cogeneration lead plant reactor vessel: system design description

    SciTech Connect

    Not Available

    1983-01-01

    The Reactor Vessel System contains the primary coolant inventory within a gas-tight pressure boundary, and provides the necessary flow paths and overpressure protection for this pressure boundary. The Reactor Vessel System also houses the components of the Reactor System, the Heat Transport System, and the Auxiliary Heat Removal System. The scope of the Reactor Vessel System includes the prestressed concrete reactor vessel (PCRV) structure with its reinforcing steel and prestressing components; liners, penetrations, closures, and cooling water tubes attached to the concrete side of the liner; the thermal barrier (insulation) on the primary coolant side of the liner; instrumentation for structural monitoring; and a pressure relief system. Specifications are presented.

  19. Numerical investigation of heat transfer in high-temperature gas-cooled reactors

    SciTech Connect

    Chen, g.; Anghaie, S.

    1995-09-01

    This paper proposes a computational model for analysis of flow and heat transfer in high-temperature gas-cooled reactors. The formulation of the problem is based on using the axisymmetric, thin layer Navier-Stokes equations. A hybrid implicit-explicit method based on finite volume approach is used to numerically solve the governing equations. A fast converging scheme is developed to accelerate the Gauss-Siedel iterative method for problems involving the wall heat flux boundary condition. Several cases are simulated and results of temperature and pressure distribution in the core are presented. Results of a parametric analysis for the assessment of the impact of power density on the convective heat transfer rate and wall temperature are discussed. A comparative analysis is conducted to identify the Nusselt number correlation that best fits the physical conditions of the high-temperature gas-cooled reactors.

  20. Safety aspects of modular high temperature gas cooled reactors

    SciTech Connect

    Kasten, P.R.

    1986-05-14

    Modular high temperature gas cooled reactors (MHTGRs) are being developed which have a high degree of passive safety. The advantageous safety characteristics are due to: the high heat capacity of the graphite core; the high temperature capability of the core components; the chemical stability and inertness of the fuel, coolant, and moderator; the high retention of fission products by the fuel coating; the single-phase characteristics of helium coolant; the inherent negative temperature coefficient of reactivity of the core; the defense in depth against air and water ingress events; and the ability to cool the reactor by passive heat transfer mechanisms following postulated accidents. Even for a number of accidents having very low probabilities of occurrence, the cumulative fission product release at the site boundary is estimated to be below those acceptable under the Protective Action Guidelines.

  1. 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.

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

    SciTech Connect

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

    2014-04-01

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

  3. Improved Pyrolysis Micro reactor Design via Computational Fluid Dynamics Simulations

    DTIC Science & Technology

    2017-05-23

    NUMBER (Include area code) 23 May 2017 Briefing Charts 25 April 2017 - 23 May 2017 Improved Pyrolysis Micro-reactor Design via Computational Fluid... PYROLYSIS MICRO-REACTOR DESIGN VIA COMPUTATIONAL FLUID DYNAMICS SIMULATIONS Ghanshyam L. Vaghjiani* DISTRIBUTION A: Approved for public release...Approved for public release, distribution unlimited. PA Clearance 17247 Chen-Source (>240 references from SciFinder as of 5/1/17): Flash pyrolysis

  4. A study on naphtha catalytic reforming reactor simulation and analysis

    PubMed Central

    Liang, Ke-min; Guo, Hai-yan; Pan, Shi-wei

    2005-01-01

    A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data. PMID:15909350

  5. A study on naphtha catalytic reforming reactor simulation and analysis.

    PubMed

    Liang, Ke-min; Guo, Hai-yan; Pan, Shi-wei

    2005-06-01

    A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data.

  6. CFD modeling of high temperature gas cooled reactors

    SciTech Connect

    Janse van Rensburg, J.J.; Viljoen, C.; Van Staden, M.P.

    2006-07-01

    This paper presents an overview of how CFD has been applied to model the gas flow and heat transfer within the PBMR (Pebble Bed Modular reactor) with the aim of providing valuable design and safety information. The thermo-hydraulic calculations are performed using the STAR-CD [1] Computational Fluid Dynamics (CFD) code and the neutronic calculations are performed using VSOP [2]. Results are presented for steady-state normal operation and for a transient De-pressurized Loss Of Forced Cooling event (DLOFC). (authors)

  7. 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.

  8. 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.

  9. WATER-GAS SHIFT KINETICS OVER IRON OXIDE CATALYSTS AT MEMBRANE REACTOR CONDITIONS

    SciTech Connect

    Carl R.F. Lund

    2002-08-02

    The kinetics of water-gas shift were studied over ferrochrome catalysts under conditions with high carbon dioxide partial pressures, such as would be expected in a membrane reactor. The catalyst activity is inhibited by increasing carbon dioxide partial pressure. A microkinetic model of the reaction kinetics was developed. The model indicated that catalyst performance could be improved by decreasing the strength of surface oxygen bonds. Literature data indicated that adding either ceria or copper to the catalyst as a promoter might impart this desired effect. Ceria-promoted ferrochrome catalysts did not perform any better than unpromoted catalyst at the conditions tested, but copper-promoted ferrochrome catalysts did offer an improvement over the base ferrochrome material. A different class of water-gas shift catalyst, sulfided CoMo/Al{sub 2}O{sub 3} is not affected by carbon dioxide and may be a good alternative to the ferrochrome system, provided other constraints, notably the requisite sulfur level and maximum temperature, are not too limiting. A model was developed for an adiabatic, high-temperature water-gas shift membrane reactor. Simulation results indicate that an excess of steam in the feed (three moles of water per mole of CO) is beneficial even in a membrane reactor as it reduces the rate of adiabatic temperature rise. The simulations also indicate that much greater improvement can be attained by improving the catalyst as opposed to improving the membrane. Further, eliminating the inhibition by carbon dioxide will have a greater impact than will increasing the catalyst activity (assuming inhibition is still operative). Follow-up research into the use of sulfide catalysts with continued kinetic and reactor modeling is suggested.

  10. 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.

  11. Biofilm reactors for treatment of gas streams containing chlorinated solvents

    SciTech Connect

    Speitel, G.E. Jr. . Dept. of Civil Engineering); McLay, D.S. )

    1993-08-01

    Air streams contaminated with chlorinated solvents are increasingly common as by-products of air-stripping and soil-vapor-extraction operations. This research investigated treatment of such gas streams with a bioreactor that supported the growth of methanotrophic bacteria. These bacteria cometabolize many chlorinated solvents. Trichloroethylene (TCE) and 1,2-dichloroethane (DCA) were selected as model contaminants. Removals ranged from 20% to 80% at influent concentrations of 300-1,000 [mu]g/L of air and packed-bed gas-resident times of 5-12 min. Biofilm models were able to describe bioreactor performance well. Pseudo-first-order rate constants from reactor modeling were considerably smaller than those measured in batch systems, suggesting that much of the biofilm was inactive, which also was supported by methane-removal data. Enzyme competition between methane and chlorinated solvents and toxicity from chlorinated solvent metabolites both appeared significant.

  12. Computational fluid dynamic simulations of chemical looping fuel reactors utilizing gaseous fuels

    SciTech Connect

    Mahalatkar, Kartikeya; Kuhlman, John; Huckaby, E. David; O'Brien, Thomas

    2011-02-01

    A computational fluid dynamic (CFD) model for the fuel reactor of chemical looping combustion technology has been developed, with special focus on accurately representing the heterogeneous chemical reactions. A continuum two-fluid model was used to describe both the gas and solid phases. Detailed sub-models to account for fluid–particle and particle–particle interaction forces were also incorporated. Two experimental cases were analyzed in this study (Son and Kim, 2006; Mattison et al., 2001). Simulations were carried out to test the capability of the CFD model to capture changes in outlet gas concentrations with changes in number of parameters such as superficial velocity, metal oxide concentration, reactor temperature, etc. For the experiments of Mattisson et al. (2001), detailed time varying outlet concentration values were compared, and it was found that CFD simulations provided a reasonable match with this data.

  13. 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.

  14. In-reactor testing of the closed cycle gas core reactor: The Nuclear Light Bulb concept

    NASA Astrophysics Data System (ADS)

    Gauntt, R. O.; Slutz, S. A.; Harms, G. A.; Latham, T. S.; Roman, W. C.; Rodgers, R. J.

    1992-10-01

    The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (greater than 1800 s) and thrust (greater than 445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (greater than 4000 K). Analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept are described. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRE. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRE for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept.

  15. 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.

  16. 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.

  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. 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)

  19. 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

  20. 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... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with... pressurized water nuclear power reactor with an operating license on October 16, 2003, except for...

  1. 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)

  2. 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... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with... pressurized water nuclear power reactor with an operating license on October 16, 2003, except for those...

  3. 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.

  4. 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.

  5. MORECA-GT: Interactive simulator for gas-turbine modular HTGR transients and heatup accidents with ATWS options

    SciTech Connect

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

    1994-03-01

    An interactive simulation code for studying postulated heatup accidents in modular high-temperature gas-cooled reactors (MHTGRs) has been adapted to assist with parametric design studies of the US Department of Energy`s (DOE`s) direct-cycle gas-turbine MHTGR concept. The studies show that the proposed MHTGR designs are very robust and can generally withstand the consequences of extremely low probability accidents with little or no damage to the reactor`s fuel or metallic components.

  6. Design of virtual SCADA simulation system for pressurized water reactor

    SciTech Connect

    Wijaksono, Umar Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-08

    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.

  7. Multiscale Simulations of ALD in Cross Flow Reactors

    DOE PAGES

    Yanguas-Gil, Angel; Libera, Joseph A.; Elam, Jeffrey W.

    2014-08-13

    In this study, we have developed a multiscale simulation code that allows us to study the impact of surface chemistry on the coating of large area substrates with high surface area/high aspect-ratio features. Our code, based on open-source libraries, takes advantage of the ALD surface chemistry to achieve an extremely efficient two-way coupling between reactor and feature length scales, and it can provide simulated quartz crystal microbalance and mass spectrometry data at any point of the reactor. By combining experimental surface characterization with simple analysis of growth profiles in a tubular cross flow reactor, we are able to extract amore » minimal set of reactions to effectively model the surface chemistry, including the presence of spurious CVD, to evaluate the impact of surface chemistry on the coating of large, high surface area substrates.« less

  8. Design of virtual SCADA simulation system for pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  9. Multiscale Simulations of ALD in Cross Flow Reactors

    SciTech Connect

    Yanguas-Gil, Angel; Libera, Joseph A.; Elam, Jeffrey W.

    2014-08-13

    In this study, we have developed a multiscale simulation code that allows us to study the impact of surface chemistry on the coating of large area substrates with high surface area/high aspect-ratio features. Our code, based on open-source libraries, takes advantage of the ALD surface chemistry to achieve an extremely efficient two-way coupling between reactor and feature length scales, and it can provide simulated quartz crystal microbalance and mass spectrometry data at any point of the reactor. By combining experimental surface characterization with simple analysis of growth profiles in a tubular cross flow reactor, we are able to extract a minimal set of reactions to effectively model the surface chemistry, including the presence of spurious CVD, to evaluate the impact of surface chemistry on the coating of large, high surface area substrates.

  10. Preparation macroconstants to simulate the core of VVER-1000 reactor

    NASA Astrophysics Data System (ADS)

    Seleznev, V. Y.

    2017-01-01

    Dynamic model is used in simulators of VVER-1000 reactor for training of operating staff and students. As a code for the simulation of neutron-physical characteristics is used DYNCO code that allows you to perform calculations of stationary, transient and emergency processes in real time to a different geometry of the reactor lattices [1]. To perform calculations using this code, you need to prepare macroconstants for each FA. One way of getting macroconstants is to use the WIMS code, which is based on the use of its own 69-group macroconstants library. This paper presents the results of calculations of FA obtained by the WIMS code for VVER-1000 reactor with different parameters of fuel and coolant, as well as the method of selection of energy groups for further calculation macroconstants.

  11. 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.

  12. 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)

  13. INITIAL 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

    2007-09-01

    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. 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 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.

  14. VERA Core Simulator methodology for pressurized water reactor cycle depletion

    DOE PAGES

    Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane; ...

    2017-01-12

    This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

  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. Thermal radiation in gas core nuclear reactors for space propulsion

    NASA Astrophysics Data System (ADS)

    Slutz, Stephen A.; Gauntt, Randall O.; Harms, Gary A.; Latham, Thomas; Roman, Ward; Rodgers, Richard J.

    1994-05-01

    A diffusive model of the radial transport of thermal radiation out of a cylindrical core of fissioning plasma is presented. The diffusion approximation is appropriate because the opacity of uranium is very high at the temperatures of interest (greater than 3000 K). We make one additional simplification of assuming constant opacity throughout the fuel. This allows the complete set of solutions to be expressed as a single function. This function is approximated analytically to facilitate parametric studies of the performance of a test module of the nuclear light bulb gas-core nuclear-rocket-engine concept, in the Annular Core Research Reactor at Sandia National Laboratories. Our findings indicate that radiation temperatures in range of 4000-6000 K are attainable, which is sufficient to test the high specific impulse potential (approximately 2000 s) of this concept.

  17. The Gas Turbine - Modular Helium Reactor: A Promising Option for Near Term Deployment

    SciTech Connect

    LaBar, Malcolm P.

    2002-07-01

    The Gas Turbine - Modular Helium Reactor (GT-MHR) is an advanced nuclear power system that offers unparalleled safety, high thermal efficiency, environmental advantages, and competitive electricity generation costs. The GT-MHR module couples a gas-cooled modular helium reactor (MHR) with a high efficiency modular Brayton cycle gas turbine (GT) energy conversion system. The reactor and power conversion systems are located in a below grade concrete silo that provides protection against sabotage. The GT-MHR safety is achieved through a combination of inherent safety characteristics and design selections that take maximum advantage of the gas-cooled reactor coated particle fuel, helium coolant and graphite moderator. The GT-MHR is projected to be economically competitive with alternative electricity generation technologies due to the high operating temperature of the gas-cooled reactor, high thermal efficiency of the Brayton cycle power conversion system, high fuel burnup (>100,000 MWd/MT), and low operation and maintenance requirements. (author)

  18. 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.

