Science.gov

Sample records for reactors cycles directs

  1. 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.

  2. METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

    DOEpatents

    Reed, G.A.

    1961-01-10

    A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

  3. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    NASA Technical Reports Server (NTRS)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  4. Hybrid reactors. [Fuel cycle

    SciTech Connect

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  5. Thermonuclear inverse magnetic pumping power cycle for stellarator reactors

    NASA Astrophysics Data System (ADS)

    Ho, D. D. M.; Kulsrud, R. M.

    1985-09-01

    A novel power cycle for direct conversion of alpha-particle energy into electricity is proposed for an ignited plasma in a stellerator reactor. The plasma column is alternately compressed and expanded in minor radius by periodic variation of the toroidal magnetic field strength. As a result of the way a stellarator is expected to work, the plasma pressure during expansion is greater than the corresponding pressure during compression. Therefore, negative work is done on the plasma during a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils, and direct electrical energy is obtained from this voltage. For a typical reactor, the average power obtained from this cycle (with a minor radius compression factor on the order of 50%) can be as much as 50% of the electrical power obtained from the thermonuclear neutrons without compressing the plasma. Thus, if it is feasible to vary the toroidal field strength, the power cycle provides an alternative scheme of energy conversion for a deuterium-tritium fueled reactor. The cycle may become an important method of energy conversion for advanced neutron-lean fueled reactors. By operating two or more reactors in tandem, the cycle can be made self-sustaining.

  6. Thermonuclear inverse magnetic pumping power cycle for stellarator reactor

    DOEpatents

    Ho, Darwin D.; Kulsrud, Russell M.

    1991-01-01

    The plasma column in a stellarator is compressed and expanded alternatively in minor radius. First a plasma in thermal balance is compressed adiabatically. The volume of the compressed plasma is maintained until the plasma reaches a new thermal equilibrium. The plasma is then expanded to its original volume. As a result of the way a stellarator works, the plasma pressure during compression is less than the corresponding pressure during expansion. Therefore, negative work is done on the plasma over a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils. Direct electrical energy is obtained from this voltage. Alternatively, after the compression step, the plasma can be expanded at constant pressure. The cycle can be made self-sustaining by operating a system of two stellarator reactors in tandem. Part of the energy derived from the expansion phase of a first stellarator reactor is used to compress the plasma in a second stellarator reactor.

  7. The Framatome ANP Indirect-Cycle Very High Temperature Reactor

    SciTech Connect

    Copsey, Bernie; Lecomte, Michel; Brinkmann, Gerd; Capitaine, Alain; Deberne, Nicolas

    2004-07-01

    Framatome ANP is developing a Very High Temperature Reactor (VHTR) design, relying on its previous experience with high temperature reactor designs, from its participation in the MODUL and the GT-MHR designs. The Framatome ANP VHTR design is based on an indirect cycle coupled to an 'off-the-shelf' combined cycle gas turbine. Although direct cycle HTR's are being promoted for their high efficiency, preliminary evaluations show that the Framatome ANP design efficiency is on par with a direct cycle while avoiding PGS (Power Generation System) developments and keeping the PGS contamination free. This concept was independently evaluated with sensitivity analysis by EDF. Moreover, the nuclear heat source of the indirect cycle could also be used to qualify the direct cycle components without risk of contamination behind the IHX, thus assisting in the preparation for the later introduction of that technology. Relying to the maximum extent on available technology, the Framatome ANP VHTR plant can demonstrate high-efficiency electricity generation and carbon-free hydrogen production. (authors)

  8. Supercritical Water Reactor Cycle for Medium Power Applications

    SciTech Connect

    BD Middleton; J Buongiorno

    2007-04-25

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency {ge}20%; Steam turbine outlet quality {ge}90%; and Pumping power {le}2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

  9. High power density reactors based on direct cooled particle beds

    NASA Astrophysics Data System (ADS)

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

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  10. Immobilization of Fast Reactor First Cycle Raffinate

    SciTech Connect

    Langley, K. F.; Partridge, B. A.; Wise, M.

    2003-02-26

    This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.

  11. Multi-cycle boiling water reactor fuel cycle optimization

    SciTech Connect

    Ottinger, K.; Maldonado, G.I.

    2013-07-01

    In this work a new computer code, BWROPT (Boiling Water Reactor Optimization), is presented. BWROPT uses the Parallel Simulated Annealing (PSA) algorithm to solve the out-of-core optimization problem coupled with an in-core optimization that determines the optimum fuel loading pattern. However it uses a Haling power profile for the depletion instead of optimizing the operating strategy. The result of this optimization is the optimum new fuel inventory and the core loading pattern for the first cycle considered in the optimization. Several changes were made to the optimization algorithm with respect to other nuclear fuel cycle optimization codes that use PSA. Instead of using constant sampling probabilities for the solution perturbation types throughout the optimization as is usually done in PSA optimizations the sampling probabilities are varied to get a better solution and/or decrease runtime. The new fuel types available for use can be sorted into an array based on any number of parameters so that each parameter can be incremented or decremented, which allows for more precise fuel type selection compared to random sampling. Also, the results are sorted by the new fuel inventory of the first cycle for ease of comparing alternative solutions. (authors)

  12. Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

    SciTech Connect

    Worrall, Andrew; Todosow, Michael

    2016-01-01

    Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include: increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance

  13. Optimization and Comparison of Direct and Indirect Supercritical Carbon Dioxide Power Plant Cycles for Nuclear Applications

    SciTech Connect

    Edwin A. Harvego; Michael G. McKellar

    2011-11-01

    There have been a number of studies involving the use of gases operating in the supercritical mode for power production and process heat applications. Supercritical carbon dioxide (CO2) is particularly attractive because it is capable of achieving relatively high power conversion cycle efficiencies in the temperature range between 550 C and 750 C. Therefore, it has the potential for use with any type of high-temperature nuclear reactor concept, assuming reactor core outlet temperatures of at least 550 C. The particular power cycle investigated in this paper is a supercritical CO2 Recompression Brayton Cycle. The CO2 Recompression Brayton Cycle can be used as either a direct or indirect power conversion cycle, depending on the reactor type and reactor outlet temperature. The advantage of this cycle when compared to the helium Brayton cycle is the lower required operating temperature; 550 C versus 850 C. However, the supercritical CO2 Recompression Brayton Cycle requires an operating pressure in the range of 20 MPa, which is considerably higher than the required helium Brayton cycle operating pressure of 8 MPa. This paper presents results of analyses performed using the UniSim process analyses software to evaluate the performance of both a direct and indirect supercritical CO2 Brayton Recompression cycle for different reactor outlet temperatures. The direct supercritical CO2 cycle transferred heat directly from a 600 MWt reactor to the supercritical CO2 working fluid supplied to the turbine generator at approximately 20 MPa. The indirect supercritical CO2 cycle assumed a helium-cooled Very High Temperature Reactor (VHTR), operating at a primary system pressure of approximately 7.0 MPa, delivered heat through an intermediate heat exchanger to the secondary indirect supercritical CO2 Brayton Recompression cycle, again operating at a pressure of about 20 MPa. For both the direct and indirect cycles, sensitivity calculations were performed for reactor outlet temperature

  14. Current status and directions for fast reactor reprocessing

    SciTech Connect

    Burch, W.D.

    1983-01-01

    The development of fast breeder reactors (FBRs) for commercial electric power production has been under way in several countries for more than 20 years. In the United States as elsewhere, early work was centered on small reactors to prove the feasibility of concepts and later was followed by larger reactors to test engineering features and to develop fuel technology. In the early 1970s, with the perceived crisis in electrical generation expected late in this century, major efforts were mounted to plan and carry out comprehensive development programs to ensure the capability to develop and begin using this new form of nuclear power by the end of this century. This comprehensive effort included the first serious efforts directed toward the supporting fuel cycle activities. However, because of the effects of the oil price rise and resulting conservation, a slowdown of industrial growth, and cut-backs in energy needs, there has been a decline in program activities. Unlike the fuel cycle for light-water reactors (LWRs), where supply and the back-end recycle and/or waste disposal activities can largely be uncoupled, recovery and recycle of fissile materials in spent fuel must be accomplished in one or two years in a practical breeder system. 3 references.

  15. Safeguards operations in the integral fast reactor fuel cycle

    SciTech Connect

    Goff, K.M.; Benedict, R.W.; Brumbach, S.B.; Dickerman, C.E.; Tompot, R.W.

    1994-08-01

    Argonne National Laboratory is currently demonstrating the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The safeguards aspects of the fuel cycle demonstration must be approved by the United States Department of Energy, but a further goal of the program is to develop a safeguards system that could gain acceptance from the Nuclear Regulatory Commission and International Atomic Energy Agency. This fuel cycle is described with emphasis on aspects that differ from aqueous reprocessing and on its improved safeguardability due to decreased attractiveness and diversion potential of all process streams, including the fuel product.

  16. Multiple reheat helium Brayton cycles for sodium fast reactors

    SciTech Connect

    Haihua Zhao; Per F. Peterson

    2008-07-01

    Sodium fast reactors (SFR) traditionally adopt the steam Rankine cycle for power conversion. The resulting potential for water-sodium reaction remains a continuing concern which at least partly delays the SFR technology commercialization and is a contributor to higher capital cost. Supercritical CO2 provides an alternative, but is also capable of sustaining energetic chemical reactions with sodium. Recent development on advanced inert-gas Brayton cycles could potentially solve this compatibility issue, increase thermal efficiency, and bring down the capital cost close to light water reactors. In this paper, helium Brayton cycles with multiple reheat and intercooling states are presented for SFRs with reactor outlet temperatures in the range of 510°C to 650°C. The resulting thermal efficiencies range from 39% and 47%, which is comparable with supercritical recompression CO2 cycles (SCO2 cycle). A systematic comparison between multiple reheat helium Brayton cycle and the SCO2 cycle is given, considering compatibility issues, plant site cooling temperature effect on plant efficiency, full plant cost optimization, and other important factors. The study indicates that the multiple reheat helium cycle is the preferred choice over SCO2 cycle for sodium fast reactors.

  17. Direct conversion nuclear reactor space power systems

    SciTech Connect

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

    1982-08-01

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

  18. Code System for Reactor Physics and Fuel Cycle Simulation.

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterativemore » processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.« less

  19. 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.

  20. Direct Energy Conversion for Fast Reactors

    SciTech Connect

    Brown, N.; Cooper, J.; Vogt, D.; Chapline, G.; Turchi, P.; Barbee Jr., T.; Farmer, J.

    2000-07-01

    Thermoelectric generators (TEG) are a well-established technology for compact low power output long-life applications. Solid state TEGs are the technology of choice for many space missions and have also been used in remote earth-based applications. Since TEGs have no moving parts and can be hermetically sealed, there is the potential for nuclear reactor power systems using TEGs to be safe, reliable and resistant to proliferation. Such power units would be constructed in a manner that would provide decades of maintenance-free operation, thereby minimizing the possibility of compromising the system during routine maintenance operations. It should be possible to construct an efficient direct energy conversion cascade from an appropriate combination of solid-state thermoelectric generators, with each stage in the cascade optimized for a particular range of temperature. Performance of cascaded thermoelectric devices could be further enhanced by exploitation of compositionally graded p-n couples, as well as radial elements to maximize utilization of the heat flux. The Jet Propulsion Laboratory in Pasadena has recently reported segmented unicouples that operate between 300 and 975 K and have conversion efficiencies of 15 percent [Caillat, 2000]. TEGs are used in nuclear-fueled power sources for space exploration, in power sources for the military, and in electrical generators on diesel engines. Second, there is a wide variety of TE materials applicable to a broad range of temperatures. New materials may lead to new TEG designs with improved thermoelectric properties (i.e. ZT approaching 3) and significantly higher efficiencies than in designs using currently available materials. Computational materials science (CMS) has made sufficient progress and there is promise for using these techniques to reduce the time and cost requirements to develop such new TE material combinations. Recent advances in CMS, coupled with increased computational power afforded by the Accelerated

  1. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  2. 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.

  3. 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.

  4. Small particle bed reactors: Sensitivity to Brayton cycle parameters

    NASA Astrophysics Data System (ADS)

    Coiner, John R.; Short, Barry J.

    Relatively simple particle bed reactor (PBR) algorithms were developed for optimizing low power closed Brayton cycle (CBC) systems. These algorithms allow the system designer to understand the relationship among key system parameters as well as the sensitivity of the PBR size and mass (a major system component) to variations in these parameters. Thus, system optimization can be achieved.

  5. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    SciTech Connect

    Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

    2002-11-01

    This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

  6. 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.

  7. 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.

  8. Synfuels from fusion: producing hydrogen with the tandem mirror reactor and thermochemical cycles

    SciTech Connect

    Ribe, F.L.; Werner, R.W.

    1981-01-21

    This report examines, for technical merit, the combination of a fusion reactor driver and a thermochemical plant as a means for producing synthetic fuel in the basic form of hydrogen. We studied: (1) one reactor type - the Tandem Mirror Reactor - wishing to use to advantage its simple central cell geometry and its direct electrical output; (2) two reactor blanket module types - a liquid metal cauldron design and a flowing Li/sub 2/O solid microsphere pellet design so as to compare the technology, the thermal-hydraulics, neutronics and tritium control in a high-temperature operating mode (approx. 1200 K); (3) three thermochemical cycles - processes in which water is used as a feedstock along with a high-temperature heat source to produce H/sub 2/ and O/sub 2/.

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

    NASA Technical Reports Server (NTRS)

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

    1983-01-01

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

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

    NASA Astrophysics Data System (ADS)

    Palmer, David J.

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

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

    SciTech Connect

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

    2006-04-01

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

  12. Detection of anomalous reactor activity using antineutrino count evolution over the course of a reactor cycle

    NASA Astrophysics Data System (ADS)

    Bulaevskaya, Vera; Bernstein, Adam

    2011-06-01

    This paper analyzes the sensitivity of antineutrino count rate measurements to changes in the fissile content of civil power reactors. Such measurements may be useful in IAEA reactor safeguards applications. We introduce a hypothesis testing procedure to identify statistically significant differences between the antineutrino count rate evolution of a standard "baseline" fuel cycle and that of an anomalous cycle, in which plutonium is removed and replaced with an equivalent fissile worth of uranium. The test would allow an inspector to detect anomalous reactor activity, or to positively confirm that the reactor is operating in a manner consistent with its declared fuel inventory and power level. We show that with a reasonable choice of detector parameters, the test can detect replacement of 82 kg of plutonium in 90 days with 95% probability, while controlling the false positive rate at 5%. We show that some improvement on this level of sensitivity may be obtained by various means, including use of the method in conjunction with existing reactor safeguards methods. We also identify a necessary and sufficient minimum daily antineutrino count rate and a maximum tolerable background rate to achieve the quoted sensitivity, and list examples of detectors in which such rates have been attained.

  13. Sensitivity Analysis of Reprocessing Cooling Times on Light Water Reactor and Sodium Fast Reactor Fuel Cycles

    SciTech Connect

    R. M. Ferrer; S. Bays; M. Pope

    2008-04-01

    The purpose of this study is to quantify the effects of variations of the Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) and fast reactor reprocessing cooling time on a Sodium Fast Reactor (SFR) assuming a single-tier fuel cycle scenario. The results from this study show the effects of different cooling times on the SFR’s transuranic (TRU) conversion ratio (CR) and transuranic fuel enrichment. Also, the decay heat, gamma heat and neutron emission of the SFR’s fresh fuel charge were evaluated. A 1000 MWth commercial-scale SFR design was selected as the baseline in this study. Both metal and oxide CR=0.50 SFR designs are investigated.

  14. Reactor applications of the Compact Fusion Advanced Rankine (CFAR) cycle for a D-T tokamak fusion reactor

    NASA Astrophysics Data System (ADS)

    Hoffman, H. A.; Logan, B. G.; Campbell, R. B.

    1988-03-01

    A preliminary design of a D-T fusion reactor blanket and MHD power conversion system is made based on the CFAR concept, and it was found that performance and costs for the reference cycle are very attractive. While much remains to be done, the potential advantage of liquid metal Rankine cycles for fusion applications are much clearer now. These include low pressures and mass flow rates, a nearly isothermal module shell which minimizes problems of thermal distortion and stresses, and an insensitivity to pressure losses in the blanket so that the two-phase MHD pressure drops in the boiling part of the blanket and the ordinary vapor pressure drops in the pebble-bed superheating zones are acceptable (the direct result of pumping a liquid rather than having to compress a gas). There are no moving parts in the high-temperature MHD power generators, no steam bottoming plant is required, only small vapor precoolers and condensers are needed because of the high heat rejection temperatures, and only a relatively small natural-draft heat exchanger is required to reject the heat to the atmosphere. The net result is a very compact fusion reactor and power conversion system which fit entirely inside an 18 meter radius reactor vault. Although a cost analysis has not yet been performed, preliminary cost estimates indicate low capital costs and a very attractive cost of electricity.