  19. 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

  20. Modeling and simulation of CANDU reactor and its regulating system

    NASA Astrophysics Data System (ADS)

    Javidnia, Hooman

    Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different

  1. Simulation of the Fuel Reactor of a Coal-Fired Chemical Looping Combustor

    NASA Astrophysics Data System (ADS)

    Mahalatkar, Kartikeya; O'Brien, Thomas; Huckaby, E. David; Kuhlman, John

    2009-06-01

    removed from the bed and returned to the air reactor without any accompanying unburned fuel. This paper presents a simulation of the gas-particle granular flow, with heat transfer and chemical reactions, in the FR. Accurate simulation of the segregation processes, depending on particle density and size differences between the carrier and the fuel, allows the design of a reactor with the desired behavior.

  2. 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.

  3. 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

  4. 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.

  5. 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.

  6. Level tracking in detailed reactor simulations

    SciTech Connect

    Aktas, B.; Mahaffy, J.H.

    1995-09-01

    We introduce a useful test problem for judging the performance of reactor safety codes in situations where moving two-phase mixture levels are present. The test problem tracks a two-phase liquid level as it rises and then falls back to its original position. Pure air exists above the level, and a low void air-water mixture is below the level. Conditions are subcooled and isothermal to remove complications resulting from failures of interfacial heat transfer packages to properly account for the level. Comparisons are made between the performance of current versions of CATHARE, RELAP5, TRAC-BF1, and TRAC-PF1. These system codes are based on finite-difference methods with a fixed, Eulerian staggered grid in space. When a partially filled cell with a mixture level discontinuity becomes the donor cell, the sharp changes in fluid properties across the interface results in numerical oscillations of various terms. Furthermore, the cell-to-cell convection of mass, momentum and energy are inaccurately predicted nearby a mixture level. To adequately model moving mixture levels, an efficient discontinuity tracking method for the finite-difference Eulerian approximations is described. This model had been implemented in the TRAC-BWR code for the two-phase mixture level tracking since the TRAC-BD1 Version (released April 1984). The result of the test problem run by the current version of TRAC-BF1/MOD1 with the mixture level tracking model shows some peculiar behavior of the variables such as velocities, pressures and interfacial terms. A systematic approach to improving performance of the tracking method is described. Implementing this approach in TRAC-BF1/MOD1 has shown a major improvement in the results.

  7. 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.

  8. 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.

  9. 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)

  10. 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.

  11. 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.

  12. 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 ...

  13. 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 ...

  14. 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.

  15. Gas phase depletion and flow dynamics in horizontal MOCVD reactors

    NASA Astrophysics Data System (ADS)

    Van de Ven, J.; Rutten, G. M. J.; Raaijmakers, M. J.; Giling, L. J.

    1986-08-01

    Growth rates of GaAs in the MOCVD process have been studied as a function of both lateral and axial position in horizontal reactor cells with rectangular cross-sections. A model to describe growth rates in laminar flow systems on the basis of concentration profiles under diffusion controlled conditions has been developed. The derivation of the growth rate equations includes the definition of an entrance length for the concentration profile to developed. In this region, growth rates appear to decrease with the 1/3 power of the axial position. Beyond this region, an exponential decrease is found. For low Rayleigh number conditions, the present experimental results show a very satisfactory agreement with the model without parameter fitting for both rectangular and tapered cells, and with both H 2 and N 2 as carrier gases. Theory also predicts that uniform deposition can be obtained over large areas in the flow direction for tapered cells, which has indeed been achieved experimentally. The influence of top-cooling in the present MOCVD system has been considered in more detail. From the experimental results, conclusions could be drawn concerning the flow characteristics. For low Rayleigh numbers (present study ≲ 700) it follows that growth rate distributions correspond with forced laminar flow characteristics. For relatively high Rayleigh numbers (present work 1700-2800), free convective effects with vortex formation are important. These conclusions are not specific for the present system, but apply to horizontal cold-wall reactors in general. On the basis of the present observations, recommendations for a cell design to obtain large area homogeneous deposition have been formulated. In addition, this work supports the conclusion that the final decomposition of trimethylgallium in the MOCVD process mainly takes place at the hot substrate and susceptor and not in the gas phase.

  16. 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.

  17. 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.

  18. 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

  19. Assessment of RELAP5-3D for Analysis of Very High Temperature Gas-Cooled Reactors

    SciTech Connect

    Chang Oh; Larry Siefken; Cliff Davis

    2005-10-01

    The RELAP5-3D© computer code is being improved for the analysis of very high temperature gas-cooled reactors. Diffusion and natural circulation can be important phenomena in gas-cooled reactors following a loss-of-coolant accident. Recent improvements to the code include the addition of models that simulate pressure loss across a pebble bed and molecular diffusion. These models were assessed using experimental data. The diffusion model was assessed using data from inverted U-tube experiments. The code’s capability to simulate natural circulation of air through a pebble bed was assessed using data from the NACOK facility. The calculated results were in reasonable agreement with the measured values.

  20. Component and system simulation models for High Flux Isotope Reactor

    SciTech Connect

    Sozer, A.

    1989-08-01

    Component models for the High Flux Isotope Reactor (HFIR) have been developed. The models are HFIR core, heat exchangers, pressurizer pumps, circulation pumps, letdown valves, primary head tank, generic transport delay (pipes), system pressure, loop pressure-flow balance, and decay heat. The models were written in FORTRAN and can be run on different computers, including IBM PCs, as they do not use any specific simulation languages such as ACSL or CSMP. 14 refs., 13 figs.

  1. Slurry phase synthesis of dimethyl ether from syngas -- A reactor model simulation

    SciTech Connect

    Mizuguchi, Masatsugu; Ogawa, Takashi; Ono, Masami,; Tomura, Keiji; Shikada, Tsutomu; Ohno, Yotaro; Fujimoto, Kaoru

    1998-12-31

    Dimethyl ether (DME) would be an attractive alternative fuel for diesel, domestic use, and power generation, if it is economically synthesized directly from syngas (derived from coal gasification or natural gas reforming). DME, which is a colorless gas with a boiling point of {minus}25 C, is chemically stable and easily liquefied under pressure. Since the properties of DME are similar to LPG, it can be handled and stored with the same manner as LPG. The authors have performed the slurry phase DME synthesis by using the 50 kg/day bench-scale unit. DME was synthesized at high yield from syngas (H{sub 2}+CO) with the newly developed catalyst system. To establish the scale-up methodology, the reactor simulation technique is essential. The authors developed a mathematical model of the slurry phase bubble column reactor for DME synthesis, which is based on their experimental results. The performance of a commercial-scale DME reactor was simulated by this model, and the results were discussed.

  2. Model biases in high-burnup fast reactor simulations

    SciTech Connect

    Touran, N.; Cheatham, J.; Petroski, R.

    2012-07-01

    A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

  3. A model for simulating autoclave-reactor pressure histories

    SciTech Connect

    Thorsness, C.B.

    1995-11-01

    Small heated-batch reactors, frequently referred to as autoclave reactors, are often used in developing information for a proposed new chemical/physical processing step. These systems often entail considerable pressure buildup during the course of operation. This report describes a model formulated to simulate well mixed autoclave reactors. The model solves a system of differential and algebraic equations which describe vapor/liquid equilibrium and chemical reactions in the reactor during a heating and cooling cycle. The model allows any number of chemical species to be defined. Phase equilibrium considerations are limited to systems with one liquid and one vapor phase, although some provisions for dealing with a second pure water liquid phase are considered. Equilibrium constraints are formulated using fugacity and activity coefficients. A new version of the general purpose differential-algebraic system solver DASSL, called DASPK, has been used to solve the system of equations. This solver has been found to work well in test problems. Selected results from a number of example problems are described. The example systems are water/nitrogen; crude oil/water; hexane/toluene; hexane/heptadecane; water/carbon dioxide; and a biomass system.

  4. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    SciTech Connect

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-09-01

    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 an outlet temperature of 850ºC 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 participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05).

  5. 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.

  6. An innovative gas-solid torbed reactor for the recycling industries

    NASA Astrophysics Data System (ADS)

    Dodson, C. E.; Lakshmanan, V. I.

    1998-07-01

    Gas-solid Torbed reactors have been developed for processing a wide range of materials. The reactors have facilitated several novel recycling projects in countries where they are used. One of the major advantages of these reactors when applied to the recycling industries is the compactness of the plant and its inherent ability to be scaled down and fully automated to better match the volume requirements of this sector.

  7. Simulant Gas Test Technique Feasibility

    DTIC Science & Technology

    1990-05-01

    constant y shock pressure. The test time is evaluated conservatively from the difference in arrival times at the expansion nozzle between the incident...relation for a constant y. The value of y at the supply gas conditions is used in this expression and a rapidly convergent Newton-Raphson iteration method...dimensional flow property/Mach number relations with assumed constant y and from the gas mixture equation of state. In all of the calculations in the present

  8. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    SciTech Connect

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven 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 experiments 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 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 several independent 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 was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  9. 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.

  10. 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.

  11. 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.

  12. ORCULT-I: a loop dynamics simulator program for the core flow test loop. [GCFR reactor simulator

    SciTech Connect

    Ball, S.J.

    1980-02-01

    A digital simulation program, ORCULT-I, was developed to predict both the steady-state and transient behavior of the core flow test loop (CFTL). ORCULT has been used to answer design questions concerning loop configurations, operating modes, bundle tests, and postulated loop-accident conditions. The CFTL is a thermal-hydraulic and structural test vehicle for performance evaluations of Gas-Cooled Fast-Breeder Reactor (GCFR) fuel-rod bundle designs. The status, capabilities, and limitations of ORCULT are described, and several sample transients are shown which demonstrate characteristics of CFTL behavior. The appendix of the report includes derivations of the model equations used in ORCULT.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Gas Flow Simulation in GCB Chambers Featuring Hot Gas Energy

    NASA Astrophysics Data System (ADS)

    Mori, Tadashi; Iwamoto, Katsuharu; Kawano, Hiromichi; Tanaka, Yasunori

    A hot gas simulation in self-blast type GCBs was performed and revised. As a result, it was foundto be important to consider the thermodynamic and transport properties of SF6-PTFE gas at high pressures and temperatures. Moreover, modelling arc diameters are also important in the chamber. They have become the main factors in determining the rise of puffer pressure in self-blast type GCBs, while they were not relevant in conventional puffer-type GCBs.

  18. Results of field trials using the NPL simulated reactor neutron field facility.

    PubMed

    Taylor, G C; Thomas, D J; Bennett, A

    2007-01-01

    The NPL simulated reactor neutron field facility provides neutron spectra similar to those found in the environs of UK gas-cooled reactors. Neutrons are generated by irradiating a thick lithium-alloy target with monoenergetic protons between 2.5 and 3.5 MeV (depending on the desired spectrum), and then moderated by a 40-cm diameter sphere of heavy water. This represents an extremely soft workplace field, with a mean neutron energy of 25 keV and, more significantly, a mean fluence to ambient dose equivalent conversion coefficient of the order of 20 pSv cm(2), approximately 20 times lower than those of the ISO standard calibration sources (252)Cf and (241)Am-Be. Results of field trials are presented, including readings from neutron spectrometers, personal dosimeters (active and passive) and neutron area survey meters, and issues with beam monitoring are discussed.

  19. 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.

  20. 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.

  1. GCFR: The European Union Gas Cooled Fast Reactor Project

    SciTech Connect

    Mitchell, Colin; Peers, Karen; Poette, Christian; Coddington, Paul; Somers, Joe; Van-Goethem, George

    2006-07-01

    In March 2005, the European Commission (EC) initiated a new 4-year Project on Gas Cooled Fast Reactors (GCFR) within its 6. Framework Programme. The EC and more than 10 participating companies, R and D organizations and universities finance the project in equal parts. The project contributes to the Generation IV ambitious goals requiring innovative solutions in terms environmental impact (robust fuel with no significant radioactive release), sustainability (core which is self sustaining and has the flexibility for waste reduction), proliferation resistant fuel cycle and economics (high coolant temperatures leading to increased thermodynamic efficiency). A matrix has been prepared for the Generation IV GFR studies to facilitate sharing the work between the members, which identifies seven combinations of design options. These option studies will lead to a pre-selection of a reference concept and alternatives and the preliminary GFR viability report. The GCFR project, which forms part of the EURATOM contribution to the Generation IV International Forum (GIF) has responsibility for the direct cycle and indirect cycle 600 MW options. In detail, the GCFR project will examine; the GFR (600 MW options) and ETDR, core and system design; GFR and ETDR safety analysis, including the analysis of selected transients; the qualification and benchmarking of the transient analysis codes through a series of benchmark exercises; and a review of candidate fuels and core materials, including their fabrication and irradiation. Education and communication to foster understanding of the growing needs for nuclear power in general and for the technology of the GCFR in particular is specific goal of the EU project. (authors)

  2. 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.

  3. 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.

  4. 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.

  5. 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

  6. 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.

  7. 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.

  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. 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...

  10. 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.

  11. 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

  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. 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.

  14. 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.

  15. 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.

  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. 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.

  18. Gas Stripping in the Simulated Pegasus Galaxy

    NASA Astrophysics Data System (ADS)

    Mercado, Francisco Javier; Samaniego, Alejandro; Wheeler, Coral; Bullock, James

    2017-01-01

    We utilize the hydrodynamic simulation code GIZMO to construct a non-cosmological idealized dwarf galaxy built to match the parameters of the observed Pegasus dwarf galaxy. This simulated galaxy will be used in a series of tests in which we will implement different methods of removing the dwarf’s gas in order to emulate the ram pressure stripping mechanism encountered by dwarf galaxies as they fall into more massive companion galaxies. These scenarios will be analyzed in order to determine the role that the removal of gas plays in rotational vs. dispersion support (Vrot/σ) of our galaxy.

  19. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR

    SciTech Connect

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.