  15. Direct Energy Conversion for Fast Reactors

    NASA Astrophysics Data System (ADS)

    Brown, N. W.; Vogt, D.; Cooper, J.; Chapline, G.; Turchi, P.

    2000-07-01

    Thermoelectric generators (TEG) are a well-established technology for compact low power output long-life applications. Solid state TEGs are the technology of choice for many space missions and remote earth-based applications. Use of solid state TEGs in these applications requires engineering designs that minimize the weight and volume of the device. Thermal to electric conversion efficiency, while an important design consideration, is not the principal design factor. However, design of a TEG for a fast reactor nuclear power plant requires higher thermal efficiencies in order to achieve competitive power generation costs.

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

    SciTech Connect

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

    2009-07-12

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

  17. Feasibility study on ultralong-cycle operation and material performance for compact liquid metal-cooled fast reactors: a review work

    SciTech Connect

    Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung; Kim, T. K.; Hong, Ser Gi

    2015-11-01

    This paper reviews the feasibility of ultralong-cycle operation on a compact liquid metal-cooled fast reactor (LMR) firstly by assessing the operation of a long-life fast reactor core and secondly by evaluating material performance in respect to both long-cycle operation and compact-size fast reactor. Many kinds of reactor concepts have been proposed, and LMR and small modular reactor (SMR) are the issued leading technologies for generation four (Gen-IV) reactor system development. The breed-and-burn strategy was proposed as a core burning strategy to operate a long cycle, and it has been evaluated in this paper with two reactor concepts: constant axial shape of neutron flux, nuclide densities, and power shape during life of energy and ultralong cycle fast reactor. In addition, Super-Safe, Small, and Simple and small modular fast reactor, compact LMR concepts, have been simulated to evaluate their long-life operation strategies. For the other practical issues, the materials for fuel, coolant, and structure have been identified and some of them are selected to have their performance optimized specifically for compact LMR with a long-cycle operation. It is believed that this comprehensive review will propose a proper direction for future reactor development and will be followed by the next step research for a complete reactor model with the other reactor components.

  18. Synfuels from fusion: producing hydrogen with the Tandem Mirror Reactor and thermochemical cycles

    SciTech Connect

    Werner, R.W.; Ribe, F.L.

    1981-01-21

    This volume contains the following sections: (1) the Tandem Mirror fusion driver, (2) the Cauldron blanket module, (3) the flowing microsphere, (4) coupling the reactor to the process, (5) the thermochemical cycles, and (6) chemical reactors and process units. (MOW)

  19. The IAEA international conference on fast reactors and related fuel cycles: highlights and main outcomes

    SciTech Connect

    Monti, S.; Toti, A.

    2013-07-01

    The 'International Conference on Fast Reactors and Related Fuel Cycles', which is regularly held every four years, represents the main international event dealing with fast reactors technology and related fuel cycles options. Main topics of the conference were new fast reactor concepts, design and simulation capabilities, safety of fast reactors, fast reactor fuels and innovative fuel cycles, analysis of past experience, fast reactor knowledge management. Particular emphasis was put on safety aspects, considering the current need of developing and harmonizing safety standards for fast reactors at the international level, taking also into account the lessons learned from the accident occurred at the Fukushima- Daiichi nuclear power plant in March 2011. Main advances in the several key areas of technological development were presented through 208 oral presentations during 41 technical sessions which shows the importance taken by fast reactors in the future of nuclear energy.

  20. Direct-energy-conversion implications of Space Nuclear Reactors

    SciTech Connect

    Morris, J.F.

    1982-08-01

    The Air Force, NASA and DOE stress space-nuclear reactor (SNR) needs in 1981 IECEC papers. SNR proposals range from 10-to-100kW /SUB e/,s with thermoelectrics through the fractional-to-several MW /SUB e/ 's with thermionic conversion to rotating bed-reactor (RBR) and NERVA ultraversions. SNR direct conversion comprises thermionic and thermoelectric generation (TEG). Thermionic energy conversion (TEC) pervades the pre-1973 in-core and out-of-core-heat-pipe concepts. SPAR and SP-100 focus on thermoelectrics because of ostensible fuel-temperature limits. A Rasor Associates mini-heat-pipe reactor verifies again the high-power capability of this SNR type--as well as TEC advantages over TEG. Finally with about 2000K effluents, directly from RBR's, NERVA's or from MHD used with them, TEC could also produce very high power levels. This paper outlines SNR needs, discusses some proposed concepts and recommends future technology programs.

  1. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 3-Surry Unit 1 Cycle 2

    SciTech Connect

    Bowman, S.M.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using selected critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations in this report is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of two reactor critical configurations for Surry Unit 1 Cycle 2. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted a direct comparison of criticality calculations using the utility-calculated isotopics with those using the isotopics generated by the SCALE-4

  2. 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.

  3. Direct nn-Scattering Measurement With the Pulsed Reactor YAGUAR

    PubMed Central

    Mitchell, G. E.; Furman, W. I.; Lychagin, E. V.; Muzichka, A. Yu.; Nekhaev, G. V.; Strelkov, A. V.; Sharapov, E. I.; Shvetsov, V. N.; Chernuhin, Yu. I.; Levakov, B. G.; Litvin, V. I.; Lyzhin, A. E.; Magda, E. P.; Crawford, B. E.; Stephenson, S. L.; Howell, C. R.; Tornow, W

    2005-01-01

    Although crucial for resolving the issue of charge symmetry in the nuclear force, direct measurement of nn-scattering by colliding free neutrons has never been performed. At present the Russian pulsed reactor YAGUAR is the best neutron source for performing such a measurement. It has a through channel where the neutron moderator is installed. The neutrons are counted by a neutron detector located 12 m from the reactor. In preliminary experiments an instantaneous value of 1.1 × 1018/cm2s was obtained for the thermal neutron flux density. The experiment will be performed by the DIANNA Collaboration as International Science & Technology Center (ISTC) project No. 2286. PMID:27308126

  4. Economic Analyiss of "Symbiotic" Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

    SciTech Connect

    Williams, Kent Alan; Shropshire, David E.

    2009-01-01

    A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle

  5. Use of RELAP5-3D for Dynamic Analysis of a Closed-Loop Brayton Cycle Coupled To a Nuclear Reactor

    NASA Astrophysics Data System (ADS)

    McCann, Larry D.

    2007-01-01

    This paper describes results of a dynamic system model for a pair of closed Brayton-cycle (CBC) loops running in parallel that are connected to a nuclear gas reactor. The model assumes direct coupling between the reactor and the Brayton-cycle loops. The RELAP5-3D (version 2.4.1) computer program was used to perform the analysis. Few reactors have ever been coupled to closed Brayton-cycle systems. As such their behavior under dynamically varying loads, startup and shut down conditions, and requirements for safe and autonomous operation are largely unknown. The model described in this paper represents the reactor, turbine, compressor, recuperator, heat rejection system and alternator. The initial results of the model indicate stable operation of the reactor-driven Brayton-cycle system. However, for analysts with mostly pressurized water reactor experience, the Brayton cycle loops coupled to a gas-cooled reactor also indicate some counter-intuitive behavior for the complete coupled system. This model has provided crucial information in evaluating the reactor design and would have been further developed for use in developing procedures for safe start up, shut down, safe-standby, and other autonomous operating modes had the plant development cycle been completed.

  6. Use of RELAP5-3D for Dynamic Analysis of a Closed-Loop Brayton Cycle Coupled To a Nuclear Reactor

    SciTech Connect

    McCann, Larry D.

    2007-01-30

    This paper describes results of a dynamic system model for a pair of closed Brayton-cycle (CBC) loops running in parallel that are connected to a nuclear gas reactor. The model assumes direct coupling between the reactor and the Brayton-cycle loops. The RELAP5-3D (version 2.4.1) computer program was used to perform the analysis. Few reactors have ever been coupled to closed Brayton-cycle systems. As such their behavior under dynamically varying loads, startup and shut down conditions, and requirements for safe and autonomous operation are largely unknown. The model described in this paper represents the reactor, turbine, compressor, recuperator, heat rejection system and alternator. The initial results of the model indicate stable operation of the reactor-driven Brayton-cycle system. However, for analysts with mostly pressurized water reactor experience, the Brayton cycle loops coupled to a gas-cooled reactor also indicate some counter-intuitive behavior for the complete coupled system. This model has provided crucial information in evaluating the reactor design and would have been further developed for use in developing procedures for safe start up, shut down, safe-standby, and other autonomous operating modes had the plant development cycle been completed.

  7. Thorium: Uranium fuel cycle in safe reactors, the time is now

    SciTech Connect

    Gat, Uri

    1995-12-31

    The thorium-uranium fuel cycle has several advantages that make it attractive. Some of these beneficial properties are of particular interest now as they help alleviate current concerns. The Th-U cycle has neutronic advantages when utilized in thermal or epithermal reactors. Some of these reactors enjoy extraordinary safety qualities. The combination of these traits suggest that now is an appropriate time to deploy and begin exploiting the Th-U fuel cycle.

  8. Closing nuclear fuel cycle with fast reactors: problems and prospects

    SciTech Connect

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V.

    2013-07-01

    The closed nuclear fuel cycle (CNFC) with fast reactors (FR) is the most promising way of nuclear energetics development because it prevents spent nuclear fuel (SNF) accumulation and minimizes radwaste volume due to minor actinides (MA) transmutation. CNFC with FR requires the elaboration of safety, environmentally acceptable and economically effective methods of treatment of SNF with high burn-up and low cooling time. The up-to-date industrially implemented SNF reprocessing technologies based on hydrometallurgical methods are not suitable for the reprocessing of SNF with high burn-up and low cooling time. The alternative dry methods (such as electrorefining in molten salts or fluoride technologies) applicable for such SNF reprocessing have not found implementation at industrial scale. So the cost of SNF reprocessing by means of dry technologies can hardly be estimated. Another problem of dry technologies is the recovery of fissionable materials pure enough for dense fuel fabrication. A combination of technical solutions performed with hydrometallurgical and dry technologies (pyro-technology) is proposed and it appears to be a promising way for the elaboration of economically, ecologically and socially accepted technology of FR SNF management. This paper deals with discussion of main principle of dry and aqueous operations combination that probably would provide safety and economic efficiency of the FR SNF reprocessing. (authors)

  9. A Parametric Study of the DUPIC Fuel Cycle to Reflect Pressurized Water Reactor Fuel Management Strategy

    SciTech Connect

    Rozon, Daniel; Shen Wei

    2001-05-15

    For both pressurized water reactor (PWR) and Canada deuterium uranium (CANDU) tandem analysis, the Direct Use of spent PWR fuel In CANDU reactor (DUPIC) fuel cycle in a CANDU 6 reactor is studied using the DRAGON/DONJON chain of codes with the ENDF/B-V and ENDF/B-VI libraries. The reference feed material is a 17 x 17 French standard 900-MW(electric) PWR fuel. The PWR spent-fuel composition is obtained from two-dimensional DRAGON assembly transport and depletion calculations. After a number of years of cooling, this defines the initial fuel nuclide field in the CANDU unit cell calculations in DRAGON, where it is further depleted with the same neutron group structure. The resulting macroscopic cross sections are condensed and tabulated to be used in a full-core model of a CANDU 6 reactor to find an optimized channel fueling rate distribution on a time-average basis. Assuming equilibrium refueling conditions and a particular refueling sequence, instantaneous full-core diffusion calculations are finally performed with the DONJON code, from which both the channel power peaking factors and local parameter effects are estimated. A generic study of the DUPIC fuel cycle is carried out using the linear reactivity model for initial enrichments ranging from 3.2 to 4.5 wt% in a PWR. Because of the uneven power histories of the spent PWR assemblies, the spent PWR fuel composition is expected to differ from one assembly to the next. Uneven mixing of the powder during DUPIC fuel fabrication may lead to uncertainties in the composition of the fuel bundle and larger peaking factors in CANDU. A mixing method for reducing composition uncertainties is discussed.

  10. Analysis of supercritical CO{sub 2} cycle control strategies and dynamic response for Generation IV Reactors.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2011-04-12

    The analysis of specific control strategies and dynamic behavior of the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle has been extended to the two reactor types selected for continued development under the Generation IV Nuclear Energy Systems Initiative; namely, the Very High Temperature Reactor (VHTR) and the Sodium-Cooled Fast Reactor (SFR). Direct application of the standard S-CO{sub 2} recompression cycle to the VHTR was found to be challenging because of the mismatch in the temperature drop of the He gaseous reactor coolant through the He-to-CO{sub 2} reactor heat exchanger (RHX) versus the temperature rise of the CO{sub 2} through the RHX. The reference VHTR features a large temperature drop of 450 C between the assumed core outlet and inlet temperatures of 850 and 400 C, respectively. This large temperature difference is an essential feature of the VHTR enabling a lower He flow rate reducing the required core velocities and pressure drop. In contrast, the standard recompression S-CO{sub 2} cycle wants to operate with a temperature rise through the RHX of about 150 C reflecting the temperature drop as the CO{sub 2} expands from 20 MPa to 7.4 MPa in the turbine and the fact that the cycle is highly recuperated such that the CO{sub 2} entering the RHX is effectively preheated. Because of this mismatch, direct application of the standard recompression cycle results in a relatively poor cycle efficiency of 44.9%. However, two approaches have been identified by which the S-CO{sub 2} cycle can be successfully adapted to the VHTR and the benefits of the S-CO{sub 2} cycle, especially a significant gain in cycle efficiency, can be realized. The first approach involves the use of three separate cascaded S-CO{sub 2} cycles. Each S-CO{sub 2} cycle is coupled to the VHTR through its own He-to-CO{sub 2} RHX in which the He temperature is reduced by 150 C. The three respective cycles have efficiencies of 54, 50, and 44%, respectively, resulting in a net cycle

  11. Analysis of closed cycle megawatt class space power systems with nuclear reactor heat sources

    NASA Technical Reports Server (NTRS)

    Juhasz, A. J.; Jones, B. I.

    1987-01-01

    The analysis and integration studies of multimegawatt nuclear power conversion systems for potential SDI applications is presented. A study is summarized which considered 3 separate types of power conversion systems for steady state power generation with a duty requirement of 1 yr at full power. The systems considered are based on the following conversion cycles: direct and indirect Brayton gas turbine, direct and indirect liquid metal Rankine, and in core thermionic. A complete mass analysis was performed for each system at power levels ranging from 1 to 25 MWe for both heat pipe and liquid droplet radiator options. In the modeling of common subsystems, reactor and shield calculations were based on multiparameter correlation and an in-house analysis for the heat rejection and other subsystems.

  12. Nuclear fuel cycle analysis of the SABR fusion-fission hybrid transmutation reactor

    NASA Astrophysics Data System (ADS)

    Sommer, Chris; Stacey, Weston; Petrovic, Bojan

    2009-11-01

    Various fuel cycles have been designed and analyzed for the Subcritical Advanced Burner Reactor (SABR). SABR is a sodium cooled fast reactor fueled with transuranics (TRU) from spent fuel of light water reactors and driven by a tokamak fusion neutron source based on ITER physics and technology. SABR employs a four batch fuel cycle using an out-to-in shuffling pattern, with the fuel being reprocessed at the end of each cycle. The reprocessing method assumes recovery rates of 99.9% of the actinides and 0.1% of the fission products remain in the recycled fuel. The reprocessing fuel cycles were analyzed to find an optimal cycle length in terms of burn up, power distribution, and materials limitations. Fuel cycles are analyzed using CEA's ERANOS2.0 code, with fuel residence times limited by radiation damage at 100, 150 and 200 dpa.

  13. Measurement control design and performance assessment in the Integral Fast Reactor fuel cycle

    SciTech Connect

    Orechwa, Y.; Bucher, R.G.

    1994-08-01

    The Integral Fast Reactor (IFR)--consisting of a metal fueled and liquid metal cooled reactor together with an attendant fuel cycle facility (FCF)--is currently undergoing a phased demonstration of the closed fuel cycle at Argonne National Laboratory. The recycle technology is pyrometalurgical based with incomplete fission product separation and all transuranics following plutonium for recycle. The equipment operates in batch mode at 500 to 1,300 C. The materials are highly radioactive and pyrophoric, thus the FCF requires remote operation. Central to the material control and accounting system for the FCF are the balances for mass measurements. The remote operation of the balances limits direct adjustment. The radiation environment requires that removal and replacement of the balances be minimized. The uniqueness of the facility precludes historical data for design and performance assessment. To assure efficient operation of the facility, the design of the measurement control system has called for procedures which assess the performance of the balances in great detail and will support capabilities for the correction of systematic changes in the performance of the balances through software.