  20. 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

  1. UNCERTAINTY QUANTIFICATION OF CALCULATED TEMPERATURES FOR ADVANCED GAS REACTOR FUEL IRRADIATION EXPERIMENTS

    SciTech Connect

    Pham, Binh Thi-Cam; Hawkes, Grant Lynn; Einerson, Jeffrey James

    2015-08-01

    This paper presents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) fuel irradiation experiments conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technology Research and Development program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR tests, the results of the numerical simulations are used in combination with statistical analysis methods to improve qualification of measured data. The temperature simulation data for AGR tests are also used for validation of the fission product transport and fuel performance simulation models. These crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. To quantify the uncertainty of AGR calculated temperatures, this study identifies and analyzes ABAQUS model parameters of potential importance to the AGR predicted fuel temperatures. The selection of input parameters for uncertainty quantification of the AGR calculated temperatures is based on the ranking of their influences on variation of temperature predictions. Thus, selected input parameters include those with high sensitivity and those with large uncertainty. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of AGR calculated temperatures. Expert judgment is used as the basis to specify the uncertainty range for selected input parameters. The input uncertainties are dynamic accounting for the effect of unplanned events and changes in thermal properties of capsule components over extended exposure to high temperature and fast neutron irradiation. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish

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

    SciTech Connect

    Blaine Grover; John Maki; David Petti

    2010-10-01

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

  3. Simulation of transients and transport in plasma processing reactors

    NASA Astrophysics Data System (ADS)

    Subramonium, Pramod

    Low temperature, high plasma density reactors are widely used for etching and deposition during microelectronics fabrication. As the wafer size increases, the requirements for azimuthal symmetry and side-to-side uniformity become more stringent. The development of plasma equipment models (PEMs) for investigating chemical, physical, and engineering scaling issues for plasma processing has significantly advanced in the recent years. PEMs in two and three dimensions have been developed with the goals of both investigating basic physical processes and for use in the design of plasma equipment. In spite of the success of these activities, issues related to three-dimensional (3D) phenomena have come to the forefront and cannot be adequately addressed by the present generation of PEMs. To design reactors with good side-to-side symmetry, design engineers require 3D information of plasma properties. The lack of inexpensive azimuthal measurement techniques for plasma properties in inductively coupled plasma (ICP) reactors makes it more difficult for experiments to provide this information. In this thesis, two-dimensional (2D) and 3D hybrid models were developed to investigate transient phenomena (time variation of plasma properties) during pulsed operation of ICP reactors. Employing the 2D model, it was demonstrated that utilizing transients during pulsed operation, energetic negative ions can be extracted from pulsed ICPs that can aid in reducing charge buildup on wafers. Energetic negative ions can be extracted from Ar/Cl2 pulsed ICPs with pulsed low frequency (1--2 MHz) substrate biases. Employing the 3D model, the impact of asymmetric pumping on plasma properties during continuous wave (CW) ICP operation and the effect of transients on these flow-induced asymmetries were quantified. Asymmetric pumping results in non-uniform species densities, which then feed back through plasma conductivity making the power deposition azimuthally asymmetric. The asymmetries in plasma

  4. 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...

  5. 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.

  6. Dynamic analysis of gas-core reactor system

    NASA Technical Reports Server (NTRS)

    Turner, K. H., Jr.

    1973-01-01

    A heat transfer analysis was incorporated into a previously developed model CODYN to obtain a model of open-cycle gaseous core reactor dynamics which can predict the heat flux at the cavity wall. The resulting model was used to study the sensitivity of the model to the value of the reactivity coefficients and to determine the system response for twenty specified perturbations. In addition, the model was used to study the effectiveness of several control systems in controlling the reactor. It was concluded that control drums located in the moderator region capable of inserting reactivity quickly provided the best control.

  7. A gas-cooled-reactor closed-Brayton-cycle demonstration with nuclear heating.

    SciTech Connect

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

    2004-09-01

    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. 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.

  10. Containment building atmosphere response during severe accidents in high temperature gas-cooled reactors

    SciTech Connect

    Kroeger, P.G.; Chan, B.C.

    1985-01-01

    Several safety evaluations for large High Temperature Gas Cooled Reactors (HTGR), using a Prestressed Concrete Reactor Vessel (PCRV) design, have concluded that Unrestricted Core Heatup Accidents (UCHA) present the most important severe accidents, resulting in the dominant source term. While the core thermohydraulic transients for such accident sequences have been presented previously, the subject of this paper is the containment building (CB) atmosphere transient, with primary emphasis on the CB atmosphere temperature and pressure, as overpressurization is the most likely failure mode.

  11. Numerical simulation of polariton Bose gas thermalization

    NASA Astrophysics Data System (ADS)

    Kartsev, P. F.; Kuznetsov, I. O.

    2016-08-01

    In this work, we present the numerical simulation of the process a Bose gas thermalization and the formation of the condensate. Our approach is based on kinetic equations and “Fermi's golden rule” in the incoherent approximation. Direct summation of terms is performed using GPGPU OpenCL parallel code using AMD Radeon HD 7970.

  12. 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.

  13. CRAW: An expert system for nuclear reactor cover gas alarm analysis

    SciTech Connect

    Zimmerman, B.D.; Rawlins, J.A.

    1986-02-01

    The CRAW expert system was developed for operational use at the Fast Flux Test Facility research nuclear reactor, which is operated for the Department of Energy by the Westinghouse Hanford Company. The CRAW expert system is based on the EXPERT-EASE commercial expert system development software. It provides ''on-the-spot'' analysis of the cause of reactor cover gas activity alarms as they are received by Operations personnel in the reactor control room. The CRAW system provides an example of how an expert system can save human time and effort in an operational setting, and also highlights some of the practical considerations of the expert system development process.

  14. Compartmental models for continuous flow reactors derived from CFD simulations.

    PubMed

    Gresch, Markus; Brügger, Raphael; Meyer, Alain; Gujer, Willi

    2009-04-01

    Reactor modeling is of major interest in environmental technology. In this context, new contaminants with higher degradation requirements increase the importance of reactor hydraulics. CFD (Computational Fluid Dynamics) may meet this challenge but is expensive for everyday use. In this paper, we provide research and practice with a methodology designed to automatically reduce the complexity of such a high-dimensional flow model to a compartmental model. The derivation is based on the concentration field of a reacting species which is included in the steady state CFD simulation. While still capturing the most important flow features, the compartmental model is fast, easy to use, and open for process modeling with yet unknown compounds. The inherent overestimation of diffusion by compartmental models has been corrected by locally adjusting turbulent fluxes. We successfully applied the methodology to the ozonation process and experimentally verified it with tracer experiments. The loss of information was quantified as a deviation from CFD performance prediction for different reactions. With increasing discretisation of the compartmental model, these deviations diminish. General advice on the necessary discretisation is given.

  15. 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

  16. 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

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. Induction simulation of gas core nuclear engine

    NASA Technical Reports Server (NTRS)

    Poole, J. W.; Vogel, C. E.

    1973-01-01

    The design, construction and operation of an induction heated plasma device known as a combined principles simulator is discussed. This device incorporates the major design features of the gas core nuclear rocket engine such as solid feed, propellant seeding, propellant injection through the walls, and a transpiration cooled, choked flow nozzle. Both argon and nitrogen were used as propellant simulating material, and sodium was used for fuel simulating material. In addition, a number of experiments were conducted utilizing depleted uranium as the fuel. The test program revealed that satisfactory operation of this device can be accomplished over a range of operating conditions and provided additional data to confirm the validity of the gas core concept.

  2. 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.

  3. 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.

  4. CASL virtual reactor predictive simulation: Grid-to-Rod Fretting wear

    NASA Astrophysics Data System (ADS)

    Lu, Roger Y.; Karoutas, Zeses; Sham, T.-L.

    2011-08-01

    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.

  5. 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.

  6. 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

  7. 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.

  8. 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.

  9. 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.

  10. 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

  11. 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.

  12. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2009-09-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight 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 new 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 ten years to demonstrate and qualify new 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 the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  13. 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.

  14. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect

    1999-10-01

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through September 1999.

  15. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect

    1999-11-01

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through October 1999.

  16. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect

    2000-02-01

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through January 2000.

  17. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect

    1998-08-01

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through July 1999.

  18. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect

    2000-01-01

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through December 1999.

  19. Engineering development of ceramic membrane reactor system for converting natural gas to hydrogen and synthesis gas for liquid transportation fuels

    SciTech Connect

    1998-05-01

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through April 1998.

  20. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect

    1999-12-01

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through November 1999.

  1. Engineering development of ceramic membrane reactor system for converting natural gas to hydrogen and synthesis gas for liquid transportation fuels

    SciTech Connect

    1998-07-01

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through June 1998.

  2. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect

    1999-03-01

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through February 1999.

  3. Experimental and Numerical Simulation of Boron Transport in a Nuclear Reactor Vessel

    NASA Astrophysics Data System (ADS)

    Radcliff, Thomas Dean

    1995-01-01

    The Westinghouse AP600 reactor design uses a gravity -forced safety injection system with nozzles in the vessel downcomer. In the event of overcooling transients, this system delivers soluble boron to the core to offset moderator defect reactivity. To evaluate the outcome of a design basis overcooling event, a tool to predict the transport of boron to the core was required. A hybrid computational fluid dynamic/fluid element tracking model was developed for this task. In this technique, the loop and safety injection flow was determined by the 1-D system code LOFTRAN. From these boundary conditions, the reactor vessel steady-state velocity field and k- varepsilon turbulence parameters were found using the 3-D computational fluid dynamics code FLOTRAN. These pointwise values were used to define the flow field characteristics in a fluid element tracking model. In this random walk model, the k-varepsilon values at a point were used to find a local turbulent diffusion time scale and a distribution of length scales. A distribution of molecular diffusion length scales were also generated. Fluid element motion was determined by combining the deterministic convective transport during a time interval with turbulent and molecular diffusion components chosen randomly from the distributions. The transient concentration distribution was determined from the time and position of the elements as they reached the core inlet. A scaling analysis of the reactor system showed that buoyancy and turbulent diffusion effects were of equal importance during overcooling transient conditions. A 1:9 scale model was constructed using air and dense gas to simulate the reactor coolant and safety injection fluid. Experiments were performed with velocities chosen to give Richardson and mixing Reynolds number scaling. Concentration of the dense gas was measured at the core inlet using a sonic nozzle/hot film anemometer instrument. The results of these experiments were used to validate the numerical

  4. 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.

  5. 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.

  6. 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.

  7. 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.

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

    DOEpatents

    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.

  9. Ambient Laboratory Coater for Advanced Gas Reactor Fuel Development

    SciTech Connect

    Duane D. Bruns; Robert M. Counce; Irma D. Lima Rojas

    2010-06-09

    this research is targeted at developing improved experimentally-based scaling relationships for the hydrodynamics of shallow, gas-spouted beds of dense particles. The work is motivated by the need to more effctively scale up shallow spouted beds used in processes such as in the coating of nuclear fuel particles where precise control of solids and gas circulation is critically important. Experimental results reported here are for a 50 mm diameter spouted bed containing two different types of bed solids (alumina and zirconia) at different static bed depths and fluidized by air and helium. Measurements of multiple local average pressures, inlet gas pressure fluctuations, and spout height were used to characterize the bed hydrodynamics for each operating condition. Follow-on studies are planned that include additional variations in bed size, particle properties, and fluidizing gas. The ultimate objective is to identify the most important non-dimensional hydrodynamic scaling groups and possible spouted-bed design correlations based on these groups.

  10. Continuous crafting of uniform colloidal nanocrystals using an inert-gas-driven microflow reactor.

    PubMed

    Tang, Hailong; He, Yanjie; Li, Bo; Jung, Jaehan; Zhang, Chuchu; Liu, Xiaobo; Lin, Zhiqun

    2015-06-07

    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.

  11. 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.

  12. 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.

  13. 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)

  14. 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.

  15. Development of a simulator for ozone/UV reactor based on CFD analysis.

    PubMed

    Kamimura, M; Furukawa, S; Hirotsuji, J

    2002-01-01

    A new CFD (Computational Fluid Dynamics) simulator for an O3/UV reactor where ozone dissolved water flows under the irradiation of UV, has been developed by combining a fluid dynamics model with a complex radical reaction model. The radical reaction model used in this simulator was found to be reasonable, because the results obtained from the simulation of a completely stirred tank reactor (CSTR) system were in good agreement with the experimental results, e.g., the concentrations of total organic carbon (TOC), hydrogen peroxide and dissolved ozone obtained from a lab-scale CSTR. Furthermore, by using this CFD simulator, the distributions of substances such as hydroxyl radical (OH*) and hydrogen peroxide in the O3/UV reactor have been investigated. These distributions showed that this CFD simulator was considered to be reasonable. In addition, the simulation results suggested that conventional reactors were not optimized.

  16. 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

  17. 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.

  18. 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.

  19. 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)

  20. Temperature effect towards methane gas production and performances of anaerobic fixed bed reactors

    NASA Astrophysics Data System (ADS)

    Budiastuti, H.; Widyabudiningsih, D.; Kurnia, D. R. D.

    2016-11-01

    This study was conducted to observe the effect of temperature towards methane gas production and performances of Anaerobic Fixed Bed Reactors (AFBR) in degrading artificial wastewater. During seeding and acclimatization as well as operation at normal loading rates the reactor was controlled at 35 ± 1 °C. The reactor temperature was then switched to room temperature to compare its performance with performance of the temperature controlled reactor. Seeding was performed by wastewater feeding at concentration of 5,300 mgCOD/L. Seeding was continued to acclimatization after COD efficiency reached about 30% and mixed liquor volatile suspended solids (MLVSS) in the reactor were in the range of 28 to 29 g/L. Acclimatization was conducted by step increased feeding while obtaining organic loading rates (OLR) of 530 mgCOD/L/day and 40 days HRT, and it was stopped when COD efficiencies almost constant at about 80% were achieved. Normal loads at OLR of 530 mgCOD/L/day and HRT of 40 days resulted in COD efficiencies in the range of 80 to 92%, produced methane gas ranged from 80 to 170 ml/day, at pH around 7 during controlled temperature. At room temperature, COD efficiencies decreased to the fluctuated range of 78 to 84%, methane gas dropped to the highest of 144 ml/day but pH range was still at around 7. At OLR 1.5 times normal OLR (795 mgCOD/L/day) and HRT of 40 days, the controlled temperature reactor shows superiority by producing efficiencies in the range of 84 to 94% and 260 mL/day of CH4 as the highest gas produced during observation. Efficiencies of COD obtained during uncontrolled temperature dropped to the range of 60 to 80%. The uncontrolled temperature reactor only produced the highest CH4 production of 48 mL/day. pH from both reactors still ranged in the normal pH range about 7. It shows that AFBR has to be controlled at around 35 °C to produce higher COD efficiencies and methane gas production.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Mesh Copy/Move/Merge Tool for Reactor Simulation Applications

    SciTech Connect

    Tautges, Timothy J.; Jain, Rajeev

    2014-04-30

    Reactor core simulations require the construction and mesh generation for core models consisting of lattices of fuel and other rods grouped into assemblies, and lattices of assemblies of several types grouped into a core model. A set of tools has been described for generating assembly and core lattice models. Both rectangular and hexagonal lattices are supported. The tools operate in three stages. First, assembly models of various types can be generated by the AssyGen tool, based on input describing the content of unit cells, the arrangement of unit cells in the lattice, and the extent of the lattice and any surrounding material. After generating the assembly model, the model is meshed with the CUBIT mesh generation toolkit, optionally based on a journal file output by AssyGen. After one or more assembly model meshes have been constructed, they are arranged in a core model using the CoreGen tool. The input for CoreGen is similar to that of AssyGen, with assembly models substituted for unit cells. AssyGen and CoreGen also annotate the models with material and volume groupings necessary for specifying materials and boundary conditions required by the analysis. The AssyGen and CoreGen tools are packaged in the open-source MeshKit library for mesh generation; download and build instructions are included in this document.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. Remote reactor repair: GTA (gas tungsten Arc) weld cracking caused by entrapped helium

    SciTech Connect

    Kanne, Jr, W R

    1988-01-01

    A repair patch was welded to the wall of a nuclear reactor tank using remotely controlled thirty-foot long robot arms. Further repair was halted when gas tungsten arc (GTA) welds joining type 304L stainless steel patches to the 304 stainless steel wall developed toe cracks in the heat-affected zone (HAZ). The role of helium in cracking was investigated using material with entrapped helium from tritium decay. As a result of this investigation, and of an extensive array of diagnostic tests performed on reactor tank wall material, helium embrittlement was shown to be the cause of the toe cracks.

  16. 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.

  17. 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.

  18. Use of lift gas in an FCC reactor riser

    SciTech Connect

    Hammershaimb, H. U.; Lomas, D. A.

    1984-10-30

    A process for converting normally liquid hydrocarbons with an active fluid catalytic cracking catalyst which comprises: passing an upflowing suspension of hot regenerated active fluid catalytic cracking catalyst in a lift gas comprising hydrocarbons including not more than 10 mole % C/sub 3/ and heavier hydrocarbons through a lower portion of a vertically orientated riser conversion zone at a velocity of from about 1.8 to less than 12.2 meters per second, and for a residence time from about 0.5 to about 15 seconds, the ratio of catalyst to hydrocarbon in said lift gas being greater than 80; and introducing the normally liquid hydrocarbons into the upflowing suspension at a locus in the riser conversion zone downstream of the lower portion to form a catalyst-hydrocarbon mixture wherein the temperature and residence time are sufficient to effect the desired conversion. The process is particularly useful for the treatment of a heavy residual feedstock.

  19. Gamma heating in reflector heat shield of gas core reactor

    NASA Technical Reports Server (NTRS)

    Lofthouse, J. H.; Kunze, J. F.; Young, T. E.; Young, R. C.

    1972-01-01

    Heating rate measurements made in a mock-up of a BeO heat shield for a gas core nuclear rocket engine yields results nominally a factor of two greater than calculated by two different methods. The disparity is thought to be caused by errors in neutron capture cross sections and gamma spectra from the low cross-section elements, D, O, and Be.

  20. Hydrodynamic instabilities in gas-liquid monolithic reactors

    SciTech Connect

    Grolman, E.; Edvinsson, R.K.; Stankiewicz, A.; Moulijn, J.A.

    1996-12-31

    Liquid holdup and pressure drop were measured during the co-current down flow of air and water through a monolith in the Taylor flow regime. The model presented accounts for the significant, up to three fold, increase in frictional pressure drop that is caused by the presence of gas bubbles. It is accurate to within 20%. In addition, the model presented is used to predict hydrodynamic stability, which is defined as the situation where all channels transport gas and liquid in the direction of mass flow. Essential for stability is a sufficiently good initial liquid distribution, which was achieved with a shower-type distributor. Furthermore, distribution was significantly enhanced by the natural occurrence of a well-mixed foam (aerated liquid) layer on top of the monolith at liquid holdup values above 0.5. The quality of the liquid distribution across the monolith follows directly from on-line, integral liquid holdup measurements. Monoliths have been developed and used extensively for catalytic automobile exhaust conversion, from which they have inherited mechanical and thermal robustness. Today, monoliths provide an interesting alternative to trickle-bed and slurry-column technologies, for solid-catalyzed gas-liquid reactions.

  1. 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.

  2. Canonical Bose gas simulations with stochastic gauges.

    PubMed

    Drummond, P D; Deuar, P; Kheruntsyan, K V

    2004-01-30

    A technique to simulate the grand canonical ensembles of interacting Bose gases is presented. Results are generated for many temperatures by averaging over energy-weighted stochastic paths, each corresponding to a solution of coupled Gross-Pitaevskii equations with phase noise. The stochastic gauge method used relies on an off-diagonal coherent-state expansion, thus taking into account all quantum correlations. As an example, the second-order spatial correlation function and momentum distribution for an interacting 1D Bose gas are calculated.

  3. 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

  4. Molten salt reactor neutronics and fuel cycle modeling and simulation with SCALE

    DOE PAGES

    Betzler, Benjamin R.; Powers, Jeffrey J.; Worrall, Andrew

    2017-03-01

    Current interest in advanced nuclear energy and molten salt reactor (MSR) concepts has enhanced interest in building the tools necessary to analyze these systems. A Python script known as ChemTriton has been developed to simulate equilibrium MSR fuel cycle performance by modeling the changing isotopic composition of an irradiated fuel salt using SCALE for neutron transport and depletion calculations. Some capabilities in ChemTriton that have improved, include a generic geometry capable of modeling multi-zone and multi-fluid systems, enhanced time-dependent feed and separations, and a critical concentration search. Although more generally applicable, the capabilities developed to date are illustrated in thismore » paper in three applied problems: (1) simulating the startup of a thorium-based MSR fuel cycle (a likely scenario requires the first of these MSRs to be started without available 233U); (2) determining the effect of the removal of different fission products on MSR operations; and (3) obtaining the equilibrium concentration of a mixed-oxide light-water reactor fuel in a two-stage fuel cycle with a sodium fast reactor. Moreover, the third problem is chosen to demonstrate versatility in an application to analyze the fuel cycle of a non-MSR system. During the first application, the initial fuel salt compositions fueled with different sources of fissile material are made feasible after (1) removing the associated nonfissile actinides after much of the initial fissile isotopes have burned and (2) optimizing the thorium concentration to maintain a critical configuration without significantly reducing breeding capability. In the second application, noble metal, volatile gas, and rare earth element fission products are shown to have a strong negative effect on criticality in a uranium-fueled thermal-spectrum MSR; their removal significantly increases core lifetime (by 30%) and fuel utilization. In the third application, the fuel of a mixed-oxide light

  5. Decomposition of phenol by hybrid gas/liquid electrical discharge reactors with zeolite catalysts.

    PubMed

    Kusić, Hrvoje; Koprivanac, Natalija; Locke, Bruce R

    2005-10-17

    Application of hybrid gas/liquid electrical discharge reactors and a liquid phase direct electrical discharge reactor for degradation of phenol in the presence and absence of zeolites have been investigated. Hybrid gas/liquid electrical discharges involve simultaneous high voltage electrical discharges in water and in the gas phase above the water surface leading to the additional OH radicals in the liquid phase and ozone formation in the gas phase with subsequent dissolution into the liquid. The role of applied zeolites, namely NH4ZSM5, FeZSM5 and HY, were also studied. Phenol degradation and production of primary phenol by-products, catechol and hydroquinone, during the treatment were monitored by HPLC measurements. The highest phenol removal results, 89.4-93.6%, were achieved by electrical discharge in combination with FeZSM5 in all three configurations of corona reactors. These results indicate that the Fenton reaction has significant influence on overall phenol removal efficiency in the electrical discharge/FeZSM5 system due to the additional OH radical formation from hydrogen peroxide generated by the water phase discharge.

  6. 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.

  7. 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

  8. 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)

  9. 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.

  10. High-temperature gas-cooled reactor safety-reliability program plan

    SciTech Connect

    Not Available

    1981-03-01

    The purpose of this document is to present a safety plan as part of an overall program plan for the design and development of the High Temperature Gas-Cooled Reactor (HTGR). This plan is intended to establish a logical framework for identifying the technology necessary to demonstrate that the requisite degree of public risk safety can be achieved economically. This plan provides a coherent system safety approach together with goals and success criterion as part of a unifying strategy for licensing a lead reactor plant in the near term. It is intended to provide guidance to program participants involved in producing a technology base for the HTGR that is fully responsive to safety consideration in the design, evaluation, licensing, public acceptance, and economic optimization of reactor systems.

  11. 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.

  12. 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.

  13. Helium circulator design considerations for modular high temperature gas-cooled reactor plant

    SciTech Connect

    McDonald, C.F.; Nichols, M.K.

    1986-12-01

    Efforts are in progress to develop a standard modular high temperature gas-cooled reactor (MHTGR) plant that is amenable to design certification and serial production. The MHTGR reference design, based on a steam cycle power conversion system, utilizes a 350 MW(t) annular reactor core with prismatic fuel elements. Flexibility in power rating is afforded by utilizing a multiplicity of the standard module. The circulator, which is an electric motor-driven helium compressor, is a key component in the primary system of the nuclear plant, since it facilitates thermal energy transfer from the reactor core to the steam generator; and, hence, to the external turbo-generator set. This paper highlights the helium circulator design considerations for the reference MHTGR plant and includes a discussion on the major features of the turbomachine concept, operational characteristics, and the technology base that exists in the US.

  14. RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

    SciTech Connect

    Chang Oh; Cliff Davis; Goon C. Park

    2007-09-01

    A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls.

  15. Safety aspects of forced flow cooldown transients in Modular High Temperature Gas-Cooled Reactors

    SciTech Connect

    Kroger, P.G. )

    1993-05-01

    During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs), the main Heat Transport System (HTS) and the Shutdown Cooling System n removed by the passive Reactor (SCS) are assumed to have failed. Decay heat is the Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This report used the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits.

  16. Laboratory simulation of cometary neutral gas ionization

    NASA Technical Reports Server (NTRS)

    Chang, Tsuey-Fen; Rahman, H. U.; White, R. S.

    1989-01-01

    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.

  17. Evaluation of model parameters for simulating TiO(2) coated UV reactors.

    PubMed

    Duran, J E; Taghipour, F; Mohseni, M

    2011-01-01

    A CFD-based model for simulating TiO(2) coated photocatalytic reactors used in drinking water treatment applications was preliminarily evaluated. The model includes aspects of hydrodynamics, mass transfer, UV-radiation field, and surface chemical reactions. Appropriate models for each of the associated physicochemical phenomena were experimentally or analytically examined. Once defined and evaluated, the individual models were integrated into a CFD-based model for simulating photocatalytic reactor performance, which was experimentally evaluated.

  18. 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

  19. Simulation of the modified K reactor supplementary safety system

    SciTech Connect

    Paik, I.K.; Canas, L.R. ); Peterson, P.F. )

    1991-01-01

    The supplementary safety system (SSS) of the K reactor provides a second line of defense to shut down the reactor if the safety and control rods fail to scram. The SSS was originally designed to inject a neutron poison solution (ink) into the reactor tank via spargers. Recently, concerns arose that the ink inventory might run out before the ink front returned to the moderator during a loss-of-ac-power transient in which the coolant pumps coast down. Thus, a new system has been added to inject additional ink through the pump suctions so that ink will arrive in the core before depletion of the sparger ink. The MODFLOW code was developed to calculate the moderator flow distribution in Savannah River site (SRS) reactors, including the effects of inertia and stratification from buoyancy forces.

  20. 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

    2016-12-01

    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.

  1. 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.

  2. Simulation of Feynman-alpha measurements from SILENE reactor using a discrete ordinates code

    SciTech Connect

    Humbert, P.; Mechitoua, B.; Verrey, B.

    2006-07-01

    In this paper we present the simulation of Feynman-{alpha} measurements from SILENE reactor using the discrete ordinates code PANDA. A 2-D cylindrical model of SILENE reactor is designed for computer simulations. Two methods are implemented for variance to mean calculation. In the first method we used the Feynman point reactor formula where the parameters (Diven factor, reactivity, detector efficiency and alpha eigenvalue) are obtained by 2-D PANDA calculations. In the second method the time dependent adjoint equations for the first two moments are solved. The calculated results are compared to the measurements. Both methods are in excellent agreement with the experimental data. (authors)

  3. 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.

  4. 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.

  5. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2012-06-01

    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&D items for the inherently-safe HTGR are described in this paper.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

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

    SciTech Connect

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

    2012-08-01

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

  12. 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.

  13. MOVPE GaN gas phase chemistry for reactor design and optimization

    SciTech Connect

    Safvi, S.A.; Thon, A.; Kuech, T.F.; Redwing, J.M.; Flynn, J.S.; Tischler, M.A.

    1997-12-31

    The results of gas phase decomposition studies are used to construct a chemistry model which is compared to data obtained from an experimental MOVPE reactor. A flow tube reactor is used to study gas phase reactions between trimethylgallium (TMG) and ammonia at high temperatures, characteristic to the metalorganic vapor phase epitaxy (MOVPE) of GaN. Experiments were performed to determine the effect of the mixing of the Group III precursors and Group V precursors on the growth rate, growth uniformity and film properties. Growth rates are predicted for simple reaction mechanisms and compared to those obtained experimentally. Quantification of the loss of reacting species due to oligomerization is made based on experimentally observed growth rates. The model is used to obtain trends in growth rate and uniformity with the purpose of moving towards better operating conditions.

  14. 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.