  14. Treatment of sewage sludge in a thermophilic membrane reactor (TMR) with alternate aeration cycles.

    PubMed

    Collivignarelli, Maria Cristina; Castagnola, Federico; Sordi, Marco; Bertanza, Giorgio

    2015-10-01

    The management of sewage sludge is becoming a more and more important issue, both at national and international level, in particular due to the uncertain recovery/disposal future options. Therefore, it is clear that the development of new technologies that can mitigate the problem at the source by reducing sludge production is necessary, such as the European Directive 2008/98/EC prescribes. This work shows the results obtained with a thermophilic membrane reactor, for processing a biological sludge derived from a wastewater treatment plant (WWTP) that treats urban and industrial wastewater. Sewage sludge was treated in a thermophilic membrane reactor (TMR), at pilot-scale (1 m(3) volume), with alternate aeration cycles. The experimentation was divided into two phases: a "startup phase" during which, starting with a psychrophilic/mesophilic biomass, thermophilic conditions were progressively reached, while feeding a highly biodegradable substrate; the obtained thermophilic biomass was then used, in the "regime phase", to digest biological sludge which was fed to the plant. Good removal yields were observed: 64% and 57% for volatile solids (VS) and total COD (CODtot), respectively, with an average hydraulic retention time (HRT) equal to 20 d, an organic loading rate (OLR) of about 1.4-1.8 kg COD m(-3) d(-1) and aeration/non aeration cycles alternated every 4 h.

  15. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    NASA Astrophysics Data System (ADS)

    Fic, Adam; Składzień, Jan; Gabriel, Michał

    2015-03-01

    Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

  16. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    PubMed

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described.

  17. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    PubMed

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. PMID:21399407

  18. A NOVEL MEMBRANE REACTOR FOR DIRECT HYDROGEN PRODUCTION FROM COAL

    SciTech Connect

    Shain Doong; Estela Ong; Mike Atroshenko; Mike Roberts; Francis Lau

    2004-04-26

    Gas Technology Institute is developing a novel concept of membrane gasifier for high efficiency, clean and low cost production of hydrogen from coal. The concept incorporates a hydrogen-selective membrane within a gasification reactor for direct extraction of hydrogen from coal synthesis gases. The objective of this project is to determine the technical and economic feasibility of this concept by screening, testing and identifying potential candidate membranes under high temperature, high pressure, and harsh environments of the coal gasification conditions. The best performing membranes will be selected for preliminary reactor design and cost estimates. To evaluate the performances of the candidate membranes under the gasification conditions, a high temperature/high pressure hydrogen permeation unit will be constructed in this project. During this reporting period, the mechanical construction of the permeation unit was completed. Commissioning and shake down tests are being conducted. The unit is capable of operation at temperatures up to 1100 C and pressures to 60 atm for evaluation of ceramic membranes such as mixed ionic conducting membrane. The membranes to be tested will be in disc form with a diameter of about 3 cm. Operation at these high temperatures and high hydrogen partial pressures will demonstrate commercially relevant hydrogen flux, 10{approx}50 cc/min/cm{sup 2}, from the membranes made of the perovskite type of ceramic material. Preliminary modeling was also performed for a tubular membrane reactor within a gasifier to estimate the required membrane area for a given gasification condition. The modeling results will be used to support the conceptual design of the membrane reactor.

  19. Metal hydrides reactors with improved dynamic characteristics for a fast cycling hydrogen compressor

    NASA Astrophysics Data System (ADS)

    Popeneciu, G.; Coldea, I.; Lupu, D.; Misan, I.; Ardelean, O.

    2009-08-01

    This paper presents an investigation of coupled heat and mass transfer process in metal hydrides hydrogen storage reactors. Hydrogen storage and compression performance of our designed and developed reactors are studied by varying the operating parameters and analyzing the effects of metal hydride bed parameters. The metal alloy selected to characterize the cycling behaviour of reactors is LaNi5, material synthesized and characterized by us in the range 20-80°C. Four types of metal hydride reactors were tested with the aim to provide a fast hydrogen absorption-desorption cycle, able to be thermally cycled at rapid rates. Some new technical solutions have been studied to make a step forward in reducing the duration of the reactors cycle, which combines the effective increase of the thermal conductivity and good permeability to hydrogen gas. Dynamic characteristic of developed fast metal hydride reactors is improved using our novel mixture metal hydride-CA conductive additive due to the increased effective thermal conductivity of the alloy bed. The advanced hydride bed design with high heat transfer capabilities can be thermally cycled at a rapid rate, under 120 seconds, in order to process high hydrogen flow rates.

  20. Modular high temperature gas-cooled reactor plant design duty cycle. Revision 3

    SciTech Connect

    Chan, T.

    1989-12-31

    This document defines the Plant Design Duty Cycle (PCDC) for the Modular High Temperature Gas-cooled Reactor (MHTGR). The duty cycle is a set of events and their design number of occurrences over the life of the plant for which the MHTGR plant shall be designed to ensure that the plant meets all the top-level requirements. The duty cycle is representative of the types of events to be expected in multiple reactor module-turbine plant configurations of the MHTGR. A synopsis of each PDDC event is presented to provide an overview of the plant response and consequence. 8 refs., 1 fig., 4 tabs.

  1. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

    NASA Astrophysics Data System (ADS)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

    1992-10-01

    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  2. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

    1992-01-01

    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  3. Design and Cold Mode Experiment of Dual Bubbling Fluidized Bed Reactors for Multiple CCR Cycles

    NASA Astrophysics Data System (ADS)

    Fang, F.; Li, Z. S.; Cai, N. S.

    The dual fluidized bed reactors are the key technology to fulfill the multiple CCR (calcination/carbonation reactions) cycles for CO2 capture from the flue gases. Firstly, the dual bubbling fluidized bed reactors were selected in this work based on analyzing different types of dual fluidized bed reactors. Secondly, the design method of dual fluidized bed reactors for CO2 capture with CCR concept was proposed. Thirdly, with the designed results, a cold mode of the dual bubbling fluidized bed reactors was built. The long-term stable operation and the continuous solid circulation between two reactors could be achieved successfully. The experimental results indicated that the solid circulation rate was increased with an increase of bed height, diameter of solid injection nozzle, and diameter of holes on the solid injection nozzle.

  4. Promising Fuel Cycle Options for R&D – Results, Insights, and Future Directions

    SciTech Connect

    Wigeland, Roald Arnold

    2015-05-01

    The Fuel Cycle Options (FCO) campaign in the U.S. DOE Fuel Cycle Research & Development Program conducted a detailed evaluation and screening of nuclear fuel cycles. The process for this study was described at the 2014 ICAPP meeting. This paper reports on detailed insights and questions from the results of the study. The comprehensive study identified continuous recycle in fast reactors as the most promising option, using either U/Pu or U/TRU recycle, and potentially in combination with thermal reactors, as reported at the ICAPP 2014 meeting. This paper describes the examination of the results in detail that indicated that there was essentially no difference in benefit between U/Pu and U/TRU recycle, prompting questions about the desirability of pursuing the more complex U/TRU approach given that the estimated greater challenges for development and deployment. The results will be reported from the current effort that further explores what, if any, benefits of TRU recycle (minor actinides in addition to plutonium recycle) may be in order to inform decisions on future R&D directions. The study also identified continuous recycle using thorium-based fuel cycles as potentially promising, in either fast or thermal systems, but with lesser benefit. Detailed examination of these results indicated that the lesser benefit was confined to only a few of the evaluation metrics, identifying the conditions under which thorium-based fuel cycles would be promising to pursue. For the most promising fuel cycles, the FCO is also conducting analyses on the potential transition to such fuel cycles to identify the issues, challenges, and the timing for critical decisions that would need to be made to avoid unnecessary delay in deployment, including investigation of issues such as the effects of a temporary lack of plutonium fuel resources or supporting infrastructure. These studies are placed in the context of an overall analysis approach designed to provide comprehensive information to

  5. Development potential for thermal reactors and their fuel cycles

    SciTech Connect

    Dodds, H.L.; Gat, U.

    1997-08-01

    The advantages of molten salt reactors (MSRs) for power production are very briefly described in this paper. The MSRs considered are those with on-line fuel processing, external cooling, and fluoride salt separation. Characteristics noted include lack of meltdown potential, small radioactive source terms, and complete burnup of fissile material. The burnup capability of MSRs would allow them to be used to dispose of plutonium while producing energy. 8 refs.

  6. Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

    2015-12-01

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  7. Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    SciTech Connect

    Shmelev, A. N. Kulikov, G. G. Kurnaev, V. A. Salahutdinov, G. H. Kulikov, E. G. Apse, V. A.

    2015-12-15

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the {sup 231}Pa–{sup 232}U–{sup 233}U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of {sup 232}U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  8. Synfuels from fusion: using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

    SciTech Connect

    Werner, R.W.

    1982-11-01

    This study is concerned with the following area: (1) the tandem mirror reactor and its physics; (2) energy balance; (3) the lithium oxide canister blanket system; (4) high-temperature blanket; (5) energy transport system-reactor to process; (6) thermochemical hydrogen processes; (7) interfacing the GA cycle; (8) matching power and temperature demands; (9) preliminary cost estimates; (10) synfuels beyond hydrogen; and (11) thermodynamics of the H/sub 2/SO/sub 4/-H/sub 2/O system. (MOW)

  9. Configuration of a molten chloride fast reactor on a thorium fuel cycle to current nuclear fuel cycle concerns

    SciTech Connect

    Ottewitte, E.H.

    1982-01-01

    Current concerns about the nuclear fuel cycle seem to center on waste management, non-proliferation, and optimum fuel utilization (including use of thorium). This thesis attempts to design a fast molten-salt reactor on the thorium fuel cycle to address these concerns and then analyzes its potential performance. The result features (1) A simplified easy-to-replace skewed-tube geometry for the core. (2) A very hard neutron spectrum which allows the useful consumption of all the actinides (no actinide waste). (3) Reduced proliferation risks on the equilibrium cycle compared to conventional fuel cycles because of the absence of carcinogenic, chemically-separable plutonium and the presence of /sup 232/U which gives a tell-tale signal and is hazardous to work with. (4) A breeding gain in the neighborhood of 0.3.

  10. Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs)

    NASA Astrophysics Data System (ADS)

    Lizon-A-Lugrin, Laure

    The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a

  11. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    SciTech Connect

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  12. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

    NASA Technical Reports Server (NTRS)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

    1996-01-01

    Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

  13. Proliferation resistance for fast reactors and related fuel cycles: issues and impacts

    SciTech Connect

    Pilat, Joseph F

    2010-01-01

    The prospects for a dramatic growth in nuclear power may depend to a significant degree on the effectiveness of, and the resources devoted to, plans to develop and implement technologies and approaches that strengthen proliferation resistance and nuclear materials accountability. The challenges for fast reactors and related fuel cycles are especially critical. They are being explored in the Generation IV Tnternational Forum (GIF) and the Tnternational Atomic Energy Agency's (IAEA's) International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiative, as well as by many states that are looking to these systems for the efficient lise of uranium resources and long-term energy security. How do any proliferation risks they may pose compare to other reactors, both existing and under development, and their fuel cycles? Can they be designed with intrinsic (technological) features to make these systems more proliferation resistant? What roles can extrinsic (institutional) features play in proliferation resistance? What are the anticipated safeguards requirements, and will new technologies and approaches need to be developed? How can safeguards be facilitated by the design process? These and other questions require a rethinking of proliferation resistance and the prospects for new technologies and other intrinsic and extrinsic features being developed that are responsive to specific issues for fast reactors and related fuel cycles and to the broader threat environment in which these systems will have to operate. There are no technologies that can wholly eliminate the risk of proliferation by a determined state, but technology and design can playa role in reducing state threats and perhaps in eliminating non-state threats. There will be a significant role for extrinsic factors, especially the various measures - from safeguards and physical protection to export controls - embodied in the international nuclear nonproliferation regime. This paper will offer

  14. Development of a Direct Evaporator for the Organic Rankine Cycle

    SciTech Connect

    Donna Post Guillen; Helge Klockow; Matthew Lehar; Sebastian Freund; Jennifer Jackson

    2011-02-01

    This paper describes research and development currently underway to place the evaporator of an Organic Rankine Cycle (ORC) system directly in the path of a hot exhaust stream produced by a gas turbine engine. The main goal of this research effort is to improve cycle efficiency and cost by eliminating the usual secondary heat transfer loop. The project’s technical objective is to eliminate the pumps, heat exchangers and all other added cost and complexity of the secondary loop by developing an evaporator that resides in the waste heat stream, yet virtually eliminates the risk of a working fluid leakage into the gaseous exhaust stream. The research team comprised of Idaho National Laboratory and General Electric Company engineers leverages previous research in advanced ORC technology to develop a new direct evaporator design that will reduce the ORC system cost by up to 15%, enabling the rapid adoption of ORCs for waste heat recovery.

  15. Hybrid sulfur cycle operation for high-temperature gas-cooled reactors

    DOEpatents

    Gorensek, Maximilian B

    2015-02-17

    A hybrid sulfur (HyS) cycle process for the production of hydrogen is provided. The process uses a proton exchange membrane (PEM) SO.sub.2-depolarized electrolyzer (SDE) for the low-temperature, electrochemical reaction step and a bayonet reactor for the high-temperature decomposition step The process can be operated at lower temperature and pressure ranges while still providing an overall energy efficient cycle process.

  16. Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

    SciTech Connect

    Middleton, Bobby; Pasch, James Jay; Kruizenga, Alan Michael; Walker, Matthew

    2016-01-01

    This report outlines the thermodynamics of a supercritical carbon dioxide (sCO2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related to both Helium and to sCO2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation

  17. A Novel Membrane Reactor for Direct Hydrogen Production From Coal

    SciTech Connect

    Shain Doong; Estela Ong; Mike Atrosphenko; Francis Lau; Mike Roberts

    2006-01-20

    Gas Technology Institute has developed a novel concept of a membrane reactor closely coupled with a coal gasifier for direct extraction of hydrogen from coal-derived syngas. The objective of this project is to determine the technical and economic feasibility of this concept by screening, testing and identifying potential candidate membranes under the coal gasification conditions. The best performing membranes were selected for preliminary reactor design and cost estimate. The overall economics of hydrogen production from this new process was assessed and compared with conventional hydrogen production technologies from coal. Several proton-conducting perovskite membranes based on the formulations of BCN (BaCe{sub 0.8}Nd{sub 0.2}O{sub 3-x}), BCY (BaCe{sub 0.8}Y{sub 0.2}O{sub 3-x}), SCE (Eu-doped SrCeO{sub 3}) and SCTm (SrCe{sub 0.95}Tm{sub 0.05}O{sub 3}) were successfully tested in a new permeation unit at temperatures between 800 and 1040 C and pressures from 1 to 12 bars. The experimental data confirm that the hydrogen flux increases with increasing hydrogen partial pressure at the feed side. The highest hydrogen flux measured was 1.0 cc/min/cm{sup 2} (STP) for the SCTm membrane at 3 bars and 1040 C. The chemical stability of the perovskite membranes with respect to CO{sub 2} and H{sub 2}S can be improved by doping with Zr, as demonstrated from the TGA (Thermal Gravimetric Analysis) tests in this project. A conceptual design, using the measured hydrogen flux data and a modeling approach, for a 1000 tons-per-day (TPD) coal gasifier shows that a membrane module can be configured within a fluidized bed gasifier without a substantial increase of the gasifier dimensions. Flowsheet simulations show that the coal to hydrogen process employing the proposed membrane reactor concept can increase the hydrogen production efficiency by more than 50% compared to the conventional process. Preliminary economic analysis also shows a 30% cost reduction for the proposed membrane

  18. 75 FR 36648 - Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technologies Subcommittee

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-28

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technologies Subcommittee AGENCY: Office of Nuclear Energy, DOE. ACTION: Notice of open meeting correction. On June 21, 2010,...

  19. Accelerator Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

    DOE PAGES

    Brown, Nicholas R.; Heidet, Florent; Haj Tahar, Malek

    2016-01-01

    This article is a review of several accelerator–reactor interface issues and nuclear fuel cycle applications of acceleratordriven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systemsmore » on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.« less

  20. Accelerator Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

    SciTech Connect

    Brown, Nicholas R.; Heidet, Florent; Haj Tahar, Malek

    2016-01-01

    This article is a review of several accelerator–reactor interface issues and nuclear fuel cycle applications of acceleratordriven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systems on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.

  1. Accelerator-Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

    NASA Astrophysics Data System (ADS)

    Heidet, Florent; Brown, Nicholas R.; Haj Tahar, Malek

    This article is a review of several accelerator-reactor interface issues and nuclear fuel cycle applications of accelerator-driven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systems on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.

  2. 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

  3. A dynamic fuel cycle analysis for a heterogeneous thorium-DUPIC recycle in CANDU reactors

    SciTech Connect

    Jeong, C. J.; Park, C. J.; Choi, H.

    2006-07-01

    A heterogeneous thorium fuel recycle scenario in a Canada deuterium uranium (CANDU) reactor has been analyzed by the dynamic analysis method. The thorium recycling is performed through a dry process which has a strong proliferation resistance. In the fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides, and fission products of a multiple thorium recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. The analysis results have shown that the heterogeneous thorium fuel cycle can be constructed through the dry process technology. It is also shown that the heterogeneous thorium fuel cycle can reduce the spent fuel inventory and save on the natural uranium resources when compared with the once-through cycle. (authors)

  4. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    SciTech Connect

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team indicates

  5. Wastes from selected activities in two light-water reactor fuel cycles

    SciTech Connect

    Palmer, C.R.; Hill, O.F.