  15. Transesterification of vegetable oils: Simulating the replacement of batch reactors with continuous reactors.

    PubMed

    Fonseca, Felipe A S; Vidal-Vieira, José A; Ravagnani, Sergio P

    2010-11-01

    A kinetic model was employed to represent biodiesel production via transesterification of vegetable oils. Reaction rate constants found in the open literature were used in order to compare the behavior of batch and continuous processes. A single continuous stirred tank reactor (CSTR) under the usual operation conditions was not capable of achieving the same productivity as a batch process. However, when reactors in series were used, the continuous process presented a behavior similar to batch processes. As a result, it was evidenced that a series of CSTRs can be an industrially feasible choice for replacing batch transesterification reactors in large scale biodiesel plants. Further, it was shown that the loss in productivity caused by changing from batch to continuous process can be compensated by means of using higher catalyst concentrations.

  16. 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).

  17. 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.

  18. 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.

  19. Proceedings of the Second Technical Meeting on High Temperature Gas-Cooled Reactors

    NASA Astrophysics Data System (ADS)

    1992-07-01

    The Second Technical Meeting on High Temperature Gas-cooled Reactors (HTGR's) was held on 11-12 Mar. 1992. The following topics were discussed: the present status and the results of Research and Development (R and D) of HTGR's; the items of R and D which should be promoted more positively in the future; and the strategy for development of high temperature engineering and examination in JAERI. This proceeding summarizes the papers that were presented.

  20. 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).

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Method and combustion reactor for converting uranium hexafluoride gas to an oxide of uranium

    SciTech Connect

    Linz, D.G.; Grier, R.K. Jr.; Wetzel, J.A.; Dada, A.G.; Connolly, J.D.

    1987-10-06

    This patent describes a method for converting uranium hexafluoride to uranium oxide in a gaseous reaction medium, comprising the steps of: (a) supplying gases individually comprising uranium hexafluoride, oxygen and hydrogen, and mixing same; (b) feeding the mixed gases as a combined jet stream into a reactor zone wherein the uranium hexafluoride is converted to an oxide of uranium; (c) supplying hydrogen gas separately in an annular flow pattern surrounding and generally parallel to the jet stream of the mixed gases into the reaction zone; and (d) supplying a quenching gas separately intermediate the reaction zone where the mixed gases convert uranium hexafluoride to an oxide of uranium.

  6. A flow reactor setup for photochemistry of biphasic gas/liquid reactions.

    PubMed

    Schachtner, Josef; Bayer, Patrick; Jacobi von Wangelin, Axel

    2016-01-01

    A home-built microreactor system for light-mediated biphasic gas/liquid reactions was assembled from simple commercial components. This paper describes in full detail the nature and function of the required building elements, the assembly of parts, and the tuning and interdependencies of the most important reactor and reaction parameters. Unlike many commercial thin-film and microchannel reactors, the described set-up operates residence times of up to 30 min which cover the typical rates of many organic reactions. The tubular microreactor was successfully applied to the photooxygenation of hydrocarbons (Schenck ene reaction). Major emphasis was laid on the realization of a constant and highly reproducible gas/liquid slug flow and the effective illumination by an appropriate light source. The optimized set of conditions enabled the shortening of reaction times by more than 99% with equal chemoselectivities. The modular home-made flow reactor can serve as a prototype model for the continuous operation of various other reactions at light/liquid/gas interfaces in student, research, and industrial laboratories.

  7. A flow reactor setup for photochemistry of biphasic gas/liquid reactions

    PubMed Central

    Schachtner, Josef; Bayer, Patrick

    2016-01-01

    Summary A home-built microreactor system for light-mediated biphasic gas/liquid reactions was assembled from simple commercial components. This paper describes in full detail the nature and function of the required building elements, the assembly of parts, and the tuning and interdependencies of the most important reactor and reaction parameters. Unlike many commercial thin-film and microchannel reactors, the described set-up operates residence times of up to 30 min which cover the typical rates of many organic reactions. The tubular microreactor was successfully applied to the photooxygenation of hydrocarbons (Schenck ene reaction). Major emphasis was laid on the realization of a constant and highly reproducible gas/liquid slug flow and the effective illumination by an appropriate light source. The optimized set of conditions enabled the shortening of reaction times by more than 99% with equal chemoselectivities. The modular home-made flow reactor can serve as a prototype model for the continuous operation of various other reactions at light/liquid/gas interfaces in student, research, and industrial laboratories. PMID:27829887

  8. 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.

  9. Anoxic gas recirculation system for fouling control in anoxic membrane reactor.

    PubMed

    Lee, Hansaem; Lee, Daeju; Hong, Seongwan; Yun, Geum Hee; Kim, Sungpyo; Hwang, Jung Ki; Lee, Woojae; Yun, Zuwhan

    2014-06-01

    Anoxic gas recirculation system was applied to control the membrane fouling in pilot-scale 4-stage anoxic membrane bioreactor (MBR). In the anaerobic-anoxic-anoxic-aerobic flow scheme, hydrophilic polytetrafluoroethylene (PTFE) membrane (0.2 μm, 7.2 m(2)/module) was submerged in the second anoxic zone. During 8 months operation, the average flux of the membrane was 21.3 L/(m(2)·hr). Chemical cleaning of the membrane was conducted only once with sodium hydroxide and sodium hypochlorite. Dissolved oxygen (DO) concentration in the second anoxic zone was maintained with an average of 0.19 ± 0.05 mg/L. Gas chromatography analysis showed that the headspace gas in the second anoxic reactor was mainly consisted of N2 (93.0% ± 2.5%), O2 (3.8% ± 0.6%), and CO2 (3.0% ± 0.5%), where the saturation DO concentration in liquid phase was 1.57 mg/L. Atmospheric O2 content (20.5% ± 0.8%) was significantly reduced in the anoxic gas. The average pH in the reactor was 7.2 ± 0.4. As a result, the recirculation of the anoxic gas was successfully applied to control the membrane fouling in the anoxic MBR. Copyright © 2014 The Research Centre for Eco-Environmental Sciences, Chinese Academy of Sciences. Published by Elsevier B.V. All rights reserved.

  10. 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

  11. 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.

  12. 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.

  13. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  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. Numerical study of cavitation and pinning effects due to gas injection through a bed of particles: application to a radial-flow moving-bed reactor

    NASA Astrophysics Data System (ADS)

    Vinay, Guillaume; Vasquez, Felaurys; Richard, Florence; Applied Mechanics Team

    2016-11-01

    In the petroleum and chemical industries, radial-flow moving-bed reactors are used to carry out chemical reactions such as catalytic reforming. Radial-flow reactors provide high capacity without increased pressure drop or greatly increased vessel dimensions. This is done by holding the catalyst in a basket forming an annular bed, and causing the gas to flow radially between the outer annulus and the central tube. Catalyst enter the top of the reactor, move through the vessel by gravity to the bottom where it is removed and then regenerated. Within the catalytic bed, the combined effects of particles motion and radial injection of the gas may lead to cavitation and pinning phenomenon that may clearly damage the reactor. We study both cavitation and pinning effects using an in-house numerical software, named PeliGRIFF (www.peligriff.com/), designed to simulate particulate flows at different scales; from the particle scale, where fluid/particle interactions are directly solved, to the particles suspension scale where the fluid/solid interactions are modeled. In the past, theoretical and experimental studies have already been conducted in order to understand the way cavitation and pinning occur. Here, we performed simulations involving a few thousands of particles aiming at reproducing experimental experiments. We will present comparisons between our numerical results and experimental results in terms of pressure drop, velocity, porosity.

  16. 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.

  17. 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.

  18. Simulation of ultrasonic inspection for sodium cooled reactors using CIVA

    SciTech Connect

    Reverdy, F.; Baque, F.; Lu, B.; Jezzine, K.; Dorval, V.; Augem, J. M.

    2011-07-01

    In-service inspection of sodium fast reactors (SFR) requires the development of non-destructive techniques adapted to the harsh conditions of the environment (opaque and hot) and the complexity of the examination (large and littered reactor block). Ultrasonic techniques are seen as suitable candidates for the inspection of SFRs and two approaches are being followed: inside inspection where transducers are directly immersed in sodium coolant and inspection from outside with transducers positioned along the wall of the main vessel. Probe design and inspection performances can be predicted by using comprehensive models that can take into account the various variables of the problem. These models are explained in this paper. (authors)

  19. Simulation and Design of an Automatic Controller for a Fast Breeder Nuclear Reactor Power Plant.

    DTIC Science & Technology

    BREEDER REACTORS, *REACTOR CONTROL, *REACTOR REACTIVITY, COMPUTER PROGRAMMING, NEUTRON TRANSPORT THEORY, REACTOR FUELS, REACTOR FUEL CLADDING , HEAT TRANSFER, COMPUTER PROGRAMS, LOGIC CIRCUITS, THESES.

  20. Gas dynamic simulations of galaxy formation

    NASA Technical Reports Server (NTRS)

    Evrard, August E.

    1993-01-01

    Results are presented from a simulation modeling the formation of a group of galaxies in a 'standard' cold, dark matter universe with delta = 1, h sub 0 = 50 km/(s(Mpc)), baryon fraction omega sub b = 0.1 and spectrum normalization sigma sub 8 = 0.6 (bias parameter b = 1.7). Initial conditions are generated within a periodic box with comoving length 16 Mpc in a manner constrained to produce a small cluster of total mass approximately 10 exp 14 solar mass. Two sets of 643 particles are used to model the dark matter and baryon fluids. Each gas particle represents 1.08 x 10 exp -8 solar mass, implying an L* galaxy is resolved by approximately 1000 particles. The system is evolved self-consistently in three dimensions using the combined N-body/hydrodynamic scheme P3MSPH up to a final redshift z = 1. Evolving to the present is prohibited by the fact that the mean density in the simulated volume is above critical and the entire volume would be going nonlinear beyond this point, We are currently analyzing another run with somewhat poorer mass resolution which was evolved to the present.

  1. 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

  2. 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)

  3. 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.

  4. COMSOL Simulations for Steady State Thermal Hydraulics Analyses of ORNL s High Flux Isotope Reactor

    SciTech Connect

    Khane, Vaibhav B; Jain, Prashant K; Freels, James D

    2012-01-01

    Simulation models for steady state thermal hydraulics analyses of Oak Ridge National Laboratory s High Flux Isotope Reactor (HFIR) have been developed using the COMSOL Multiphysics simulation software. A single fuel plate and coolant channel of each type of HFIR fuel element was modeled in three dimensions; coupling to adjacent plates and channels was accounted for by using periodic boundary conditions. The standard k- turbulence model was used in simulating turbulent flow with conjugate heat transfer. The COMSOL models were developed to be fully parameterized to allow assessing impacts of fuel fabrication tolerances and uncertainties related to low enriched uranium (LEU) fuel design and reactor operating parameters. Heat source input for the simulations was obtained from separate Monte Carlo N Particle calculations for the axially non-contoured LEU fuel designs at the beginning of the reactor cycle. Mesh refinement studies have been performed to calibrate the models against the pressure drop measured across the HFIR core.

  5. 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)

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

    SciTech Connect

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

    2010-12-23

    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 (k{sub eff}) is in almost linear relations with the change of the fuel volume to coolant ratio.

  7. 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.

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

    PubMed

    Laurenzis, A; Heits, H; Wübker, S; 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 (m3pb). 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-3pb h-1 at a load of 150 g of toluene m-3pb h-1. Such a removal rate with a trickle-bed reactor was not reported before.

  9. Growth of plant root cultures in liquid- and gas-dispersed reactor environments.

    PubMed

    McKelvey, S A; Gehrig, J A; Hollar, K A; Curtis, W R

    1993-01-01

    The growth of Agrobacterium transformed "hairy root" cultures of Hyoscyamus muticus was examined in various liquid- and gas-dispersed bioreactor configurations. Reactor runs were replicated to provide statistical comparisons of nutrient availability on culture performance. Accumulated tissue mass in submerged air-sparged reactors was 31% of gyratory shake-flask controls. Experiments demonstrate that poor performance of sparged reactors is not due to bubble shear damage, carbon dioxide stripping, settling, or flotation of roots. Impaired oxygen transfer due to channeling and stagnation of the liquid phase are the apparent causes of poor growth. Roots grown on a medium-perfused inclined plane grew at 48% of gyratory controls. This demonstrates the ability of cultures to partially compensate for poor liquid distribution through vascular transport of nutrients. A reactor configuration in which the medium is sprayed over the roots and permitted to drain down through the root tissue was able to provide growth rates which are statistically indistinguishable (95% T-test) from gyratory shake-flask controls. In this type of spray/trickle-bed configuration, it is shown that distribution of the roots becomes a key factor in controlling the rate of growth. Implications of these results regarding design and scale-up of bioreactors to produce fine chemicals from root cultures are discussed.

  10. 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.

  11. Studies on disintegrating spherical fuel elements of high temperature gas-cooled reactor by a electrochemical method

    NASA Astrophysics Data System (ADS)

    Tian, Lifang; Wen, Mingfen; Chen, Jing

    2013-01-01

    Spherical fuel elements of a high temperature gas-cooled reactor were disintegrated through a electrochemical method with NaNO3 as electrolyte. The X-ray diffraction spectra and total carbon contents of the graphite fragments were determined, and the results agreed with those from simulated fuel elements. After conducting the characterization analysis and the leaching experiment of coated fuel particles, the uranium concentrations of leaching solutions and spent electrolyte were found to be at background levels. The results demonstrate the effectiveness of the improved electrochemical method with NaNO3 as electrolyte in disintegrating the unirradiated fuel elements without any damage to the coated fuel particles. Moreover, the method avoided unexpected radioactivity contamination to the graphite matrix and spent electrolyte.

  12. Special power supply and control system for the gas-cooled fast reactor-core flow test loop

    SciTech Connect

    Hudson, T.L.

    1981-09-01

    The test bundle in the Gas-Cooled Fast Reactor-Core Flow Test Loop (GCFR-CFTL) requires a source of electrical power that can be controlled accurately and reliably over a wide range of steady-state and transient power levels and skewed power distributions to simulate GCFR operating conditions. Both ac and dc power systems were studied, and only those employing silicon-controlled rectifiers (SCRs) could meet the requirements. This report summarizes the studies, tests, evaluations, and development work leading to the selection. it also presents the design, procurement, testing, and evaluation of the first 500-kVa LMPL supply. The results show that the LMPL can control 60-Hz sine wave power from 200 W to 500 kVA.