    1980-07-01

    This report presents projected volumes and radioactivities of wastes from the production of electrical energy using light-water reactors (LWR). The projections are based upon data developed for a recent environmental impact statement in which the transuranic wastes (i.e., those wastes containing certain long-lived alpha emitters at concentrations of at least 370 becquerels, or 10 nCi, per gram of waste) from fuel cycle activities were characterized. In addition, since the WG.7 assumed that all fuel cycle wastes except mill tailings are placed in a mined geologic repository, the nontransuranic wastes from several activities are included in the projections reported. The LWR fuel cycles considered are the LWR, once-through fuel cycle (Strategy 1), in which spent fuel is packaged in metal canisters and then isolated in geologic formations; and the LWR U/Pu recycle fuel cycle (Strategy 2), wherein spent fuel is reprocessed for recovery and recycle of uranium and plutonium in LWRs. The wastes projected for the two LWR fuel cycles are summarized. The reactor operations and decommissioning were found to dominate the rate of waste generation in each cycle. These activities account for at least 85% of the fuel cycle waste volume (not including head-end wastes) when normalized to per unit electrical energy generated. At 10 years out of reactor, however, spent fuel elements in Strategy 1 represent 98% of the fuel cycle activity but only 4% of the volume. Similarly, the packaged high-level waste, fuel hulls and hardware in Strategy 2 concentrate greater than 95% of the activity in 2% of the waste volume.

  6. Process Model of A Fusion Fuel Recovery System for a Direct Drive IFE Power Reactor

    NASA Astrophysics Data System (ADS)

    Natta, Saswathi; Aristova, Maria; Gentile, Charles

    2008-11-01

    A task has been initiated to develop a detailed representative model for the fuel recovery system (FRS) in the prospective direct drive inertial fusion energy (IFE) reactor. As part of the conceptual design phase of the project, a chemical process model is developed in order to observe the interaction of system components. This process model is developed using FEMLAB Multiphysics software with the corresponding chemical engineering module (CEM). Initially, the reactants, system structure, and processes are defined using known chemical species of the target chamber exhaust. Each step within the Fuel recovery system is modeled compartmentally and then merged to form the closed loop fuel recovery system. The output, which includes physical properties and chemical content of the products, is analyzed after each step of the system to determine the most efficient and productive system parameters. This will serve to attenuate possible bottlenecks in the system. This modeling evaluation is instrumental in optimizing and closing the fusion fuel cycle in a direct drive IFE power reactor. The results of the modeling are presented in this paper.

  7. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2004-02-01

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

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

    SciTech Connect

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

    2004-02-04

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

  10. Design of a Solar Reactor to Split CO2 Via Isothermal Redox Cycling of Ceria

    SciTech Connect

    Bader, R; Chandran, RB; Venstrom, LJ; Sedler, SJ; Krenzke, PT; De Smith, RM; Banerjee, A; Chase, TR; Davidson, JH; Lipinski, W

    2014-12-23

    The design procedure for a 3 kWth prototype solar thermochemical reactor to implement isothermal redox cycling of ceria for CO2 splitting is presented. The reactor uses beds of mm-sized porous ceria particles contained in the annulus of concentric alumina tube assemblies that line the cylindrical wall of a solar cavity receiver. The porous particle beds provide high surface area for the heterogeneous reactions, rapid heat and mass transfer, and low pressure drop. Redox cycling is accomplished by alternating flows of inert sweep gas and CO2 through the bed. The gas flow rates and cycle step durations are selected by scaling the results from small-scale experiments. Thermal and thermo-mechanical models of the reactor and reactive element tubes are developed to predict the steady-state temperature and stress distributions for nominal operating conditions. The simulation results indicate that the target temperature of 1773K will be reached in the prototype reactor and that the Mohr-Coulomb static factor of safety is above two everywhere in the tubes, indicating that thermo-mechanical stresses in the tubes remain acceptably low.

  11. Direct magnetocaloric characterization and simulation of thermomagnetic cycles.

    PubMed

    Porcari, G; Buzzi, M; Cugini, F; Pellicelli, R; Pernechele, C; Caron, L; Brück, E; Solzi, M

    2013-07-01

    An experimental setup for the direct measurement of the magnetocaloric effect capable of simulating high frequency magnetothermal cycles on laboratory-scale samples is described. The study of the magnetocaloric properties of working materials under operative conditions is fundamental for the development of innovative devices. Frequency and time dependent characterization can provide essential information on intrinsic features such as magnetic field induced fatigue in materials undergoing first order magnetic phase transitions. A full characterization of the adiabatic temperature change performed for a sample of Gadolinium across its Curie transition shows the good agreement between our results and literature data and in-field differential scanning calorimetry. PMID:23902084

  12. Direct magnetocaloric characterization and simulation of thermomagnetic cycles

    NASA Astrophysics Data System (ADS)

    Porcari, G.; Buzzi, M.; Cugini, F.; Pellicelli, R.; Pernechele, C.; Caron, L.; Brück, E.; Solzi, M.

    2013-07-01

    An experimental setup for the direct measurement of the magnetocaloric effect capable of simulating high frequency magnetothermal cycles on laboratory-scale samples is described. The study of the magnetocaloric properties of working materials under operative conditions is fundamental for the development of innovative devices. Frequency and time dependent characterization can provide essential information on intrinsic features such as magnetic field induced fatigue in materials undergoing first order magnetic phase transitions. A full characterization of the adiabatic temperature change performed for a sample of Gadolinium across its Curie transition shows the good agreement between our results and literature data and in-field differential scanning calorimetry.

  13. Comparative study of pulsed and steady-state tokamak reactor burn cycles

    SciTech Connect

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1984-05-01

    Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles.

  14. Science and cycling: current knowledge and future directions for research.

    PubMed

    Atkinson, Greg; Davison, Richard; Jeukendrup, Asker; Passfield, Louis

    2003-09-01

    In this holistic review of cycling science, the objectives are: (1) to identify the various human and environmental factors that influence cycling power output and velocity; (2) to discuss, with the aid of a schematic model, the often complex interrelationships between these factors; and (3) to suggest future directions for research to help clarify how cycling performance can be optimized, given different race disciplines, environments and riders. Most successful cyclists, irrespective of the race discipline, have a high maximal aerobic power output measured from an incremental test, and an ability to work at relatively high power outputs for long periods. The relationship between these characteristics and inherent physiological factors such as muscle capilliarization and muscle fibre type is complicated by inter-individual differences in selecting cadence for different race conditions. More research is needed on high-class professional riders, since they probably represent the pinnacle of natural selection for, and physiological adaptation to, endurance exercise. Recent advances in mathematical modelling and bicycle-mounted strain gauges, which can measure power directly in races, are starting to help unravel the interrelationships between the various resistive forces on the bicycle (e.g. air and rolling resistance, gravity). Interventions on rider position to optimize aerodynamics should also consider the impact on power output of the rider. All-terrain bicycle (ATB) racing is a neglected discipline in terms of the characterization of power outputs in race conditions and the modelling of the effects of the different design of bicycle frame and components on the magnitude of resistive forces. A direct application of mathematical models of cycling velocity has been in identifying optimal pacing strategies for different race conditions. Such data should, nevertheless, be considered alongside physiological optimization of power output in a race. An even distribution

  15. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

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

  16. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    NASA Astrophysics Data System (ADS)

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

    2015-12-01

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  17. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    SciTech Connect

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A. Ignatiev, V. V.; Subbotin, S. A. Tsibulskiy, V. F.

    2015-12-15

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  18. Fuel Cycle System Analysis Implications of Sodium-Cooled Metal-Fueled Fast Reactor Transuranic Conversion Ratio

    SciTech Connect

    Steven J. Piet; Edward A. Hoffman; Samuel E. Bays; Gretchen E. Matthern; Jacob J. Jacobson; Ryan Clement; David W. Gerts

    2013-03-01

    If advanced fuel cycles are to include a large number of fast reactors (FRs), what should be the transuranic (TRU) conversion ratio (CR)? The nuclear energy era started with the assumption that they should be breeder reactors (CR > 1), but the full range of possible CRs eventually received attention. For example, during the recent U.S. Global Nuclear Energy Partnership program, the proposal was burner reactors (CR < 1). Yet, more recently, Massachusetts Institute of Technology's "Future of the Nuclear Fuel Cycle" proposed CR [approximately] 1. Meanwhile, the French company EDF remains focused on breeders. At least one of the reasons for the differences of approach is different fuel cycle objectives. To clarify matters, this paper analyzes the impact of TRU CR on many parameters relevant to fuel cycle systems and therefore spans a broad range of topic areas. The analyses are based on a FR physics parameter scan of TRU CR from 0 to [approximately]1.8 in a sodium-cooled metal-fueled FR (SMFR), in which the fuel from uranium-oxide-fueled light water reactors (LWRs) is recycled directly to FRs and FRs displace LWRs in the fleet. In this instance, the FRs are sodium cooled and metal fueled. Generally, it is assumed that all TRU elements are recycled, which maximizes uranium ore utilization for a given TRU CR and waste radiotoxicity reduction and is consistent with the assumption of used metal fuel separated by electrochemical means. In these analyses, the fuel burnup was constrained by imposing a neutron fluence limit to fuel cladding to the same constant value. This paper first presents static, time-independent measures of performance for the LWR [right arrow] FR fuel cycle, including mass, heat, gamma emission, radiotoxicity, and the two figures of merit for materials for weapon attractiveness developed by C. Bathke et al. No new fuel cycle will achieve a static equilibrium in the foreseeable future. Therefore, additional analyses are shown with dynamic, time

  19. Measuring of fissile isotope partial antineutrino spectra in direct experiment at nuclear reactor

    SciTech Connect

    Sinev, V. V.

    2009-11-15

    The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta-decay reaction positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-proliferation purposes.

  20. The benefits of a fast reactor closed fuel cycle in the UK

    SciTech Connect

    Gregg, R.; Hesketh, K.

    2013-07-01

    The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size, so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the

  1. Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

    NASA Astrophysics Data System (ADS)

    Jaradat, Safwan Qasim Mohammad

    Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

  2. Apparatus and process to eliminate diffusional limitations in a membrane biological reactor by pressure cycling

    DOEpatents

    Efthymiou, George S.; Shuler, Michael L.

    1989-08-29

    An improved multilayer continuous biological membrane reactor and a process to eliminate diffusional limitations in membrane reactors in achieved by causing a convective flux of nutrient to move into and out of an immobilized biocatalyst cell layer. In a pressure cycled mode, by increasing and decreasing the pressure in the respective layers, the differential pressure between the gaseous layer and the nutrient layer is alternately changed from positive to negative. The intermittent change in pressure differential accelerates the transfer of nutrient from the nutrient layers to the biocatalyst cell layer, the transfer of product from the cell layer to the nutrient layer and the transfer of byproduct gas from the cell layer to the gaseous layer. Such intermittent cycling substantially eliminates mass transfer gradients in diffusion inhibited systems and greatly increases product yield and throughput in both inhibited and noninhibited systems.

  3. Impacts of Heterogeneous Recycle in Fast Reactors on Overall Fuel Cycle

    SciTech Connect

    Temitope A. Taiwo; Samuel E. Bays; Abdullatif M. Yacout; Edward M. Hoffman; Michael Todosow; Taek K. Kim; Massimo Salvatores

    2011-03-01

    A study in the United States has evaluated the attributes of the heterogeneous recycle approach for plutonium and minor actinide transmutation in fast reactor fuel cycles, with comparison to the homogeneous recycle approach, where pertinent. The work investigated the characteristics, advantages, and disadvantages of the approach in the overall fuel cycle, including reactor transmutation, systems and safety impacts, fuel separation and fabrication issues, and proliferation risk and transportation impacts. For this evaluation, data from previous and ongoing national studies on heterogeneous recycle were reviewed and synthesized. Where useful, information from international sources was included in the findings. The intent of the work was to provide a comprehensive assessment of the heterogeneous recycle approach at the current time.

  4. High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

    For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

  5. An evaluation of waste radiotoxicity reduction for a fast burner reactor closed fuel cycle: NEA benchmark results

    SciTech Connect

    Grimm, K.N.; Hill, R.N.; Wase, D.C.

    1995-12-01

    As part of a program proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed. In this paper, the fuel cycle performance of the metal-fueled benchmark is evaluated in detail. Benchmark results assess the reactor performance and toxicity behavior in a closed nuclear fuel cycle for a parametric variation of the conversion ratio between 0.5 and 1.0. Results indicate that a fast burner reactor closed fuel cycle can be utilized to significantly reduce the radiotoxicity destined for ultimate disposal.

  6. Maintenance Cycle Extension in the IRIS Advanced Light Water Reactor Plant Design

    SciTech Connect

    Galvin, Mark R.; Todreas, Neil E.; Conway, Larry E.

    2003-09-15

    New nuclear power generation in the United States will be realized only if the economic performance can be made competitive with other methods of electrical power generation. The economic performance of a nuclear power plant can be significantly improved by increasing the time spent on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described that can be used to resolve, in the design phase, maintenance-related operating cycle length barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the International Reactor, Innovative and Secure (IRIS) design. IRIS is an advanced light water nuclear power plant that is being designed to maximize this on-line generating time by increasing the operating cycle length. This is consequently a maintenance strategy paper using the IRIS plant as the example.Potential IRIS operating cycle length maintenance-related barriers, determined by modification of an earlier operating pressurized water reactor (PWR) plant cycle length analysis to account for differences between the design of IRIS and this operating PWR, are presented. The proposed methodology to resolve these maintenance-related barriers by the design process is described. The results of applying the methodology to two potential IRIS cycle length barriers, relief valve testing and emergency heat removal system testing, are presented.

  7. POWER CYCLE AND STRESS ANALYSES FOR HIGH TEMPERATURE GAS-COOLED REACTOR

    SciTech Connect

    Oh, Chang H; Davis, Cliff; Hawkes, Brian D; Sherman, Steven R

    2007-05-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with three turbines and four compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with three stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and a 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles to

  8. The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

    NASA Astrophysics Data System (ADS)

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

    2012-06-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. 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 and Thorium is 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 is filled by fresh natural Uranium/Thorium fuel. 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 and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

  9. The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

    SciTech Connect

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

    2012-06-06

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. 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 and Thorium is 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 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

  10. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 5 - North Anna Unit 1 Cycle 5

    SciTech Connect

    Bowman, S.M.

    1993-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor (AFR) criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial pressurized-water reactors (PWR). The analysis methodology selected for all calculations reported herein was the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted comparison of criticality calculations directly using the utility-calculated isotopics to those using the isotopics generated by the SCALE-4 SAS2H

  11. 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.

  12. Once-through thorium fuel cycle evaluation for TVA's Browns Ferry-3 Boiling Water Reactor

    SciTech Connect

    Hopkins, G.C.

    1982-05-01

    This report documents benchmark evaluations to test thorium lattice predictive methods and neutron cross sections against available data and summarizes specific evaluations of the once-through thorium cycle when applied to the Browns Ferry-3 BWR. It was concluded that appreciable uncertainties in thorium cycle nuclear data cloud the ability to reliably predict the fuel cycle performance and that power reactor irradiations of ThO/sub 2/ rods in BWRs are desirable to resolve uncertainties. Benchmark evaluations indicated that the ENDF/B-IV data used in the evaluations should cause an underprediction of U-233/ThO/sub 2/ fuel reactivity, and, therefore, the results of the preliminary evaluations completed under the program should be conservative.

  13. A Novel Fuel/Reactor Cycle to Implement the 300 Years Nuclear Waste Policy Approach - 12377

    SciTech Connect

    Carelli, M.D.; Franceschini, F.; Lahoda, E.J.; Petrovic, B.

    2012-07-01

    A thorium-based fuel cycle system can effectively burn the currently accumulated commercial used nuclear fuel and move to a sustainable equilibrium where the actinide levels in the high level waste are low enough to yield a radiotoxicity after 300 years lower than that of the equivalent uranium ore. The second step of the Westinghouse approach to solving the waste 'problem' has been completed. The thorium fuel cycle has indeed the potential of burning the legacy TRU and achieve the waste objective proposed. Initial evaluations have been started for the third step, development and selection of appropriate reactors. Indications are that the probability of show-stoppers is rather remote. It is, therefore, believed that development of the thorium cycle and associated technologies will provide a permanent solution to the waste management. Westinghouse is open to the widest collaboration to make this a reality. (authors)

  14. The benefits of an advanced fast reactor fuel cycle for plutonium management

    SciTech Connect

    Hannum, W.H.; McFarlane, H.F.; Wade, D.C.; Hill, R.N.

    1996-12-31

    The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium and long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a `focus area` for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed.

  15. Effect of temperature and cycle length on microbial competition in PHB-producing sequencing batch reactor.

    PubMed

    Jiang, Yang; Marang, Leonie; Kleerebezem, Robbert; Muyzer, Gerard; van Loosdrecht, Mark C M

    2011-05-01

    The impact of temperature and cycle length on microbial competition between polyhydroxybutyrate (PHB)-producing populations enriched in feast-famine sequencing batch reactors (SBRs) was investigated at temperatures of 20 °C and 30 °C, and in a cycle length range of 1-18 h. In this study, the microbial community structure of the PHB-producing enrichments was found to be strongly dependent on temperature, but not on cycle length. Zoogloea and Plasticicumulans acidivorans dominated the SBRs operated at 20 °C and 30 °C, respectively. Both enrichments accumulated PHB more than 75% of cell dry weight. Short-term temperature change experiments revealed that P. acidivorans was more temperature sensitive as compared with Zoogloea. This is particularly true for the PHB degradation, resulting in incomplete PHB degradation in P. acidivorans at 20 °C. Incomplete PHB degradation limited biomass growth and allowed Zoogloea to outcompete P. acidivorans. The PHB content at the end of the feast phase correlated well with the cycle length at a constant solid retention time (SRT). These results suggest that to establish enrichment with the capacity to store a high fraction of PHB, the number of cycles per SRT should be minimized independent of the temperature.

  16. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3

    SciTech Connect

    Bowman, S.M.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of three reactor critical configurations for the Sequoyah Unit 2 Cycle 3. This unit and cycle were chosen because of the relevance in spent fuel benchmark applications: (1) the unit had a significantly long downtime of 2.7 years during the middle of cycle (MOC) 3, and (2) the core consisted entirely of burned fuel at the MOC restart. The first benchmark critical calculation was the MOC restart at hot, full-power (HFP) critical conditions. The

  17. Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5

    SciTech Connect

    Bowman, S.M.; Suto, T. |

    1996-10-01

    ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k{sub eff} of 1. 0040{+-}0.0005.

  18. Feasibility Study on Thermal-Hydraulic Performance of Innovative Water Reactor for Flexible Fuel Cycle (FLWR)

    SciTech Connect

    Akira, Ohnuki; Kazuyuki, Takase; Masatoshi, Kureta; Hiroyuki, Yoshida; Hidesada, Tamai; Wei, Liu; Toru, Nakatsuka; Takeharu, Misawa; Hajime, Akimoto

    2006-07-01

    R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is started at Japan Atomic Energy Agency (JAEA) in collaboration with power company, reactor vendors, universities since 2002. The FLWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the FLWR because of the tight lattice configuration. In this paper, we will show the R and D plan and summarize experimental studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility. Most important objective of the large-scale test is to resolve a fundamental subject whether the core cooling under a tight-lattice configuration is feasible. The characteristics of critical power and flow behavior are investigated under different geometrical configuration and boundary conditions. The configuration parameter is the gap between rods (FY2004) and the rod bowing (FY2005). We have confirmed the thermal-hydraulic feasibility from the experimental results. (authors)

  19. Nutrient removal in a sequencing batch reactor operated with short anaerobic/aerobic cycles.

    PubMed

    Freitas, F; Temudo, M; Almeida, J S; Reis, M A M

    2003-01-01

    A single sequencing batch reactor operated with short intermittent aeration cycles was used to simultaneously remove carbon, nitrogen and phosphorus. The complete cycle, comprising feeding, anaerobiosis, aerobiosis, settling and decanting, was only 36 minutes long. The system has shown high and stable nutrient removal at 30 degrees C with acetate as carbon source and it has proved to be rather robust and dynamic, efficiently adapting to most of the changes in operating parameters tested: presence of nitrate in the feeding medium, different substrates (propionate and butyrate), temperature and nutrient shock loads. For the optimum conditions used, a removal efficiency of over 90% was obtained for each nutrient. Description of the population kinetics was obtained for each operating condition, by performing batch tests. Kinetic and stoichiometric parameters were used to infer the relative contribution of each group of microorganisms on SBR performance. Compared to the traditional SBR operated with cycles of 6 hours, the use of short intermittent aeration cycles of 36 minutes corresponds to a 40% reduction on aeration time.

  20. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis.

    PubMed

    Kunz, Ulrich; Turek, Thomas

    2009-11-30

    Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid.

  1. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis

    PubMed Central

    Turek, Thomas

    2009-01-01

    Summary Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid. PMID:20300506

  2. Nuclear reactor melt-retention structure to mitigate direct containment heating

    DOEpatents

    Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

    1991-01-01

    A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

  3. Modeling and Simulations for the High Flux Isotope Reactor Cycle 400

    SciTech Connect

    Ilas, Germina; Chandler, David; Ade, Brian J; Sunny, Eva E; Betzler, Benjamin R; Pinkston, Daniel

    2015-03-01

    A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. At this time, the core model used to perform analyses in support of HFIR operation is an MCNP model for the beginning of Cycle 400, which was documented in detail in a 2005 technical report. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed to serve as reference for the design of an LEU fuel for HFIR. The recent enhancements in modeling and simulations for HFIR that are discussed in the present report include: (1) revision of the 2005 MCNP model for the beginning of Cycle 400 to improve the modeling data and assumptions as necessary based on appropriate primary reference sources HFIR drawings and reports; (2) improvement of the fuel region model, including an explicit representation for the involute fuel plate geometry that is characteristic to HFIR fuel; and (3) revision of the Monte Carlo-based depletion model for HFIR in use since 2009 but never documented in detail, with the development of a new depletion model for the HFIR explicit fuel plate representation. The new HFIR models for Cycle 400 are used to determine various metrics of relevance to reactor performance and safety assessments. The calculated metrics are compared, where possible, with measurement data from preconstruction critical experiments at HFIR, data included in the current HFIR safety analysis report, and/or data from previous calculations performed with different methods or codes. The results of the analyses show that the models presented in this report provide a robust and reliable basis for HFIR analyses.

  4. Improving Cycling Performance: Transcranial Direct Current Stimulation Increases Time to Exhaustion in Cycling

    PubMed Central

    Bertollo, Maurizio; Boggio, Paulo Sergio; Fregni, Felipe

    2015-01-01

    The central nervous system seems to have an important role in fatigue and exercise tolerance. Novel noninvasive techniques of neuromodulation can provide insights on the relationship between brain function and exercise performance. The purpose of this study was to determine the effects of transcranial direct current stimulation (tDCS) on physical performance and physiological and perceptual variables with regard to fatigue and exercise tolerance. Eleven physically active subjects participated in an incremental test on a cycle simulator to define peak power output. During 3 visits, the subjects experienced 3 stimulation conditions (anodal, cathodal, or sham tDCS—with an interval of at least 48 h between conditions) in a randomized, counterbalanced order to measure the effects of tDCS on time to exhaustion at 80% of peak power. Stimulation was administered before each test over 13 min at a current intensity of 2.0 mA. In each session, the Brunel Mood State questionnaire was given twice: after stimulation and after the time-to-exhaustion test. Further, during the tests, the electromyographic activity of the vastus lateralis and rectus femoris muscles, perceived exertion, and heart rate were recorded. RM-ANOVA showed that the subjects performed better during anodal primary motor cortex stimulation (491 ± 100 s) compared with cathodal stimulation (443 ± 11 s) and sham (407 ± 69 s). No significant difference was observed between the cathodal and sham conditions. The effect sizes confirmed the greater effect of anodal M1 tDCS (anodal x cathodal = 0.47; anodal x sham = 0.77; and cathodal x sham = 0.29). Magnitude-based inference suggested the anodal condition to be positive versus the cathodal and sham conditions. There were no differences among the three stimulation conditions in RPE (p = 0.07) or heart rate (p = 0.73). However, as hypothesized, RM- ANOVA revealed a main effect of time for the two variables (RPE and HR: p < 0.001). EMG activity also did not differ

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

    NASA Astrophysics Data System (ADS)

    Dugan, Edward T.; Kutikkad, Kiratadas

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

  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. Population exposure from the fuel cycle: Review and future direction

    SciTech Connect

    Richmond, C.R.

    1987-01-01

    The legacy of radiation exposures confronting man arises from two historical sources of energy, the sun and radioactive decay. Contemporary man continues to be dependent on these two energy sources, which include the nuclear fuel cycle. Radiation exposures from all energy sources should be examined, with particular emphasis on the nuclear fuel cycle, incidents such as Chernobyl and Three Mile Island. In addition to risk estimation, concepts such as de minimis, life shortening as a measure of risk, and competing risks as projected into the future must be considered in placing radiation exposures in perspective. The utility of these concepts is in characterizing population exposures for decision makers in a manner that the public may judge acceptable. All these viewpoints are essential in the evaluation of population exposure from the nuclear fuel cycle.

  8. Progress and Future Directions in North American Carbon Cycle Science

    NASA Astrophysics Data System (ADS)

    Michalak, Anna; Huntzinger, Deborah; Shrestha, Gyami

    2013-05-01

    The North American Carbon Program (NACP) convened its fourth biennial "All Investigators" meeting (AIM4, http://www.nacarbon.org/meeting_2013) to review progress in understanding the dynamics of the carbon cycle of North America and adjacent oceans and to chart a course for a more integrative and holistic approach to future research. The meeting was structured around the six decadal goals outlined in the new "A U.S. Carbon Cycle Science Plan" (Michalak et al., University Corporation for Atmospheric Research, 2011, available at http://www.carboncyclescience.gov) and focused on (1) diagnosis of the atmospheric carbon cycle, (2) drivers of anthropogenic emissions, (3) vulnerability of carbon stocks to change, (4) ecosystem impacts of change, (5) carbon management, and (6) decision support.

  9. Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detector

    SciTech Connect

    Hellfeld, D.; Dazeley, S.; Bernstein, A.; Marianno, C.

    2015-11-25

    The potential of elastic antineutrino-electron scattering (ν¯e + e → ν¯e + e) in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13 km standoff from a 3.758 GWt light water nuclear reactor. Background was estimated via independent simulations and by appropriately scaling published measurements from similar detectors. Many potential backgrounds were considered, including solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclide and water-borne radon decays, and gamma rays from the photomultiplier tubes, detector walls, and surrounding rock. The detector response was modeled using a GEANT4-based simulation package. The results indicate that with the use of low radioactivity PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. The directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. Lastly, the results provide a list of theoretical conditions that, if satisfied in practice, would enable nuclear reactor antineutrino directionality in a Gd-doped water Cherenkov detector approximately 10 km from a large power reactor.

  10. Use of cermet fueled nuclear reactors for direct nuclear propulsion

    SciTech Connect

    Bhattacharyya, S.K.; Carlson, L.W.; Kuczen, K.D.; Hanan, N.A.; Palmer, R.G.; Von Hoomissen, J.; Chiu, W.; Haaland, R.

    1988-07-01

    There has been a renewal of interest in Direct Nuclear Propulsion (DNP) because of the Air Force Forecast II recommendation for the development of the technology. Several nuclear concepts have been proposed to meet the Direct Nuclear Propulsion challenge. In this paper we will present results of an initial study of the potential of a cermet fueled nuclear system in providing the desired DNP capabilities and featuring a set of unique safety characteristics. The concept of cermet fuel for DNP applications was first developed by ANL and GE working independently more than 20 years ago. The two organizations came to several remarkably consistent conclusions. The present work has consisted of collecting a unified set of design parameters from the set of design results produced in the earlier work. The conclusion of this exercise was that a cermet-fueled DNP design looked extremely promising from performance and safety considerations and that it deserves serious consideration when the decision to develop one or more concepts for DNP is made.

  11. Development of a Scale Model for High Flux Isotope Reactor Cycle 400

    SciTech Connect

    Ilas, Dan

    2012-03-01

    The development of a comprehensive SCALE computational model for the High Flux Isotope Reactor (HFIR) is documented and discussed in this report. The SCALE model has equivalent features and functionality as the reference MCNP model for Cycle 400 that has been used extensively for HFIR safety analyses and for HFIR experiment design and analyses. Numerical comparisons of the SCALE and MCNP models for the multiplication constant, power density distribution in the fuel, and neutron fluxes at several locations in HFIR indicate excellent agreement between the results predicted with the two models. The SCALE HFIR model is presented in sufficient detail to provide the users of the model with a tool that can be easily customized for various safety analysis or experiment design requirements.

  12. Preliminary design of ultra-long cycle fast reactor employing breed-and-burn strategy

    SciTech Connect

    Tak, T. W.; Yu, H.; Kim, J. H.; Lee, D.; Kim, T. K.

    2012-07-01

    A new design of ultra-long cycle fast reactor with power rate of 1000 MWe (UCFR) has been developed based on the strategy of breed-and burn. The bottom region of the core with low enriched uranium (LEU) plays a role of igniter of the core burning and the upper natural uranium (NU) region acts as blanket for breeding. Fissile materials are bred in the blanket and the active core moves upward at a speed of 5.4 cm/year. Through the core depletion calculation using Monte Carlo code, McCARD, it is confirmed that a full power operation of 60 years without refueling is feasible. Core performance characteristics have been evaluated in terms of axial/radial power shapes, reactivity feedback coefficients, etc. This design will serve as a base model for further design study of UCFRs using LWR spent fuels in the blanket region. (authors)

  13. Design of a Simplified Closed Brayton Cycle for a Space Reactor Application

    SciTech Connect

    Guimaraes, Lamartine N. F.; Camillo, Giannino Ponchio; Placco, Guilherme Moreira

    2009-03-16

    The Nuclear Energy Division (ENU) of the Institute for Advanced Studies (IEAv) has started a preliminary design study for a Closed Brayton Cycle Loop (CBCL) aimed at a space reactor application. The main objectives of the study are: 1) to establish a starting concept for the CBCL components specifications, and 2) to build a demonstrative simulator of CBCL. This preliminary design study is been developed around the NOELLE 60290 turbo machine. The actual nuclear reactor study is being conducted independently. Because of that, a conventional heat source is being used for the CBCL, in this preliminary design phase. This paper describes details of the CBCL mechanical design and the steady state simulator of the CBCL operating with NOELLE 60290 turbo machine. In principle, several gases are being considered as working fluid, as for instance: air, helium, nitrogen, CO2 and gas mixtures such as helium and xenon. However, for this first application pure helium will be used as working fluid. Simplified models of heat and mass transfer were developed to simulate thermal components. A new graphical interface was developed for the simulator to display the thermal process variables in steady state and to keep track of the modifications being implemented at the NOELLE 60290 turbo machine in order to build the CBCL. A set of new results are being produced. These new results help to establish the hot and cold source geometry allowing for price estimating costs for building the actual device. These fresh new results will be presented and discussed.

  14. Flexible Fuel Cycle Initiative for the Harmonized Deployment of Gen-IV Reactors

    NASA Astrophysics Data System (ADS)

    Fukasawa, Tetsuo; Yamashita, Junichi; Hoshino, Kuniyoshi; Fujimura, Koji; Sasahira, Akira

    Generation IV type fast reactors (FR) are expected to be commercially deployed instead of light water reactors (LWR) from around 2050. Replacement of LWR to FR needs flexibility due to uncertain factors such as FR deployment rate which affects the FR fuel (Pu) supply amount from LWR spent fuel reprocessing and the capacity of related facilities. If the FR deployment rate is as currently planned, more Pu must be prepared by expanding LWR reprocessing. If the FR deployment rate decreases, LWR reprocessing must be reduced to avoid excess Pu. To cope with this issue we proposed the innovative system called Flexible Fuel Cycle Initiative (FFCI) that has integral reprocessing for LWR and FR spent fuels. LWR reprocessing in FFCI only carries out about 90% U recovery and residual material with Pu, U (˜5%), minor actinides (MA) and fission products (FP) goes to FR reprocessing for the planned FR deployment rate. For any decrease in the FR deployment rate temporary storage will be used. Coexistence of Pu/U with MA and FP until just before Pu/U usage in the FR provides high proliferation resistance. Preliminary evaluation revealed that FFCI can reduce the LWR reprocessing capacity and LWR spent fuel storage amount compared with current plan (reference system) if the FR deployment rate decreases. Several FR deployment scenarios and countermeasures such as FFCI were investigated.

  15. Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond

    SciTech Connect

    Ovanes, M.; Chan, P. S. W.; Mao, J.; Alderson, N.; Hopwood, J. M.