  13. Boosting Gas Involved Reactions at Nanochannel Reactor with Joint Gas-Solid-Liquid Interfaces and Controlled Wettability.

    PubMed

    Mi, Li; Yu, Jiachao; He, Fei; Jiang, Ling; Wu, Yafeng; Yang, Lijun; Han, Xiaofeng; Li, Ying; Liu, Anran; Wei, Wei; Zhang, Yuanjian; Tian, Ye; Liu, Songqin; Jiang, Lei

    2017-08-02

    The low solubility of gases in aqueous solution is the major kinetic limitation of reactions that involve gases. To address this challenge, we report a nanochannel reactor with joint gas-solid-liquid interfaces and controlled wettability. As a proof of concept, a porous anodic alumina (PAA) nanochannel membrane with different wettability is used for glucose oxidase (GOx) immobilization, which contacts with glucose aqueous solution on one side, while the other side gets in touch with the gas phase directly. Interestingly, it is observed that the O2 could participate in the enzymatic reaction directly from gas phase through the proposed nanochannels, and a hydrophobic interface is more favorable for the enzymatic reaction due to the rearrangement of GOx structure as well as the high gas adhesion. As a result, the catalytic efficiency of GOx in the proposed interface is increased up to 80-fold compared with that of the free state in traditional aqueous air-saturated electrolyte. This triphase interface with controlled wettability can be generally applied to immobilize enzymes or catalysts with gas substrates for high efficiency.

  14. Simulation of two-phase flow using lattice gas automata methods

    SciTech Connect

    Tsumaya, Akira; Ohashi, Hirotada; Akiyama, Mamoru

    1996-08-01

    Two-phase flow simulation has been primarily based on experimental data in the sense that constitutive relations necessary for solving fundamental equations are experimentally determined. This assures validity of simulation of two-phase flow within the experimental conditions, but it is difficult to predict the behavior of two-phase flow under extreme or complex conditions which occur, for example, in severe accidents of nuclear reactors. Lattice gas automaton (LGA) simulation has recently attracted attention as a method for numerical simulation of multi phase flow. The authors extend phase-separation LGA models and develop methods for two-phase flow simulation. First, they newly added a flow model to the immiscible lattice gas model and applied it to two-dimensional Poiseuille flow. They obtained a result looking like lubricated pipelining of crude oil with water. Also, considering the gravity effect, they introduced a buoyancy force into the liquid-gas model. As a result, they demonstrated that gas bubbles of various diameters rise and gradually coalesce each other turning into larger bubbles. Using these newly developed LGA models, they succeeded in simulating various flow patterns of two-phase flow.

  15. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    NASA Astrophysics Data System (ADS)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

  16. Ab initio simulation of radiation damage in nuclear reactor pressure vessel materials

    NASA Astrophysics Data System (ADS)

    Watts, Daniel; Finkenstadt, Daniel

    2012-02-01

    Using Kinetic Monte Carlo we developed a code to study point defect hopping in BCC metallic alloys using energetics and attempt frequencies calculated using VASP, an electronic structure software package. Our code provides a way of simulating the effects of neutron radiation on potential reactor materials. Specifically we will compare the Molybdenum-Chromium alloy system to steel alloys for use in nuclear reactor pressure vessels.

  17. 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.

  18. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    NASA Astrophysics Data System (ADS)

    Ilham, Muhammad; Su'ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  19. 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

  20. Biological reduction of chlorate in a gas-lift reactor using hydrogen as an energy source.

    PubMed

    Kroon, A G M; van Ginkel, C G

    2004-01-01

    Chlorate release into the environment occurs with its manufacture and use. Biological reduction of chlorate offers an attractive option to decrease this release. A hydrogen gas-lift reactor with microorganisms attached to pumice particles was used for the treatment of wastewater containing high concentrations of chlorate. The microorganisms used chlorate as an electron acceptor and hydrogen gas as a reducing agent. After a start-up period of only a few weeks, chlorate reduction rates of 3.2 mmol L(-1) h(-1) were achieved during continuous operation. During this period, a hydrogen consumption rate of 14.5 mmol L(-1) h(-1) was observed. Complete removal of chlorate was maintained at hydraulic retention times of 6 h. This study clearly demonstrates the potential of hydrogen gas-lift bioreactors for the treatment of chlorate-containing waste streams.

  1. 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

  2. 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.

  3. Hot-Gas Filter Testing with a Transport Reactor Development Unit

    SciTech Connect

    Swanson, M.L.; Ness, R.O., Jr.

    1996-12-31

    The objective of the hot-gas cleanup (HGC) work on the transport reactor demonstration unit (TRDU) located at the Environmental Research Center is to demonstrate acceptable performance of hot-gas filter elements in a pilot-scale system prior to long-term demonstration tests. The primary focus of the experimental effort in the 2-year project will be the testing of hot- gas filter elements as a function of particulate collection efficiency, filter pressure differential, filter cleanability, and durability during relatively short-term operation (100-200 hours). A filter vessel will be used in combination with the TRDU to evaluate the performance of selected hot- gas filter elements under gasification operating conditions. This work will directly support the Power Systems Development Facility utilizing the M.W. Kellogg transport reactor located at Wilsonville, Alabama and indirectly the Foster Wheeler advanced pressurized fluid-bed combustor, also located at Wilsonville and the Clean Coal IV Pinon Pine IGCC Power Project. This program has a phased approach involving modification and upgrades to the TRDU and the fabrication, assembly, and operation of a hot-gas filter vessel (HGFV) capable of operating at the outlet design conditions of the TRDU. Phase 1 upgraded the TRDU based upon past operating experiences. Additions included a nitrogen supply system upgrade, upgraded LASH auger and 1807 coal feed lines, the addition of a second pressurized coal feed hopper and a dipleg ash hopper, and modifications to spoil the performance of the primary cyclone. Phase 2 included the HGFV design, procurement, and installation. Phases 3 through 5 consist of 200-hour hot-gas filter tests under gasification conditions using the TRDU at temperatures of 540-650{degrees}C (1000-1200{degrees}F), 9.3 bar, and face velocities of 1.4, 2. and 3.8 cm/s, respectively. The increased face velocities are achieved by removing candles between each test.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Code and model extensions of the THATCH code for modular high temperature gas-cooled reactors

    SciTech Connect

    Kroger, P.G.; Kennett, R.J. )

    1993-05-01

    This report documents several model extensions and improvements of the THATCH code, a code to model thermal and fluid flow transients in High Temperature Gas-Cooled Reactors. A heat exchanger model was added, which can be used to represent the steam generator of the main Heat Transport System or the auxiliary Shutdown Cooling System. This addition permits the modeling of forced flow cooldown transients with the THATCH code. An enhanced upper head model, considering the actual conical and spherical shape of the upper plenum and reactor upper head was added, permitting more accurate modeling of the heat transfer in thisregion. The revised models are described, and the changes and addition to the input records are documented.

  10. 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.

  11. Preliminary Study of Turbulent Flow in the Lower Plenum of a Gas-Cooled Reactor

    SciTech Connect

    D.P. Guillen; H.M.McIlroy

    2007-09-01

    A preliminary study of the turbulent flow in a scaled model of a portion of the lower plenum of a gas-cooled advanced reactor concept has been conducted. The reactor is configured such that hot gases at various temperatures exit the coolant channels in the reactor core, where they empty into a lower plenum and mix together with a crossflow past vertical cylindrical support columns, then exit through an outlet duct. An accurate assessment of the flow behavior will be necessary prior to final design to ensure that material structural limits are not exceeded. In this work, an idealized model was created to mimic a region of the lower plenum for a simplified set of conditions that enabled the flow to be treated as an isothermal, incompressible fluid with constant properties. This is a first step towards assessing complex thermal fluid phenomena in advanced reactor designs. Once such flows can be computed with confidence, heated flows will be examined. Experimental data was obtained using three-dimensional Particle Image Velocimetry (PIV) to obtain non-intrusive flow measurements for an unheated geometry. Computational fluid dynamic (CFD) predictions of the flow were made using a commercial CFD code and compared to the experimental data. The work presented here is intended to be scoping in nature, since the purpose of this work is to identify improvements that can be made to subsequent computations and experiments. Rigorous validation of computational predictions will eventually be necessary for design and analysis of new reactor concepts, as well as for safety analysis and licensing calculations.

  12. 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.

  13. 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.

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

    SciTech Connect

    Turinsky, Paul J.; Kothe, Douglas B.

    2016-05-15

    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

  15. 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.

  16. 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.

  17. A two-dimensional numerical model of gas mixing and deposition in a rotating disk CVD reactor

    SciTech Connect

    Winters, W.S.; Evans, G.H.; Greif, R.

    1996-04-01

    Gas phase transport with mixing and surface chemistry is studied in an axisymmetric, isothermal rotating disk chemical vapor deposition reactor. A simple one-step surface reaction is used to model deposition of gallium on the rotating surface. Partitioning of the inlet flow into separate gas streams of different species can lead to nonuniform deposition on the growth surface. The nonuniformity is caused by incomplete radial diffusion of gas species; depending on reactor temperature and pressure it can be worsened by large buoyant flow instabilities. The nonuniformity is relatively insensitive to the magnitude of the specified sticking coefficient.

  18. 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.

  19. 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.

  20. 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)

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. A utility assessment of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    SciTech Connect

    Bliss, H.E.; Grier, C.A. ); Crews, M.R. ); Fernandez, R.T.; Heard, J.W.; Hinkle, W.D. ); Pschirer, D.M.; Sharpe, R.O. )

    1991-01-01

    A team of electric utility representatives conducted an in-depth, independent evaluation of the current Modular High Temperature Gas-Cooled Reactor (MHTGR) design. The emphasis was on the fuel design with respect to safety, the licensability of the proposed containment concept, refueling operations and equipment, spent fuel storage capacity, staffing projections, and the economic competitiveness. Specific comments and recommendations are provided as a contribution towards enhancing the MHTGR design, licensability and acceptance from a utility's view. Individual sections have been indexed separately for inclusion on the data base.

  7. Method for fabricating wrought components for high-temperature gas-cooled reactors and product

    DOEpatents

    Thompson, Larry D.; Johnson, Jr., William R.

    1985-01-01

    A method and alloys for fabricating wrought components of a high-temperature gas-cooled reactor are disclosed. These wrought, nickel-based alloys, which exhibit strength and excellent resistance to carburization at elevated temperatures, include aluminum and titanium in amounts and ratios to promote the growth of carburization resistant films while preserving the wrought character of the alloys. These alloys also include substantial amounts of molybdenum and/or tungsten as solid-solution strengtheners. Chromium may be included in concentrations less than 10% to assist in fabrication. Minor amounts of carbon and one or more carbide-forming metals also contribute to high-temperature strength.

  8. A microscale multi-inlet vortex nanoprecipitation reactor: Turbulence measurement and simulation

    NASA Astrophysics Data System (ADS)

    Cheng, Janine Chungyin; Olsen, Michael G.; Fox, Rodney O.

    2009-05-01

    Microscale reactors capable of generating turbulent flow are used in Flash NanoPrecipitation, an approach to produce functional nanoparticles with unique optical, mechanical and chemical properties. Microreactor design and optimization could be greatly enhanced by developing reliable computational models of the nanoprecipitation process. A microscale multi-inlet vortex nanoprecipitation reactor was investigated using microscopic particle image velocimetry and computational fluid dynamics. Velocity data such as the mean velocity and turbulent kinetic energy displayed good agreement between experiment and simulation over flow conditions ranging from fully laminar to turbulent, demonstrating the accuracy of the simulation model over the entire turbulent transition range.

  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. 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

  11. 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.

  12. 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.

  13. Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

    1976-01-01

    The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

  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. Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

    1976-01-01

    The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

  16. 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

  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. 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.

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

    DOEpatents

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    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.

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

    DOEpatents

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pocoima, CA; Benander, Robert E [Pacoima, CA

    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.

  2. A chemical reactor network for oxides of nitrogen emission prediction in gas turbine combustor

    NASA Astrophysics Data System (ADS)

    Hao, Nguyen Thanh

    2014-06-01

    This study presents the use of a new chemical reactor network (CRN) model and non-uniform injectors to predict the NOx emission pollutant in gas turbine combustor. The CRN uses information from Computational Fluid Dynamics (CFD) combustion analysis with two injectors of CH4-air mixture. The injectors of CH4-air mixture have different lean equivalence ratio, and they control fuel flow to stabilize combustion and adjust combustor's equivalence ratio. Non-uniform injector is applied to improve the burning process of the turbine combustor. The results of the new CRN for NOx prediction in the gas turbine combustor show very good agreement with the experimental data from Korea Electric Power Research Institute.

  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. Insights into large-scale cell-culture reactors: II. Gas-phase mixing and CO₂ stripping.

    PubMed

    Sieblist, Christian; Hägeholz, Oliver; Aehle, Mathias; Jenzsch, Marco; Pohlscheidt, Michael; Lübbert, Andreas

    2011-12-01

    Most discussions about stirred tank bioreactors for cell cultures focus on liquid-phase motions and neglect the importance of the gas phase for mixing, power input and especially CO(2) stripping. Particularly in large production reactors, CO(2) removal from the culture is known to be a major problem. Here, we show that stripping is mainly affected by the change of the gas composition during the movement of the gas phase through the bioreactor from the sparger system towards the headspace. A mathematical model for CO(2)-stripping and O(2)-mass transfer is presented taking gas-residence times into account. The gas phase is not moving through the reactor in form of a plug flow as often assumed. The model is validated by measurement data. Further measurement results are presented that show how the gas is partly recirculated by the impellers, thus increasing the gas-residence time. The gas-residence times can be measured easily with stimulus-response techniques. The results offer further insights on the gas-residence time distributions in stirred tank reactors. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  5. 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.

  6. 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.

  7. Use of albedo for neutron reflector regions in reactor core 3-D simulations

    NASA Astrophysics Data System (ADS)

    Mohanakrishnan, P.