    2012-07-01

    The Enhanced CANDU 6{sup R} (ECo{sup R}) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

  16. Feasibility study of fuel cladding performance for application in ultra-long cycle fast reactor

    NASA Astrophysics Data System (ADS)

    Jung, Ju Ang; Kim, Seung Hyun; Shin, Sang Hun; Bang, In Cheol; Kim, Ji Hyun

    2013-09-01

    As a part of the research and development activities for long-life core sodium-cooled fast reactors, the cladding performance of the ultra-long cycle fast reactor (UCFR) is evaluated with two design power levels (1000 MWe and 100 MWe) and cladding peak temperatures (873 K and 923 K). The key design concept of the UCFR is that it is non-refueling during its 30-60 years of operation. This concept may require a maximum peak cladding temperature of 923 K and a cladding radiation damage of over 200 dpa (displacements per atom). Therefore, for the design of the UCFR, deformation due to thermal creep, irradiation creep, and swelling must be taken into consideration through quantitative evaluations. As candidate cladding materials for use in UCFRs, ferritic-martensitic (FM) steels, oxide dispersion strengthened (ODS) steels, and SiC-based composite materials are studied using deformation behavior modeling for a feasibility evaluation. The results of this study indicate that SiC is a potential UCFR cladding material, with the exception of irradiation creep due to high neutron fluence stemming from its long operating time of about 30-60 years.

  17. Effect of cycle changes on simultaneous biological nutrient removal in a sequencing batch reactor (SBR).

    PubMed

    Coma, M; Puig, S; Monclús, H; Balaguer, M D; Colprim, J

    2010-03-01

    The destabilization of a microbial population is sometimes hard to solve when different biological reactions are coupled in the same reactor as in sequencing batch reactors (SBRs). This paper will try to guide through practical experiences the recovery of simultaneous nitrogen and phosphorus removal in an SBR after increasing the demand of wastewater treatment by taking advantage of its flexibility. The results demonstrate that the length of phases and the optimization of influent distribution are key factors in stabilizing the system for long-term periods with high nutrient removal (88%, 93% and 99% of carbon, nitrogen and phosphorus, respectively). In order to recover a biological nutrient removal (BNR) system, different interactions such as simultaneous nitrification and denitrification and also phosphorus removal must be taken into account. As a general conclusion, it can be stated there is no such thing as a perfect SBR operation, and that much will depend on the state of the BNR system. Hence, the SBR operating strategy must be based on a dynamic cycle definition in line with process efficiency. PMID:20426270

  18. A Preliminary and Simplified Closed Brayton Cycle Modeling for a Space Reactor Application

    SciTech Connect

    Guimaraes, Lamartine Nogueira Frutuoso; Camillo, Giannino Ponchio

    2008-01-21

    The Nuclear Energy Division (ENU) of the Institute for Advanced Studies (IEAv) has started a preliminary design study for a Closed Brayton Cycle Loop (CBCL) aimed at a space reactor application. The main objectives of the study are: 1) to establish a starting concept for the CBCL components specifications, and 2) to build a demonstrative simulator of CBCL. This preliminary design study is developing the CBCL around the NOELLE 60290 turbo machine. The actual nuclear reactor study is being conducted independently. Because of that, a conventional heat source is being used for the CBCL, in this preliminary design phase. This paper describes the steady state simulator of the CBCL operating with NOELLE 60290 turbo machine. In principle, several gases are being considered as working fluid, as for instance: air, helium, nitrogen, CO{sub 2} and gas mixtures such as helium and xenon. However, for this first application pure helium will be used as working fluid. Simplified models of heat and mass transfer were developed to simulate thermal components. Future efforts will focus on implementing a graphical interface to display the thermal process variables in steady state and to keep track of the modifications being implemented at the NOELLE 60290 turbo machine in order to build the CBCL.

  19. DIRECT ENERGY CONVERSION FISSION REACTOR FOR THE PERIOD JANUARY 1, 2002 THROUGH MARCH 31, 2002

    SciTech Connect

    L.C. BROWN

    2002-03-31

    Direct energy conversion is the only potential means for producing electrical energy from a fission reactor without the Carnot efficiency limitations. This project was undertaken by Sandia National Laboratories, Los Alamos National Laboratories, The University of Florida, Texas A&M University and General Atomics to explore the possibilities of direct energy conversion. Other means of producing electrical energy from a fission reactor, without any moving parts, are also within the statement of proposed work. This report documents the efforts of General Atomics. Sandia National Laboratories, the lead laboratory, provides overall project reporting and documentation. The highlights of this reporting period are: (1) Cooling of the vapor core reactor and the MHD generator was incorporated into the Vapor Core Reactor model using standard heat transfer calculation methods. (2) Fission product removal, previously modeled as independent systems for each class of fission product, was incorporated into the overall fuel recycle loop of the Vapor Core Reactor. The model showed that the circulating activity levels are quite low. (3) Material distribution calculations were made for the ''pom-pom'' style cathode for the Fission Electric Cell. Use of a pom-pom cathode will eliminate the problem of hoop stress in the thin spherical cathode caused by the electric field.

  20. Heuristic optimization of pressurized water reactor fuel cycle design under general constraints

    SciTech Connect

    Moon, H.; Levine, S.H. ); Mahgerefteh, M. )

    1989-12-01

    Optimization techniques in fuel management have directed modern fuel cycle designs to use low-leakage loading patterns. Future optimization calculations involving low-leakage patterns must utilize nucleonic models that are both fast operationally and rigorous. A two-dimensional two-group diffusion theory code is developed and lattice homogenization constants are generated using a modified LEOPARD code to fulfill these criteria. Based on these two codes, a heuristic optimization study is performed that considers the general constraints (e.g., spent-fuel storage limit and mechanical burnup limit) given to a utility fuel cycle designer. The optimum cycle length that minimizes the fuel cost is {approximately} 600 effective full-power days for the conditions assumed.

  1. Parametric analyses of single-zone thorium-fueled molten salt reactor fuel cycle options

    SciTech Connect

    Powers, J.J.; Worrall, A.; Gehin, J.C.; Harrison, T.J.; Sunny, E.E.

    2013-07-01

    Analyses of fuel cycle options based on thorium-fueled Molten Salt Reactors (MSRs) have been performed in support of fuel cycle screening and evaluation activities for the United States Department of Energy. The MSR options considered are based on thermal spectrum MSRs with 3 different separations levels: full recycling, limited recycling, and 'once-through' operation without active separations. A single-fluid, single-zone 2250 MWth (1000 MWe) MSR concept consisting of a fuel-bearing molten salt with graphite moderator and reflectors was used as the basis for this study. Radiation transport and isotopic depletion calculations were performed using SCALE 6.1 with ENDF/B-VII nuclear data. New methodology developed at Oak Ridge National Laboratory (ORNL) enables MSR analysis using SCALE, modeling material feed and removal by taking user-specified parameters and performing multiple SCALE/TRITON simulations to determine the resulting equilibrium operating conditions. Parametric analyses examined the sensitivity of the performance of a thorium MSR to variations in the separations efficiency for protactinium and fission products. Results indicate that self-sustained operation is possible with full or limited recycling but once-through operation would require an external neutron source. (authors)

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

    NASA Astrophysics Data System (ADS)

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

    2010-06-01

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

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

    SciTech Connect

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

    2010-06-22

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

  4. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, George P.

    1988-01-01

    A high-power-density laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems.

  5. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, G.P.

    1987-02-20

    A high-power-density-laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems. 25 figs.

  6. Direct conversion of methane to aromatics in a catalytic co-ionic membrane reactor.

    PubMed

    Morejudo, S H; Zanón, R; Escolástico, S; Yuste-Tirados, I; Malerød-Fjeld, H; Vestre, P K; Coors, W G; Martínez, A; Norby, T; Serra, J M; Kjølseth, C

    2016-08-01

    Nonoxidative methane dehydroaromatization (MDA: 6CH4 ↔ C6H6 + 9H2) using shape-selective Mo/zeolite catalysts is a key technology for exploitation of stranded natural gas reserves by direct conversion into transportable liquids. However, this reaction faces two major issues: The one-pass conversion is limited by thermodynamics, and the catalyst deactivates quickly through kinetically favored formation of coke. We show that integration of an electrochemical BaZrO3-based membrane exhibiting both proton and oxide ion conductivity into an MDA reactor gives rise to high aromatic yields and improved catalyst stability. These effects originate from the simultaneous extraction of hydrogen and distributed injection of oxide ions along the reactor length. Further, we demonstrate that the electrochemical co-ionic membrane reactor enables high carbon efficiencies (up to 80%) that improve the technoeconomic process viability.

  7. Direct conversion of methane to aromatics in a catalytic co-ionic membrane reactor.

    PubMed

    Morejudo, S H; Zanón, R; Escolástico, S; Yuste-Tirados, I; Malerød-Fjeld, H; Vestre, P K; Coors, W G; Martínez, A; Norby, T; Serra, J M; Kjølseth, C

    2016-08-01

    Nonoxidative methane dehydroaromatization (MDA: 6CH4 ↔ C6H6 + 9H2) using shape-selective Mo/zeolite catalysts is a key technology for exploitation of stranded natural gas reserves by direct conversion into transportable liquids. However, this reaction faces two major issues: The one-pass conversion is limited by thermodynamics, and the catalyst deactivates quickly through kinetically favored formation of coke. We show that integration of an electrochemical BaZrO3-based membrane exhibiting both proton and oxide ion conductivity into an MDA reactor gives rise to high aromatic yields and improved catalyst stability. These effects originate from the simultaneous extraction of hydrogen and distributed injection of oxide ions along the reactor length. Further, we demonstrate that the electrochemical co-ionic membrane reactor enables high carbon efficiencies (up to 80%) that improve the technoeconomic process viability. PMID:27493179

  8. Enhancing VHTR passive safety and economy with thermal radiation based direct reactor auxiliary cooling system

    SciTech Connect

    Zhao, H.; Zhang, H.; Zou, L.; Sun, X.

    2012-07-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The RVACS can be characterized as a surface-based decay heat removal system. It is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to the core volume) and decay heat removal capability (proportional to the vessel surface area). Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environmental side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps or annular regions formed between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions among the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very

  9. Enhancing VHTR Passive Safety and Economy with Thermal Radiation Based Direct Reactor Auxiliary Cooling System

    SciTech Connect

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or

  10. Direct Estimation of Power Distribution in Reactors for Nuclear Thermal Space Propulsion

    NASA Astrophysics Data System (ADS)

    Aldemir, Tunc; Miller, Don W.; Burghelea, Andrei

    2004-02-01

    A recently proposed constant temperature power sensor (CTPS) has the capability to directly measure the local power deposition rate in nuclear reactor cores proposed for space thermal propulsion. Such a capability reduces the uncertainties in the estimated power peaking factors and hence increases the reliability of the nuclear engine. The CTPS operation is sensitive to the changes in the local thermal conditions. A procedure is described for the automatic on-line calibration of the sensor through estimation of changes in thermal .conditions.

  11. Direct Estimation of Power Distribution in Reactors for Nuclear Thermal Space Propulsion

    SciTech Connect

    Aldemir, Tunc; Miller, Don W.; Burghelea, Andrei

    2004-02-04

    A recently proposed constant temperature power sensor (CTPS) has the capability to directly measure the local power deposition rate in nuclear reactor cores proposed for space thermal propulsion. Such a capability reduces the uncertainties in the estimated power peaking factors and hence increases the reliability of the nuclear engine. The CTPS operation is sensitive to the changes in the local thermal conditions. A procedure is described for the automatic on-line calibration of the sensor through estimation of changes in thermal conditions.

  12. Data Reconciliation in the Steam-Turbine Cycle of a Boiling Water Reactor

    SciTech Connect

    Sunde, Svein; Berg, Oivind; Dahlberg, Lennart; Fridqvist, Nils-Olof

    2003-08-15

    A mathematical model for a boiling water reactor steam-turbine cycle was assembled by means of a configurable, steady-state modeling tool TEMPO. The model was connected to live plant data and intermittently fitted to these by minimization of a weighted least-squares object function. The improvement in precision achieved by this reconciliation was assessed from quantities calculated from the model equations linearized around the minimum and from Monte Carlo simulations. It was found that the inclusion of the flow-passing characteristics of the turbines in the model equations significantly improved the precision as compared to simple mass and energy balances, whereas heat transfer calculations in feedwater heaters did not. Under the assumption of linear model equations, the quality of the fit can also be expressed as a goodness-of-fit Q. Typical values for Q were in the order of 0.9. For a validated model Q may be used as a fault detection indicator, and Q dropped to very low values in known cases of disagreement between the model and the plant state. The sensitivity of Q toward measurement faults is discussed in relation to redundancy. The results of the linearized theory and Monte Carlo simulations differed somewhat, and if a more accurate analysis is required, this is better based on the latter. In practical application of the presently employed techniques, however, assessment of uncertainties in raw data is an important prerequisite.

  13. The thermodynamics of pyrochemical processes for liquid metal reactor fuel cycles

    SciTech Connect

    Johnson, I.

    1987-01-01

    The thermodynamic basis for pyrochemical processes for the recovery and purification of fuel for the liquid metal reactor fuel cycle is described. These processes involve the transport of the uranium and plutonium from one liquid alloy to another through a molten salt. The processes discussed use liquid alloys of cadmium, zinc, and magnesium and molten chloride salts. The oxidation-reduction steps are done either chemically by the use of an auxiliary redox couple or electrochemically by the use of an external electrical supply. The same basic thermodynamics apply to both the salt transport and the electrotransport processes. Large deviations from ideal solution behavior of the actinides and lanthanides in the liquid alloys have a major influence on the solubilities and the performance of both the salt transport and electrotransport processes. Separation of plutonium and uranium from each other and decontamination from the more noble fission product elements can be achieved using both transport processes. The thermodynamic analysis is used to make process design computations for different process conditions.

  14. Effect of reactor temperature on direct growth of carbon nanomaterials on stainless steel

    NASA Astrophysics Data System (ADS)

    Edzatty, A. N.; Syazwan, S. M.; Norzilah, A. H.; Jamaludin, S. B.

    2016-07-01

    Currently, carbon nanomaterials (CNMs) are widely used for various applications due to their extraordinary electrical, thermal and mechanical properties. In this work, CNMs were directly grown on the stainless steel (SS316) via chemical vapor deposition (CVD). Acetone was used as a carbon source and argon was used as carrier gas, to transport the acetone vapor into the reactor when the reaction occurred. Different reactor temperature such as 700, 750, 800, 850 and 900 °C were used to study their effect on CNMs growth. The growth time and argon flow rate were fixed at 30 minutes and 200 ml/min, respectively. Characterization of the morphology of the SS316 surface after CNMs growth using Scanning Electron Microscopy (SEM) showed that the diameter of grown-CNMs increased with the reactor temperature. Energy Dispersive X-ray (EDX) was used to analyze the chemical composition of the SS316 before and after CNMs growth, where the results showed that reduction of catalyst elements such as iron (Fe) and nickel (Ni) at high temperature (700 - 900 °C). Atomic Force Microscopy (AFM) analysis showed that the nano-sized hills were in the range from 21 to 80 nm. The best reactor temperature to produce CNMs was at 800 °C.

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

    NASA Astrophysics Data System (ADS)

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

    2003-01-01

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

  16. 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.

  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. 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.

  19. 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)

  20. Bacterial structure of aerobic granules is determined by aeration mode and nitrogen load in the reactor cycle.

    PubMed

    Cydzik-Kwiatkowska, Agnieszka

    2015-04-01

    This study investigated how the microbial composition of biomass and kinetics of nitrogen conversions in aerobic granular reactors treating high-ammonium supernatant depended on nitrogen load and the number of anoxic phases in the cycle. Excellent ammonium removal and predomination of full nitrification was observed in the reactors operated at 1.1 kg TKN m(-3) d(-1) and with anoxic phases in the cycle. In all reactors, Proteobacteria and Actinobacteria predominated, comprising between 90.14% and 98.59% of OTUs. Extracellular polymeric substances-producing bacteria, such as Rhodocyclales, Xanthomonadaceae, Sphingomonadales and Rhizobiales, were identified in biomass from all reactors, though in different proportions. Under constant aeration, bacteria capable of autotrophic nitrification were found in granules, whereas under variable aeration heterotrophic nitrifiers such as Pseudomonas sp. and Paracoccus sp. were identified. Constant aeration promoted more even bacteria distribution among taxa; with 1 anoxic phase, Paracoccus aminophilus predominated (62.73% of OTUs); with 2 phases, Corynebacterium sp. predominated (65.10% of OTUs).

  1. Progress in understanding of direct containment heating phenomena in pressurized light water reactors

    SciTech Connect

    Ginsberg, T.; Tutu, N.K.

    1988-01-01

    Progress is described in development of a mechanistic understanding of direct containment heating phemonena arising during high-pressure melt ejection accidents in pressurized water reactor systems. The experimental data base is discussed which forms the basis for current assessments of containment pressure response using current lumped-parameter containment analysis methods. The deficiencies in available methods and supporting data base required to describe major phenomena occurring in the reactor cavity, intermediate subcompartments and containment dome are highlighted. Code calculation results presented in the literature are cited which demonstrate that the progress in understanding of DCH phenomena has also resulted in current predictions of containment pressure loadings which are significantly lower than are predicted by idealized, thermodynamic equilibrium calculations. Current methods are, nonetheless, still predicting containment-threatening loadings for large participating melt masses under high-pressure ejection conditions. Recommendations for future research are discussed. 36 refs., 5 figs., 1 tab.