    1989-10-01

    In this paper we present two new simplified schemes for the application of the albedo concept of replacing the reflector in 3-D reactor core simulations. Both involve the numerical derivation of albedoes from the fluxes at the core- (blanket-) reflector interface obtained from sample calculations including the reflector. Diffusion theory is used for core calculations in both cases. In the first scheme a new method for "diagonalising" the albedo matrix is demonstrated. This achieves easy applicability of the albedo parameters in core simulations of a fast breeder reactor core, resulting in significant savings in computing efforts. The second scheme, applied to light water reactors, achieves better accuracy in core periphery power predictions with the use of only uniform coarse meshes throughout the core and the numerically derived albedoes.

  8. 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. Copyright © 2011 Elsevier Ltd. All rights reserved.

  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

    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.

  10. 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.

  11. Large-Scale Simulation of Nuclear Reactors: Issues and Perspectives

    SciTech Connect

    Merzari, Elia; Obabko, Aleks; Fischer, Paul; Halford, Noah; Walker, Justin; Siegel, Andrew; Yu, Yiqi

    2015-01-01

    Numerical simulation has been an intrinsic part of nuclear engineering research since its inception. In recent years a transition is occurring toward predictive, first-principle-based tools such as computational fluid dynamics. Even with the advent of petascale computing, however, such tools still have significant limitations. In the present work some of these issues, and in particular the presence of massive multiscale separation, are discussed, as well as some of the research conducted to mitigate them. Petascale simulations at high fidelity (large eddy simulation/direct numerical simulation) were conducted with the massively parallel spectral element code Nek5000 on a series of representative problems. These simulations shed light on the requirements of several types of simulation: (1) axial flow around fuel rods, with particular attention to wall effects; (2) natural convection in the primary vessel; and (3) flow in a rod bundle in the presence of spacing devices. The focus of the work presented here is on the lessons learned and the requirements to perform these simulations at exascale. Additional physical insight gained from these simulations is also emphasized.

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

    DOE PAGES

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

    2009-01-01

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

  13. 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.

  14. Transport simulation of ITER (International Thermonuclear Engineering Reactor) startup

    SciTech Connect

    Attenberger, S.E.; Houlberg, W.A.

    1989-01-01

    The present International Thermonuclear Engineering Reactor (ITER) reference configurations are the Technology Phase,'' in which the plasma current is maintained noninductively at a subignition density, and the Physics Phase,'' which is ignited but requires inductive maintenance of the current. The WHIST 1.5-D transport code is used to evaluate the volt-second requirements of both configurations. A slow current ramp (60-80's) is required for fixed-radius startup in ITER to avoid hollow current density profiles. To reach the operating point requires about 203 V{center dot}s for the Technology Phase (18 MA) and about 270 V{center dot}s for the Physics Phase (22 MA). The resistive losses can be reduced with expanding-radius startup. 5 refs., 4 figs.

  15. 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.

  16. 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)

  17. 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

  18. 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.

  19. CFD coupled kinetic modeling and simulation of hot wall vertical tubular reactor for deposition of SiC crystal from MTS

    NASA Astrophysics Data System (ADS)

    Mollick, P. K.; Venugopalan, R.; Srivastava, D.

    2017-10-01

    Chemical Vapor Deposition (CVD) process is generally carried out in a hot wall reactor of vertical or horizontal type keeping the substrate inside the chamber on which deposition is targeted. Present study is focused to explain the role of hydrodynamics and temperature conditions on the overall coating rates inside a hot wall vertical tubular reactor. Deposition of β-Silicon Carbide crystals from Methytricholorosilane catalyzed by hydrogen is modeled here considering growth kinetics which can be successfully described - using only two steps. Finite Element Method based simulation is performed to obtain the flow and temperature profiles inside the hot wall reactor. Model equations for kinetics are derived in differential form based on mass balance considering transport of species. Kinetic parameters were approximated comparing the experimentally found coating rates as reported earlier. Present model is seen to fit reasonably well for the wide variation of gas flow rates as well as temperature. Apart from the flow rates of total fluid at inlet and initial wall temperature of reactor, sample position and the inlet diameter of the reactor are found to be key important parameters for the desired coating to take place. Model prediction thus can provide better knowledge in order to carefully choose process parameters in designing the reactor for achieving optimized deposition rates by CVD with desired properties.

  20. FLASH simulations of 120MJ target explosions in LIFE reactor chamber

    NASA Astrophysics Data System (ADS)

    Sacks, Ryan; Moses, Gregory; Fatenejad, Milad

    2012-10-01

    The LIFE conceptual reactor designfootnotetextMoses, E.I., Ignition on the National Ignition Facility: a path towards inertial fusion energy, Nucl. Fusion 49 104022 is a 12 m diameter reaction chamber with a steel first wall. The chamber is filled with 6 μg/cm^3 Xenon gas to protect the wall from fusion burn products. Indirect drive 120 MJ fusion targets are shot at 13 Hz repetition rate. For purposes of simulating the target explosion the target is approximated as a 1 g lead hohlraum. Fusion burn product energy is added to the Pb in a 100 ps flattop source at a rate of 12 MJ / 100 ps. The additional 13 MJ of fusion energy is assumed to be radiated as prompt x-rays. The resulting spherical micro-explosion of the heated Pb target into the surrounding Xe is simulated in 2D using the FLASH radiation hydrodynamics code. The FLASH codefootnotetextFryxell, B., Olson, K. et al.,FLASH: An Adaptive Mesh Hydrodynamics Code for Modeling Astrophysical Thermonuclear Flashes, Astro. Journal Sup. Series., 131, 273 is an AMR block-structured, parallel scalable radiation hydrodynamics code. FLASH has separate electron and ion temperatures and single group or multi-group radiation diffusion. Shock generation in the Xe and mixing of the Pb and Xe behind the shock due to Rayleigh-Taylor instability is investigated. Comparison with results from the 1D BUCKY radiation hydrodynamics code will be presented. This work was supported by Lawrence Livermore National Laboratory under contract number B587835.

  1. 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.

  2. 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

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

    DOE PAGES

    Gaston, Derek R.; Permann, Cody J.; Peterson, John W.; ...

    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

  4. Simulation of Nanoparticle Production in Premixed Aerosol Flow Reactors by Interfacing Fluid Mechanics and Particle Dynamics

    NASA Astrophysics Data System (ADS)

    Schild, A.; Gutsch, A.; Mühlenweg, H.; Pratsinis, S. E.

    1999-06-01

    The interaction of fluid mechanics and particle dynamics at the very early stages of flame synthesis largely affects the characteristics of the product powder. Detailed simulations provide a better understanding of these processes, which take place in a few milliseconds, and offer the possibility to influence the product characteristics by intelligent selection of the process parameters. The present paper reports on the simulation of titania powder formation by TiCl4 oxidation in an aerosol flow reactor. A commercially available fluid mechanics code is used for the detailed calculation of the fluid flow and the chemical reaction at non-isothermal conditions. This code is then interfaced with a model for aggregate particle dynamics neglecting the spread of the particle size distribution. The simulation shows the onset of the particle formation in the reactor and calculates the dynamic evolution of the aggregate particle size, number of primary particles per aggregate and the specific surface area throughout the reactor. The presented, newly developed calculation technique allows for the first time the simulation of particle formation processes under the authentic, complex conditions as found in actual aerosol reactors.

  5. 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.

  6. 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.

  7. A pilot-scale jet bubbling reactor for wet flue gas desulfurization with pyrolusite.

    PubMed

    Su, Shi-jun; Zhu, Xiao-fan; Liu, Yong-jun; Jiang, Wen-ju; Jin, Yan

    2005-01-01

    MnO2 in pyrolusite can react with SO2 in flue gas and obtain by-product MnSO4 x H2O. A pilot scale jet bubbling reactor was applied in this work. Different factors affecting both SO2 absorption efficiency and Mn2+ extraction rate have been investigated, these factors include temperature of inlet gas flue, ration of liquid/solid mass flow rate (L/S), pyrolusite grade, and SO2 concentration in the inlet flue gas. In the meantime, the procedure of purification of absorption liquid was also discussed. Experiment results indicated that the increase of temperature from 30 to 70 K caused the increase of SO2 absorption efficiency from 81.4% to 91.2%. And when SO2 concentration in the inlet flue gas increased from 500 to 3000 ppm, SO2 absorption efficiency and Mn2+ extraction rate decreased from 98.1% to 82.2% and from 82.8% to 61.7%, respectively. The content of MnO2 in pyrolusite had a neglectable effect on SO2 absorption efficiency. Low L/S was good for both removal of SO2 and Mn2+ extraction. The absorption liquid was filtrated and purified to remove Si, Mg, Ca, Fe, Al and heavy metals, last product MnSO4 x H2O was obtained which quality could reach China GB1622-86, the industry grade standards.

  8. Hot-gas filter testing with a transport reactor development unit

    SciTech Connect

    Swanson, M.L.; Ness, R.O. Jr.; Mann, M.D.

    1996-12-31

    The objective of the hot-gas cleanup work on the transport reactor development unit (TRDU) located at the Energy and Environmental Research Center is to demonstrate acceptable performance of hot-gas filter elements in a pilot-scale system prior to long-term demonstration tests. The primary focus of the experimental effort in the 2-year project will be the testing of hot-gas filter elements as a function of particulate collection efficiency, filter pressure differential, filter cleanability, and durability during relatively short-term operation (100--200 hours). The selected hot-gas filter elements currently being tested include a Fibroplate{trademark} ceramic tube sheet and Fibrosic{trademark} candle filters from industrial Filter and Pump Mfg.; silicon carbide fiber ceramic candle filters from 3M; and metal and Vitropore ceramic candle filters from Pall Advanced Separation Systems. These elements will be used in three 200-hour filter tests under gasification conditions using the TRDU at filter temperatures of approximately 540 C at 9.3 bar and face velocities ranging from 1.4 to 3.8 cm/s. Changes in filter face velocity are achieved by removing six candles between each test. Results from the first 200-hour test are presented.

  9. Hydrogen production from simulated hot coke oven gas by using oxygen-permeable ceramics

    SciTech Connect

    Hongwei Cheng; Yuwen Zhang; Xionggang Lu; Weizhong Ding; Qian Li

    2009-01-15

    Hydrogen production from simulated hot coke oven gas (HCOG) was investigated in a BaCo{sub 0.7}Fe{sub 0.2}Nb{sub 0.1}O{sub 3-{delta}} (BCFNO) membrane reactor combined with a Ni/Mg(Al)O catalyst by the partial oxidation with toluene as a model tar compound under atmospheric pressure. The reaction results indicated that toluene was completely converted to H{sub 2} and CO in the catalytic reforming of the simulated HCOG in the temperature range from 825 to 875{sup o}C. Both thermodynamically predicated values and experimental data showed that the selective oxidation of toluene took precedence over that of CH{sub 4} in the reforming reaction. At optimized reaction conditions, the dense oxygen-permeable membrane has an oxygen permeation flux around 12.3 mL cm{sup -2} min{sup -1}, and a CH{sub 4} conversion of 86%, a CO{sub 2} conversion of 99%, a H{sub 2} yield of 88%, and a CO yield of 87% have been achieved. When the toluene and methane were reformed, the amount of H{sub 2} in the reaction effluent gas was about 2 times more than that of original H{sub 2} in simulated HCOG. The results reveal that it is feasible for hydrogen production from HCOG by reforming hydrocarbon compounds in a ceramic oxygen-permeable membrane reactor. 27 refs., 10 figs., 3 abs.

  10. Automated in Situ Measurement of Gas Solubility in Liquids with a Simple Tube-in-Tube Reactor.

    PubMed

    Zhang, Jisong; Teixeira, Andrew R; Zhang, Haomiao; Jensen, Klavs F

    2017-08-15

    Data on the solubilities of gases in liquids are foundational for assessing a variety of multiphase separations and gas-liquid reactions. Taking advantage of the tube-in-tube reactor design built with semipermeable Teflon AF-2400 tubes, liquids can be rapidly saturated without direct contacting of gas and liquid. The gas solubility can be determined by performing steady-state flux balances of both the gas and liquid flowing into the reactor system. Using this type of reactor, a fully automated strategy has been developed for the rapid in situ measurement of gas solubilities in liquids. The developed strategy enables precise gas solubility measurements within 2-5 min compared with 4-5 h using conventional methods. This technique can be extended to the discrete multipoint steady-state and continuous ramped-multipoint data acquisition methods. The accuracy of this method has been validated against several gas-liquid systems, showing less than 2% deviation from known values. Finally, this strategy has been extended to measure the temperature dependence of gas solubilities in situ and to estimate the local enthalpy of dissolution across a defined temperature range.

  11. Decomposition of simulated odors in municipal wastewater treatment plants by a wire-plate pulse corona reactor.

    PubMed

    Ruan, Jian-Jun; Li, Wei; Shi, Yao; Nie, Yong; Wang, Xin; Tan, Tian-En

    2005-04-01

    Decomposition of simulated odors in municipal wastewater treatment plants was investigated experimentally by a wire-plate pulse corona reactor. A new type of high pulse voltage source with a thyratron switch and a Blumlein pulse forming network (BPFN) was adopted in our experiments, and the testing malodorants were ammonia, ethanethiol and tri-methyl amine, respectively. The maximum output power of the pulse voltage source and the maximum peak voltage were 1 kW and 100 kV. The experiments were conducted at the gas-flow rate of 4.0-23.0 m3 h(-1). Important parameters, including peak voltage, pulse frequency, capacitance (inductance) of the BPFN, gas-flow rate, initial concentration, which influenced on the removal efficiency, were investigated. The results show that the odors can be treated effectively. Almost 100% removal efficiency was obtained for 32 mg m(-3) ammonia at the gas-flow rate of 4.0 m(3) h(-1). The maximum removal efficiencies of 85 mg m(-3) ethanethiol and 750 mg m(-3) tri-methyl amine at 10.0 m(3) h(-1) were 98% and 91%, respectively. The energy yield of 110 mg m(-3) ammonia was 2.99 g kWh(-1) when specific energy density was 106 Jl(-1). In the cases of ammonia, ethanethiol and tri-methyl amine removal, ozone and nitrogen oxides were observed in the exit gas. The carbon and sulfur elements of ethanethiol and tri-methyl amine were mainly converted to carbon dioxide, carbon monoxide and sulfur dioxide. Moreover, the ammonium nitrates and sulfur were discovered in the reactor.

  12. Experimental and numerical investigations of high temperature gas heat transfer and flow in a VHTR reactor core

    NASA Astrophysics Data System (ADS)

    Valentin Rodriguez, Francisco Ivan

    High pressure/high temperature forced and natural convection experiments have been conducted in support of the development of a Very High Temperature Reactor (VHTR) with a prismatic core. VHTRs are designed with the capability to withstand accidents by preventing nuclear fuel meltdown, using passive safety mechanisms; a product of advanced reactor designs including the implementation of inert gases like helium as coolants. The present experiments utilize a high temperature/high pressure gas flow test facility constructed for forced and natural circulation experiments. This work examines fundamental aspects of high temperature gas heat transfer applied to VHTR operational and accident scenarios. Two different types of experiments, forced convection and natural circulation, were conducted under high pressure and high temperature conditions using three different gases: air, nitrogen and helium. The experimental data were analyzed to obtain heat transfer coefficient data in the form of Nusselt numbers as a function of Reynolds, Grashof and Prandtl numbers. This work also examines the flow laminarization phenomenon (turbulent flows displaying much lower heat transfer parameters than expected due to intense heating conditions) in detail for a full range of Reynolds numbers including: laminar, transition and turbulent flows under forced convection and its impact on heat transfer. This phenomenon could give rise to deterioration in convection heat transfer and occurrence of hot spots in the reactor core. Forced and mixed convection data analyzed indicated the occurrence of flow laminarization phenomenon due to the buoyancy and acceleration effects induced by strong heating. Turbulence parameters were also measured using a hot wire anemometer in forced convection experiments to confirm the existence of the flow laminarization phenomenon. In particular, these results demonstrated the influence of pressure on delayed transition between laminar and turbulent flow. The heat

  13. 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.

  14. A Microscale Multi-Inlet Vortex Nanoprecipitation Reactor: Turbulence Measurement and Simulation

    NASA Astrophysics Data System (ADS)

    Chungyin Cheng, Janine; Olsen, Michael G.; Fox, Rodney O.

    2009-11-01

    Microreactors capable of generating turbulent flow are used in Flash NanoPrecipitation, a novel approach to produce functional nanoparticles. Microreactor design and optimization can be greatly enhanced by developing reliable computational models. Scalar mixing in a multi-inlet micro-vortex reactor has been previously studied and a RANS simulation with a scalar mixing model had been successfully validated for fully turbulent flow. To further validate the flow field, the reactor, operated in flow regimes from laminar to turbulent, was investigated using microscopic particle image velocimetry and computational fluid dynamics. In the experiments, instantaneous velocity fields were recorded on three different planes in the reactor and flow statistics such as the mean velocity and turbulent kinetic energy were computed for comparison with flow simulations. Laminar simulations were performed for the low-Reynolds-number cases, and large-eddy simulations (LES) were performed for higher-Reynolds-number cases. The simulation results were in good agreement with experiments for all cases, demonstrating the accuracy of the simulation models in the laminar-turbulent transition range.

  15. 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.

  16. 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

  17. 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.

  18. Size limitations for microwave cavity to simulate heating of blanket material in fusion reactor

    SciTech Connect

    Wolf, D.

    1987-01-01

    The power profile in the blanket material of a nuclear fusion reactor can be simulated by using microwaves at 200 MHz. Using these microwaves, ceramic breeder materials can be thermally tested to determine their acceptability as blanket materials without entering a nuclear fusion environment. A resonating cavity design is employed which can achieve uniform cross sectional heating in the plane transverse to the neutron flux. As the sample size increases in height and width, higher order modes, above the dominant mode, are propagated and destroy the approximation to the heating produced in a fusion reactor. The limits at which these modes develop are determined in the paper.

  19. Hot halo gas in numerical simulations of galaxy mergers

    NASA Astrophysics Data System (ADS)

    Sinha, Manodeep

    Galaxy merger simulations have explored the behavior of gas within a galactic disk, yet the dynamics of hot gas within the galaxy halo has been neglected. We report on the results of high-resolution hydrodynamic simulations of colliding galaxies with hot halo gas. We explore a range of mass ratios, gas fractions and orbital configurations to constrain the shocks and the dynamics of the gas within the progenitor halos. We find that: (i) A strong shock is produced in the galaxy halos before the first passage, increasing the temperature of the gas by almost an order of magnitude to ˜ 10 6.3 K. (ii) The X-ray luminosity of the shock is strongly dependent on the gas fraction. It is ≳ 1039 erg/s for gas fractions larger than 10%. (iii) We find an analytic fit to the maximum X-ray luminosity of the shock as a function of merger parameters. This fit can be used in semi-analytic recipes for galaxy formation to estimate the total X-ray emission from shocks in merging galaxies. (iv) The hot diffuse gas in the simulation also produces X-ray luminosities as large as 1042 erg/s. This contributes to the total X-ray background in the Universe. (v) ˜ 10--20% of the initial gas mass is unbound from the galaxies for equal-mass mergers, while 3--5% of the gas mass is released for the 3:1 and 10:1 mergers. This unbound gas ends up far from the galaxy and can be a feasible mechanism for metal enrichment of the IGM. We use an analytical halo merger tree to estimate the fraction of gas mass lost over the history of the Universe.

  20. Flow design and simulation of a gas compression system for hydrogen fusion energy production

    NASA Astrophysics Data System (ADS)

    Avital, E. J.; Salvatore, E.; Munjiza, A.; Suponitsky, V.; Plant, D.; Laberge, M.

    2017-08-01

    An innovative gas compression system is proposed and computationally researched to achieve a short time response as needed in engineering applications such as hydrogen fusion energy reactors and high speed hammers. The system consists of a reservoir containing high pressure gas connected to a straight tube which in turn is connected to a spherical duct, where at the sphere’s centre plasma resides in the case of a fusion reactor. Diaphragm located inside the straight tube separates the reservoir’s high pressure gas from the rest of the plenum. Once the diaphragm is breached the high pressure gas enters the plenum to drive pistons located on the inner wall of the spherical duct that will eventually end compressing the plasma. Quasi-1D and axisymmetric flow formulations are used to design and analyse the flow dynamics. A spike is designed for the interface between the straight tube and the spherical duct to provide a smooth geometry transition for the flow. Flow simulations show high supersonic flow hitting the end of the spherical duct, generating a return shock wave propagating upstream and raising the pressure above the reservoir pressure as in the hammer wave problem, potentially giving temporary pressure boost to the pistons. Good agreement is revealed between the two flow formulations pointing to the usefulness of the quasi-1D formulation as a rapid solver. Nevertheless, a mild time delay in the axisymmetric flow simulation occurred due to moderate two-dimensionality effects. The compression system is settled down in a few milliseconds for a spherical duct of 0.8 m diameter using Helium gas and a uniform duct cross-section area. Various system geometries are analysed using instantaneous and time history flow plots.

  1. 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.

  2. 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.

  3. A Review of Gas-Cooled Reactor Concepts for SDI Applications

    DTIC Science & Technology

    1989-08-01

    Reactor B-2 B.2 NERVA Derivative Reactor B-2 B.3 Pellet-Bed Reactor B-5 B.4 PLUTO Derivative Reactor B-7 B.5 NERVA/ PLUTO Hybrid Reactor...Method °-2 D. 2 Burst-Mode Mass Estimates D-3 D.2.1 Pellet-Bed Reactor D-3 D.2.2 PLUTO Derivative Reactor D-3 D.2.3 NERVA/ PLUTO Hybrid Reactor...Reactor D-5 D.2.9 Foam-Fuel Reactor D-5 D. 3 MMWSS-Mode Mass Estimates D-6 D. 3 .1 Pellet Bed Reactor D-6 D.3.2 PLUTO Derivative D-6 D.3.3

  4. 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.

  5. 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

  6. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect

    S. B. Grover

    2007-05-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (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 new 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. The 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 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.1,2 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 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 final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment

  7. 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.

  8. N2O and NO emissions during autotrophic nitrogen removal in a granular sludge reactor--a simulation study.

    PubMed

    Van Hulle, S W H; Callens, J; Mampaey, K E; van Loosdrecht, M C M; Volcke, E I P

    2012-01-01

    This contribution deals with NO and N2O emissions during autotrophic nitrogen removal in a granular sludge reactor. Two possible model scenarios describing this emission by ammonium- oxidizing biomass have been compared in a simulation study of a granular sludge reactor for one-stage partial nitritation--Anammox. No significant difference between these two scenarios was noticed. The influence of the bulk oxygen concentration, granule size, reactor temperature and ammonium load on the NO and N2O emissions has been assessed. The simulation results indicate that emission maxima of NO and N2O coincide with the region for optimal Anammox conversion. Also, most of the NO and N2O are present in the off-gas, owing to the limited solubility of both gases. The size of granules needs to be large enough not to limit optimal Anammox activity, but not too large as this implies an elevated production of N2O. Temperature has a significant influence on N2O emission, as a higher temperature results in a better N-removal efficiency and a lowered N2O production. Statistical analysis of the results showed that there is a strong correlation between nitrite accumulation and N2O production. Further, three regions of operation can be distinguished: a region with high N2O, NO and nitrite concentration; a region with high N2 concentrations and, as such, high removal percentages; and a region with high oxygen and nitrate concentrations. There is some overlap between the first two regions, which is in line with the fact that maximum emission of NO and N2O coincides with the region for optimal Anammox conversion.

  9. Modelling of slaughterhouse solid waste anaerobic digestion: determination of parameters and continuous reactor simulation.

    PubMed

    López, Iván; Borzacconi, Liliana

    2010-10-01

    A model based on the work of Angelidaki et al. (1993) was applied to simulate the anaerobic biodegradation of ruminal contents. In this study, two fractions of solids with different biodegradation rates were considered. A first-order kinetic was used for the easily biodegradable fraction and a kinetic expression that is function of the extracellular enzyme concentration was used for the slowly biodegradable fraction. Batch experiments were performed to obtain an accumulated methane curve that was then used to obtain the model parameters. For this determination, a methodology derived from the "multiple-shooting" method was successfully used. Monte Carlo simulations allowed a confidence range to be obtained for each parameter. Simulations of a continuous reactor were performed using the optimal set of model parameters. The final steady-states were determined as functions of the operational conditions (solids load and residence time). The simulations showed that methane flow peaked at a flow rate of 0.5-0.8 Nm(3)/d/m(reactor)(3) at a residence time of 10-20 days. Simulations allow the adequate selection of operating conditions of a continuous reactor.

  10. 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.

  11. 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.

  12. 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)

  13. Molecular Dynamics Simulations of Gas Transport in Polymer Films

    NASA Astrophysics Data System (ADS)

    Whitley, David; Butler, Simon; Adolf, David

    2010-03-01

    Parallel molecular dynamics simulations have been carried out to determine the permeability of O2 and N2 through polyethylene terephthalate, polypropylene and cis(1-4) polybutadiene. The permeability of both mixed and unmixed gas penetrants is studied within films of these well known gas barrier polymers. Results are obtained either through the solubility and diffusion (i.e. P=D*S) or via the permeability directly. Encouraging results are obtained. Additional analysis focuses on ``unmixed/mixed gas'' intracomparisons of the simulated permeability data in addition to corresponding penetrant and host polymer local dynamics.

  14. Liquid-Gas Mixtures in Contact with Walls: Molecular Simulations

    NASA Astrophysics Data System (ADS)

    Markus Dammer, Stephan

    2005-11-01

    We perform molecular dynamics simulations of liquid-gas mixtures in contact to solid walls. We present results concerning Lennard-Jones systems composed of three particle species, namely liquid, foreign gas, and wall particles, which are frozen on a lattice: (i) Close to the wall we observe a layering of the fluid which becomes more pronounced for increasingly hydrophilic walls. (ii) Close to smooth hydrophobic walls we find a two orders of magnitude increase in the number density of gas, which will favor bubble nucleation. (iii) To characterize the walls, we determined the contact angle by simulations of droplets and compare the result to Laplace's estimate of surface energies.

  15. 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)

  16. 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.

  17. Gas composition influence on ion energy distribution functions in an industrial ICP reactor with biased cathode

    NASA Astrophysics Data System (ADS)

    Peterson, David; Shannon, Steven; Coumou, David; White, Scott

    2016-09-01

    An industrial ICP reactor consisting of a top planar coil and RF biased lower electrode has been characterized using a hairpin resonator probe and gridded ion energy analyzer to measure electron density in the bulk plasma and ion energy distribution function (IEDF) at the surface of the biased cathode. Argon and oxygen were run at constant total flow with 20mTorr downstream pressure control with varying flow ratios between the two gases ranging from 0% to 100% oxygen content. ICP and bias power were adjusted to maintain constant electron density and sheath bias over this mixing matrix at four different setpoints reflecting high density / high bias, high density / low bias, low density / high bias, and low density / low bias. Although the fundamental parameters governing RF sheath behavior were held constant, several trends in ion energy distribution are observed with respect to gas composition (aside from the obvious influence of ion mass) that show considerable variation in measured IEDF particularly that can be attributed to ion collisions in the sheath as well as gas heating variation due to gas composition.

  18. 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.

  19. 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).

  20. Dry Scrubbing of Aluminum Cell Gases: Design and Operating Characteristics of a Novel Gas/Solids Reactor

    NASA Astrophysics Data System (ADS)

    Lamb, W. D.; Reeve, Martin R.; Dethloff, F. H.; Leinum, Magne

    1982-11-01

    Engineering details of a pilot plant reactor are described. It comprises a vertical cylindrical vessel with a tangential bottom gas entry. Countercurrent spiraling gas-solids flow is achieved. Reacted solids can be withdrawn from the bottom or the top using a rising axial gas jet. The reactor was evaluated by testing in a dry scrubber system treating 14,000 m3/h of gas from prebake cells. At inlet concentrations of 30-60 mg/m3 it achieved 99.5% scrubbing efficiency with aluminas of a surface area of 45-80 m2/g at feed rates considerably less than cell requirements. Potential benefits are: 1) control of metal purity by segregation of scrubber catch to selected cells, 2) scrubbing high HF inlet concentrations at full feed rate, and 3) meeting more stringent working environment and stack emission requirements.