  2. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, G.P.

    1983-09-29

    The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.

  3. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    SciTech Connect

    Lv, Quiping; Sun, Xiaodong; Chtistensen, Richard; Blue, Thomas; Yoder, Graydon; Wilson, Dane

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  4. Direct In Situ Quantification of HO2 from a Flow Reactor.

    PubMed

    Brumfield, Brian; Sun, Wenting; Ju, Yiguang; Wysocki, Gerard

    2013-03-21

    The first direct in situ measurements of hydroperoxyl radical (HO2) at atmospheric pressure from the exit of a laminar flow reactor have been carried out using mid-infrared Faraday rotation spectroscopy. HO2 was generated by oxidation of dimethyl ether, a potential renewable biofuel with a simple molecular structure but rich low-temperature oxidation chemistry. On the basis of the results of nonlinear fitting of the experimental data to a theoretical spectroscopic model, the technique offers an estimated sensitivity of <1 ppmv over a reactor exit temperature range of 398-673 K. Accurate in situ measurement of this species will aid in quantitative modeling of low-temperature and high-pressure combustion kinetics. PMID:26291349

  5. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14

    SciTech Connect

    Schneider, K.J.

    1982-09-01

    The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation.

  6. Modular assembly for supporting, straining, and directing flow to a core in a nuclear reactor

    DOEpatents

    Pennell, William E.

    1977-01-01

    A reactor core support arrangement for supporting, straining, and providing fluid flow to the core and periphery of a nuclear reactor during normal operation. A plurality of removable inlet modular units are contained within permanent liners in the lower supporting plate of the reactor vessel lower internals. During normal operation (1) each inlet modular unit directs main coolant flow to a plurality of core assemblies, the latter being removably supported in receptacles in the upper portion of the modular unit and (2) each inlet modular unit may direct bypass flow to a low pressure annular region of the reactor vessel. Each inlet modular unit may include special fluid seals interposed between mating surfaces of the inlet modular units and the core assemblies and between the inlet modular units and the liners, to minimize leakage and achieve an hydraulic balance. Utilizing the hydraulic balance, the modular units are held in the liners and the assemblies are held in the modular unit receptacles by their own respective weight. Included as part of the permanent liners below the horizontal support plate are generally hexagonal axial debris barriers. The axial debris barriers collectively form a bottom boundary of a secondary high pressure plenum, the upper boundary of which is the bottom surface of the horizontal support plate. Peripheral liners include radial debris barriers which collectively form a barrier against debris entry radially. During normal operation primary coolant inlet openings in the liner, below the axial debris barriers, pass a large amount of coolant into the inlet modular units, and secondary coolant inlet openings in the portion of the liners within the secondary plenum pass a small amount of coolant into the inlet modular units. The secondary coolant inlet openings also provide alternative coolant inlet flow paths in the unlikely event of blockage of the primary inlet openings. The primary inlet openings have characteristics which limit the

  7. Importance of the (n,gamma) Cm-247 Evaluation on Neutron Emission in Fast Reactor Fuel Cycle Analysis

    SciTech Connect

    Benoit Forget; Mehdi Asgari; Rodolfo M. Ferrer

    2007-11-01

    As part of the GNEP program, it is envisioned to build a fast reactor for the transmutation of minor actinides. The spent nuclear fuel from the current fleet of light water reactors would be recycled, the current baseline is the UREX+1a process, and would act as a feed for the fast reactor. As the fuel is irradiated in a fast reactor a certain quantity of minor actinides would thus build up in the fuel stream creating possible concerns with the neutron emission of these minor actinides for fuel transportation, handling and fabrication. Past neutronic analyses had not tracked minor actinides above Cm-246 in the transmutation chain, because of the small influence on the overall reactor performance and cycle parameters. However, when trying to quantify the neutron emission from the recycled fuel with high minor actinide content, these higher isotopes play an essential role and should be included in the analysis. In this paper, the influence of tracking these minor actinides on the calculated neutron emission is presented. Also presented is the particular influence of choosing a different evaluated cross section data set to represent the minor actinides above Cm-246. The first representation uses the cross-sections provided by MC2-2 for all isotopes, while the second representation uses infinitely diluted ENDF/BVII.0 cross-sections for Cm-247 to Cf-252 and MC2-2 for all other isotopes.

  8. Steam Generator Component Model in a Combined Cycle of Power Conversion Unit for Very High Temperature Gas-Cooled Reactor

    SciTech Connect

    Oh, Chang H; Han, James; Barner, Robert; Sherman, Steven R

    2007-06-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP), Very High Temperature Gas-Cooled Reactor (VHTR) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. A combined cycle is considered as one of the power conversion units to be coupled to the very high-temperature gas-cooled reactor (VHTR). The combined cycle configuration consists of a Brayton top cycle coupled to a Rankine bottoming cycle by means of a steam generator. A detailed sizing and pressure drop model of a steam generator is not available in the HYSYS processes code. Therefore a four region model was developed for implementation into HYSYS. The focus of this study was the validation of a HYSYS steam generator model of two phase flow correlations. The correlations calculated the size and heat exchange of the steam generator. To assess the model, those calculations were input into a RELAP5 model and its results were compared with HYSYS results. The comparison showed many differences in parameters such as the heat transfer coefficients and revealed the different methods used by the codes. Despite differences in approach, the overall results of heat transfer were in good agreement.

  9. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    SciTech Connect

    Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

    2013-07-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  10. Short Contact Time Direct Coal Liquefaction Using a Novel Batch Reactor

    SciTech Connect

    He Huang; Michael T. Klein; William H. Calkins

    1997-01-30

    The primary objective of this research is to optimize the design and operation of the bench scale batch reactor (SCTBR) for studying direct coal liquefaction at short contact times (.01 to 10 minutes or longer). Additional objectives are to study the kinetics of direct coal liquefaction particularly at short reaction times and to investigate the role of organic oxygen components of coal and their reaction pathways during coal liquefaction. Many of those objectives have already been achieved. This quarterly report discusses further kinetic studies of the liquefaction in tetralin of a Montana Lignite, Wyodak-Anderson subbituminous coal, Illinois #6 hv bituminous coal, Pittsburgh #8 hv bituminous coals, and Pocohontas lV bituminous coal at short contact times. All of these coals showed a distinct extraction stage. Further work has also been done to attempt to clarify the role of the liquefaction solvent in the direct liquefaction process.

  11. Engine-cycle analysis for a particle-bed reactor nuclear rocket. Final report, May-Jul 90

    SciTech Connect

    Suzuki, D.E.

    1991-03-01

    This report addresses three candidate engine cycles for a particle bed nuclear rocket; bleed cycle with uncooled carbon - carbon composite nozzle; bleed cycle with regeneratively cooled Aluminum nozzle; expander cycle with regeneratively cooled Aluminum nozzle. The analysis was performed using the SALT System Analysis Language Translator code with the following amendments; particle bed reactor was modeled as a simple heater; a regeneratively cooled nozzle model was added which includes the heating of the coolant due to hot exhaust gases and nuclear heating of nozzle. The conclusion of the analysis were the topping cycle should be pursued for Mars missions and the bleed cycle should be pursued for OTV (Orbital Transfer Vehicle) missions. This study indicates that a regeneratively cooled aluminum nozzle can be sufficiently cooled to allow its use with a PBR rocket engine. This result is based on nozzle heating due to hot exhaust gases at a maximum chamber temperature and nuclear heating effects. The highest temperatures occur at the nozzle throat, where a composite or alloy coating could protect the aluminum. Further investigation of nozzle cooling should include modeling the nozzle with more nodes, and including more accurate dimensions for the nozzle wall thicknesses and coolant flow passages. The study also indicates that an expander cycle with a cooled aluminum nozzle can operate with a high pressure PBR at realistic TPA efficiencies. Further investigation should include the improvements to the regeneratively cooled nozzle model and more accurate performance maps for the TPA components.

  12. Role of fast reactor and its cycle to reduce nuclear waste burden

    SciTech Connect

    Arie, Kazuo; Oomori, Takashi; Okita, Takeshi; Kawashima, Masatoshi; Kotake, Shoji; Fuji-ie, Yoichi

    2013-07-01

    The role of the metal fuel fast reactor with recycling of actinides and the five long-lived fission products based on the concept of the Self-Consistent Nuclear Energy System has been examined by evaluating the reduction of nuclear wastes during the transition period to this reactor system. The evaluation was done in comparison to an LWR once-through case and a conventional actinide recycling oxide fast reactor. As a result, it is quantitatively clarified that a metal fuel fast reactor with actinide and the five long-lived fission products (I{sup 129}, Tc{sup 99}, Zr{sup 93}, Cs{sup 135} and Sn{sup 126}) recycling could play a significant role in reducing the nuclear waste burden including the current LWR wastes. This can be achieved by using a fast neutron spectrum reactor enhanced with metal fuel that brings high capability as a 'waste burner'. (authors)

  13. ENERGY EFFICIENCY LIMITS FOR A RECUPERATIVE BAYONET SULFURIC ACID DECOMPOSITION REACTOR FOR SULFUR CYCLE THERMOCHEMICAL HYDROGEN PRODUCTION

    SciTech Connect

    Gorensek, M.; Edwards, T.

    2009-06-11

    A recuperative bayonet reactor design for the high-temperature sulfuric acid decomposition step in sulfur-based thermochemical hydrogen cycles was evaluated using pinch analysis in conjunction with statistical methods. The objective was to establish the minimum energy requirement. Taking hydrogen production via alkaline electrolysis with nuclear power as the benchmark, the acid decomposition step can consume no more than 450 kJ/mol SO{sub 2} for sulfur cycles to be competitive. The lowest value of the minimum heating target, 320.9 kJ/mol SO{sub 2}, was found at the highest pressure (90 bar) and peak process temperature (900 C) considered, and at a feed concentration of 42.5 mol% H{sub 2}SO{sub 4}. This should be low enough for a practical water-splitting process, even including the additional energy required to concentrate the acid feed. Lower temperatures consistently gave higher minimum heating targets. The lowest peak process temperature that could meet the 450-kJ/mol SO{sub 2} benchmark was 750 C. If the decomposition reactor were to be heated indirectly by an advanced gas-cooled reactor heat source (50 C temperature difference between primary and secondary coolants, 25 C minimum temperature difference between the secondary coolant and the process), then sulfur cycles using this concept could be competitive with alkaline electrolysis provided the primary heat source temperature is at least 825 C. The bayonet design will not be practical if the (primary heat source) reactor outlet temperature is below 825 C.

  14. Scaling analysis for the direct reactor auxiliary cooling system for FHRs

    SciTech Connect

    Lv, Q.; Kim, I. H.; Sun, X.; Christensen, R. N.; Blue, T. E.; Yoder, G.; Wilson, D.; Sabharwall, P.

    2015-04-01

    The Direct Reactor Auxiliary Cooling System (DRACS) is a passive residual heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three natural circulation/convection loops that rely on buoyancy as the driving force and are coupled via two heat exchangers, namely, the DRACS heat exchanger and the natural draft heat exchanger. A fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during reactor normal operation, and to activate the DRACS in accidents when the reactor is shut down. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. In this paper, a detailed scaling analysis for the DRACS is performed, which will provide guidance for the design of scaled-down DRACS test facilities. Based on the Boussinesq assumption and one-dimensional flow formulation, the governing equations are non-dimensionalized by introducing appropriate dimensionless parameters. The key dimensionless numbers that characterize the DRACS system are obtained from the non-dimensional governing equations. Based on the dimensionless numbers and non-dimensional governing equations, similarity laws are proposed. In addition, a scaling methodology has been developed, which consists of a core scaling and a loop scaling. The consistency between the core and loop scaling is examined via the reference volume ratio, which can be obtained from both the core and loop scaling processes. The scaling methodology and similarity laws have been applied to obtain a scientific design of a scaled-down high-temperature DRACS test facility.

  15. Increasing the reliability of the shutdown of 500 - 750-kV overhead lines equipped with shunt reactors in an unsuccessful three-phase automatic repeated closure cycle

    SciTech Connect

    Kuz'micheva, K. I.; Merzlyakov, A. S.; Fokin, G. G.

    2013-05-15

    The reasons for circuit-breaker failures during repeated disconnection of 500 - 750 kV overhead lines with shunt reactors in a cycle of unsuccessful three-phase automatic reconnection (TARC) are analyzed. Recommendations are made for increasing the operating reliability of power transmission lines with shunt reactors when there is unsuccessful reconnection.

  16. Core design of long life-cycle fast reactors operating without reactivity margin

    SciTech Connect

    Aristova, E. N.; Baydin, D. F.; Gol'din, V. Y.; Pestryakova, G. A.; Stoynov, M. I.

    2012-07-01

    In this paper we consider a possibility of designing a fast reactor core that operates without reactivity margin for a long time. This study is based on the physical principle of fast reactor operating in a self-adjustable neutron-nuclear regime (SANNR-1) introduced by L.P. Feoktistov (1988-1993) and improved by V. Ya. Gol'din SANNR-2 (1995). The mathematical modeling of active zones of fast reactors in SANNR modes is held by authors since 1992. The numerical simulation is based on solving the neutron transport equation coupled with quasi-diffusion equations. The calculations have been performed using standard 26 energy groups. We use a hierarchy of spatial models of 1D, 1.5D, 2D, and 3D geometries. The spatial models of higher dimensionality are used for verification of results. The calculations showed that operation of the reactor in this mode increases its efficiency, safety and simplifies management. It is possible to achieve continuous work of the reactor in SANNR-2 during 7-10 years without fuel overloads by means of further optimization of the mode. Small reactivity margin is used only for the reactor start up. After first 10-15 days the reactor in SANNR-2 operates without reactivity margin. (authors)

  17. The 22-Year Hale Cycle in Cosmic Ray Flux - Evidence for Direct Heliospheric Modulation

    NASA Astrophysics Data System (ADS)

    Thomas, S. R.; Owens, M. J.; Lockwood, M.

    2014-01-01

    The ability to predict times of greater galactic cosmic ray (GCR) fluxes is important for reducing the hazards caused by these particles to satellite communications, aviation, or astronauts. The 11-year solar-cycle variation in cosmic rays is highly correlated with the strength of the heliospheric magnetic field. Differences in GCR flux during alternate solar cycles yield a 22-year cycle, known as the Hale Cycle, which is thought to be due to different particle drift patterns when the northern solar pole has predominantly positive (denoted as qA>0 cycle) or negative ( qA<0) polarities. This results in the onset of the peak cosmic-ray flux at Earth occurring earlier during qA>0 cycles than for qA<0 cycles, which in turn causes the peak to be more dome-shaped for qA>0 and more sharply peaked for qA<0. In this study, we demonstrate that properties of the large-scale heliospheric magnetic field are different during the declining phase of the qA<0 and qA>0 solar cycles, when the difference in GCR flux is most apparent. This suggests that particle drifts may not be the sole mechanism responsible for the Hale Cycle in GCR flux at Earth. However, we also demonstrate that these polarity-dependent heliospheric differences are evident during the space-age but are much less clear in earlier data: using geomagnetic reconstructions, we show that for the period of 1905 - 1965, alternate polarities do not give as significant a difference during the declining phase of the solar cycle. Thus we suggest that the 22-year cycle in cosmic-ray flux is at least partly the result of direct modulation by the heliospheric magnetic field and that this effect may be primarily limited to the grand solar maximum of the space-age.

  18. Effect of bed characters on the direct synthesis of dimethyldichlorosilane in fluidized bed reactor.

    PubMed

    Zhang, Pan; Duan, Ji H; Chen, Guang H; Wang, Wei W

    2015-01-01

    This paper presents the numerical investigation of the effects of the general bed characteristics such as superficial gas velocities, bed temperature, bed heights and particle size, on the direct synthesis in a 3D fluidized bed reactor. A 3D model for the gas flow, heat transfer, and mass transfer was coupled to the direct synthesis reaction mechanism verified in the literature. The model was verified by comparing the simulated reaction rate and dimethyldichlorosilane (M2) selectivity with the experimental data in the open literature and real production data. Computed results indicate that superficial gas velocities, bed temperature, bed heights, and particle size have vital effect on the reaction rates and/or M2 selectivity.

  19. Effect of bed characters on the direct synthesis of dimethyldichlorosilane in fluidized bed reactor.

    PubMed

    Zhang, Pan; Duan, Ji H; Chen, Guang H; Wang, Wei W

    2015-01-01

    This paper presents the numerical investigation of the effects of the general bed characteristics such as superficial gas velocities, bed temperature, bed heights and particle size, on the direct synthesis in a 3D fluidized bed reactor. A 3D model for the gas flow, heat transfer, and mass transfer was coupled to the direct synthesis reaction mechanism verified in the literature. The model was verified by comparing the simulated reaction rate and dimethyldichlorosilane (M2) selectivity with the experimental data in the open literature and real production data. Computed results indicate that superficial gas velocities, bed temperature, bed heights, and particle size have vital effect on the reaction rates and/or M2 selectivity. PMID:25742729

  20. Effect of Bed Characters on the Direct Synthesis of Dimethyldichlorosilane in Fluidized Bed Reactor

    PubMed Central

    Zhang, Pan; Duan, Ji H.; Chen, Guang H.; Wang, Wei W.

    2015-01-01

    This paper presents the numerical investigation of the effects of the general bed characteristics such as superficial gas velocities, bed temperature, bed heights and particle size, on the direct synthesis in a 3D fluidized bed reactor. A 3D model for the gas flow, heat transfer, and mass transfer was coupled to the direct synthesis reaction mechanism verified in the literature. The model was verified by comparing the simulated reaction rate and dimethyldichlorosilane (M2) selectivity with the experimental data in the open literature and real production data. Computed results indicate that superficial gas velocities, bed temperature, bed heights, and particle size have vital effect on the reaction rates and/or M2 selectivity. PMID:25742729

  1. Fuel cycles and envisioned roles of fast neutron reactors and hybrids

    NASA Astrophysics Data System (ADS)

    Salvatores, Massimo

    2012-06-01

    Future innovative nuclear fuel cycles will require insuring sustainability in terms of safe operation, optimal use of resources, radioactive waste minimization and reduced risk of proliferation. The present paper introduces some basic notions and fundamental fuel cycle strategies. The simulation approach needed to evaluate the impact of the different fuel cycle alternatives will also be shortly discussed.

  2. Fuel cycles and envisioned roles of fast neutron reactors and hybrids

    SciTech Connect

    Salvatores, Massimo

    2012-06-19

    Future innovative nuclear fuel cycles will require insuring sustainability in terms of safe operation, optimal use of resources, radioactive waste minimization and reduced risk of proliferation. The present paper introduces some basic notions and fundamental fuel cycle strategies. The simulation approach needed to evaluate the impact of the different fuel cycle alternatives will also be shortly discussed.

  3. Record of Cycling Operation of the Natural Nuclear Reactor in the Oklo/Okelobondo Area in Gabon

    NASA Astrophysics Data System (ADS)

    Meshik, A. P.; Hohenberg, C. M.; Pravdivtseva, O. V.

    2004-10-01

    Using selective laser extraction technique combined with sensitive ion-counting mass spectrometry, we have analyzed the isotopic structure of fission noble gases in U-free La-Ce-Sr-Ca aluminous hydroxy phosphate associated with the 2 billion yr old Oklo natural nuclear reactor. In addition to elevated abundances of fission-produced Zr, Ce, and Sr, we discovered high (up to 0.03 cm3 STP/g) concentrations of fission Xe and Kr, the largest ever observed in any natural material. The specific isotopic structure of xenon in this mineral defines a cycling operation for the reactor with 30-min active pulses separated by 2.5h dormant periods. Thus, nature not only created conditions for self-sustained nuclear chain reactions, but also provided clues on how to retain nuclear wastes, including fission Xe and Kr, and prevent uncontrolled runaway chain reaction.

  4. Record of cycling operation of the natural nuclear reactor in the Oklo/Okelobondo area in Gabon.

    PubMed

    Meshik, A P; Hohenberg, C M; Pravdivtseva, O V

    2004-10-29

    Using selective laser extraction technique combined with sensitive ion-counting mass spectrometry, we have analyzed the isotopic structure of fission noble gases in U-free La-Ce-Sr-Ca aluminous hydroxy phosphate associated with the 2 billion yr old Oklo natural nuclear reactor. In addition to elevated abundances of fission-produced Zr, Ce, and Sr, we discovered high (up to 0.03 cm(3) STP/g) concentrations of fission Xe and Kr, the largest ever observed in any natural material. The specific isotopic structure of xenon in this mineral defines a cycling operation for the reactor with 30-min active pulses separated by 2.5 h dormant periods. Thus, nature not only created conditions for self-sustained nuclear chain reactions, but also provided clues on how to retain nuclear wastes, including fission Xe and Kr, and prevent uncontrolled runaway chain reaction.

  5. Extension of the supercritical carbon dioxide brayton cycle to low reactor power operation: investigations using the coupled anl plant dynamics code-SAS4A/SASSYS-1 liquid metal reactor code system.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2012-05-10

    Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle enabling removal of power from an autonomous load following Sodium-Cooled Fast Reactor (SFR) down to decay heat levels such that the S-CO{sub 2} cycle can be used to cool the reactor until decay heat can be removed by the normal shutdown heat removal system or a passive decay heat removal system such as Direct Reactor Auxiliary Cooling System (DRACS) loops with DRACS in-vessel heat exchangers. This capability of the new control strategy eliminates the need for use of a separate shutdown heat removal system which might also use supercritical CO{sub 2}. It has been found that this capability can be achieved by introducing a new control mechanism involving shaft speed control for the common shaft joining the turbine and two compressors following reduction of the load demand from the electrical grid to zero. Following disconnection of the generator from the electrical grid, heat is removed from the intermediate sodium circuit through the sodium-to-CO{sub 2} heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO{sub 2}-to-water cooler. To investigate the effectiveness of shaft speed control, calculations are carried out using the coupled Plant Dynamics Code-SAS4A/SASSYS-1 code for a linear load reduction transient for a 1000 MWt metallic-fueled SFR with autonomous load following. No deliberate motion of control rods or adjustment of sodium pump speeds is assumed to take place. It is assumed that the S-CO{sub 2} turbomachinery shaft speed linearly decreases from 100 to 20% nominal following reduction of grid load to zero. The reactor power is calculated to autonomously decrease down to 3% nominal providing a lengthy window in time for the switchover to the normal shutdown heat removal system or for a passive decay heat removal system to become effective. However, the

  6. Electric power generating plant having direct coupled steam and compressed air cycles

    DOEpatents

    Drost, Monte K.

    1982-01-01

    An electric power generating plant is provided with a Compressed Air Energy Storage (CAES) system which is directly coupled to the steam cycle of the generating plant. The CAES system is charged by the steam boiler during off peak hours, and drives a separate generator during peak load hours. The steam boiler load is thereby levelized throughout an operating day.

  7. Electric power generating plant having direct-coupled steam and compressed-air cycles

    DOEpatents

    Drost, M.K.

    1981-01-07

    An electric power generating plant is provided with a Compressed Air Energy Storage (CAES) system which is directly coupled to the steam cycle of the generating plant. The CAES system is charged by the steam boiler during off peak hours, and drives a separate generator during peak load hours. The steam boiler load is thereby levelized throughout an operating day.

  8. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    SciTech Connect

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

    2012-04-01

    The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water

  9. Scaling Analysis for the Direct Reactor Auxillary Cooling System For AHTRS

    SciTech Connect

    Lv, Q. NMN; Wang, X. NMN; Sun, X NMN; Christensen, R. N.; Blue, T. E.; Yoder Jr, Graydon L; Wilson, Dane F; Subharwall, Piyush; Adams, I.

    2013-01-01

    The Direct Reactor Auxiliary Cooling System (DRACS) is a passive heat removal system proposed for the Advanced High-Temperature Reactor (AHTR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops relying completely on buoyancy as the driving force. In the DRACS, two heat exchangers, namely, the DRACS Heat Exchanger (DHX) and the Natural Draft Heat Exchanger (NDHX) are used to couple these loops. In addition, a fluidic diode is employed to minimize the parasitic flow during normal operation of the reactor and to activate the DRACS in accidents. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for AHTRs built and tested in the literature. In this paper, a detailed scaling analysis for the DRACS is performed, which will provide guidance for the design of the scaled-down DRACS test facilities. Based on the Boussinesq assumption and one-dimensional flow formulation, the governing equations are non-dimensionalized by introducing appropriate dimensionless parameters. The key dimensionless numbers that characterize the DRACS system are obtained straightforwardly from the non-dimensional governing equations. Based on the dimensionless numbers and non-dimensional governing equations, similarity laws are proposed. In addition, a scaling methodology has also been developed, which consists of the core scaling and loop scaling. The consistence between the core and loop scaling is examined through the reference volume ratio, which can be obtained from the core and loop scaling processes. The scaling methodology and similarity laws have been applied to obtain a design of the scaled-down high-temperature DRACS test facility (HTDF).

  10. SHORT CONTACT TIME DIRECT COAL LIQUEFACTION USING A NOVEL BATCH REACTOR

    SciTech Connect

    Michael T. Klein; William H. Calkins

    1997-10-29

    The overall goal of this research is to develop an understanding of the Direct Coal Liquefaction process at the molecular level. Many approaches have been used to study this process including kinetic studies, study of the liquefaction products, study of the effect of reaction variables, such as temperature, solvent type and composition, the changing nature and composition of the coal during liquefaction, and the distribution in the liquefaction products of the hydrogen consumed. While all these studies have contributed to our growing knowledge of the liquefaction process, an adequate understanding of direct liquefaction still eludes us. This is due to many reasons including: the complexity and variable nature of coal itself and the many different chemical reactions which are occurring simultaneously during direct coal liquefaction. We believe that a study of the liquefaction process at the very early stages will avoid the complexities of secondary reactions associated with free radical high temperature processes that are clearly involved in direct coal liquefaction. This prompted us to devise a reactor system which avoids long heat up and cool-down times associated with previous kinetic studies, and allows kinetic measurements even at as short as the first few seconds of the liquefaction reaction.

  11. Short Contact Time Direct Coal Liquefaction Using a Novel Batch Reactor

    SciTech Connect

    He Huang; Michael T. Klein; William H. Calkins

    1997-04-03

    The primary objective of this research is to optimize the design and operation of the bench scale batch reactor (SCTBR) for studying direct coal liquefaction at short contact times (.01 to 10 minutes or longer) . An additional objective is to study the kinetics of direct coal liquefaction particularly at short reaction times. Both of these objectives have been nearly achieved, however this work has shown the great importance of the liquefaction solvent characteristics and the solvent-catalyst interaction on the liquefaction process. This has prompted us to do a preliminary investigation of solvents and the solvent-catalyst systems in coal liquefaction. SUMMARY AND CONCLUSIONS 1) Conversion vs time data have been extended to 5 coals of ranks from lignite to low volatile bituminous coal. A broad range of reaction rates have been observed with a maximum in the high volatile bituminous range. 2) A series of direct coal liquefaction runs have been made using a range of nitrogen containing solvents that given high liquefaction conversions of coal. These runs are now being analyzed. 3) The coalification process has been shown by TGA to go through an intermediate stage which may account for the greater reactivity of bituminous coals in the direct coal liquefaction process. 4) It was shown that coal rank can be accurately determined by thermogravimetric analysis

  12. Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

    SciTech Connect

    Bruce G. Schnitzler

    2012-01-01

    well as open loop systems for direct nuclear thermal propulsion. Although a number of fast spectrum reactor and engine designs suitable for direct nuclear thermal propulsion were proposed and designed, none were built. This report summarizes status results of evaluations of small nuclear reactor designs suitable for direct nuclear thermal propulsion.

  13. Safeguards and Non-proliferation Issues as Related to Advanced Fuel Cycle and Advanced Fast Reactor Development with Processing of Reactor Fuel

    SciTech Connect

    Rahmat Aryaeinejad; Jerry D. Cole; Mark W. Drigert; Dee E. Vaden

    2006-10-01

    The goal of this work is to establish basic data and techniques to enable safeguards appropriate to a new generation of nuclear power systems that will be based on fast spectrum reactors and mixed actinide fuels containing significant quantities of "minor" actinides, possibly due to reprocessing, and determination of what new radiation signatures and parameters need to be considered. The research effort focuses on several problems associated with the use of fuel having significantly different actinide inventories that current practice and on the development of innovative techniques using new radiation signatures and other parameters useful for safeguards and monitoring. In addition, the development of new distinctive radiation signatures as an aid in controlling proliferation of nuclear materials has parallel applications to support Gen-IV and current advanced fuel cycle initiative (AFCI) goals as well as the anticipated Global Nuclear Energy Partnership (GNEP).

  14. Surveillance strategy for an extended operating cycle in commercial nuclear reactors

    SciTech Connect

    McHenry, R.S.; Moore, T.J.; Maurer, J.H.; Todreas, N.E.

    1997-05-01

    The impetus for improved economic performance of commercial nuclear power plants can be partially satisfied by increasing plant capacity factors through operating cycle extension. One aspect of an operating cycle extension effort is the modification of plant surveillance programs to complete required regulatory and investment protection surveillance activities within the extended planned outage schedule. The goal of this paper is to introduce a general strategy for existing power plants to transition their surveillance programs to an extended operating cycle up to 48 months in length, and to test the feasibility of this strategy through the complete analysis of the surveillance programs at operating BWR and PWR case study plants. The reconciliation of surveillances at these plants demonstrates that surveillance performance will not preclude 48 month operating cycles. Those surveillance activities that could not be resolved to an extended cycle are identified for further study. Finally, a number of general issues are presented that should be considered before implementing a cycle extension effort.

  15. PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? EXTENDING CYCLE BURNUP

    SciTech Connect

    Primm, Trent; Chandler, David

    2009-01-01

    Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting HFIR from high enriched to low enriched uranium (20 wt % 235U) fuel requires extending the end-of-life burnup value for HFIR fuel from the current nominal value of 2200 MWD to 2600 MWD. The current fuel fabrication procedure is discussed and changes that would be required to this procedure are identified. Design and safety related analyses that are required for the certification of a new fuel are identified. Qualification tests and comments regarding the regulatory approval process are provided along with a conceptual schedule.

  16. Short contact time direct coal liquefaction using a novel batch reactor. Quarterly report, 1996

    SciTech Connect

    Klein, M.T.; Calkins, W.H.; Huang, H.

    1996-05-01

    The objective of this research is to optimize the design and operation of the bench scale batch reactor (SCTBR) for coal liquefaction at short contact times (0.01 to 10 minutes or longer). Additional objectives are to study the kinetics of direct coal liquefaction particularly at short reaction times, and to investigate the role of the organic oxygen components of coal and their reaction pathways during liquefaction. Many of those objectives have already been achieved and others are still in progress. This quarterly report covers further progress toward those objectives. Much of the previous quarterly report was concerned mainly in the retrograde reactions occurring during the liquefaction process. This report is largely devoted to the kinetics and mechanisms of the liquefaction process itself and the influence of the liquefaction solvents.

  17. Preparation of Biodiesel from Microalgae and Palm Oil by Direct Transesterification in a Batch Microwave Reactor

    NASA Astrophysics Data System (ADS)

    Marwan; Suhendrayatna; Indarti, E.

    2015-06-01

    The present work was aimed to study the so-called direct transesterification of microalgae lipids to biodiesel in a batch microwave reactor. As a comparison, preparation of palm oil to biodiesel by alkaline catalyzed ethanolysis was also carried out. Palm oil biodiesel was recovered close to an equilibrium conversion (94-96% yield) under microwave heating for at least 6 min, while the conventional method required more than 45 minutes reaching the same yield. A very short reaction time suggests the benefit of microwave effect over conventional heating method in making biodiesel. FTIR analysis revealed the presence of fatty acid ethyl esters with no undesired chemical groups or compounds formed due to local heat generated by microwave effect, thus the conversion only followed transesterification route. Oil containing microalgae of Chlorella sp. isolated from the local brackish water pond was used as a potential source of biodiesel. High yield of biodiesel (above 0.6 g/g of dried algae) was also attainable for the direct transesterification of microalgae in the microwave reactor. Effect of water content of the algae biomass became insignificant at 11.9%(w/w) or less, related to the algae biomass dried for longer than 6 h. Fast transesterification of the algal oil towards equilibrium conversion was obtained at reaction time of 6 min, and at longer times the biodiesel yield remains unchanged. FAME profile indicates unsaturated fatty acids as major constituents. It was shown that microwave irradiation contributes not only to enhance the transeseterification, but also to assist effective release of fatty acid containing molecules (e.g. triacylglycerol, free fatty acids and phospholipids) from algal cells.

  18. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 4-Three Mile Island Unit 1 Cycle 5

    SciTech Connect

    DeHart, M.D.

    1995-01-01

    The requirements of ANSI/ANS-8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using relevant and well-documented critical configurations from commercial pressurized water reactors. The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Isotopic densities for spent fuel assemblies in the core were calculated using the SCALE-4 SAS2H analytical sequence. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code family was used to extract the necessary isotopic densities from SAS2H results and to provide the data in the format required for SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) for the critical configuration. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all calculations. This volume of the report documents a reactor critical calculation for GPU Nuclear Corporation's Three Mile Island Unit 1 (TMI-1) during hot, zero-power startup testing for the beginning of cycle 5. This unit and cycle were selected because of their relevance in spent fuel benchmark applications: (1) cycle 5 startup occurred after an especially long downtime of 6.6 years; and (2) the core consisted primarily (75%) of burned fuel, with

  19. 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.

  20. Stride-Cycle Influences on Goal-Directed Head Movements Made During Walking

    NASA Technical Reports Server (NTRS)

    Peters, Brian T.; vanEmmerik, Richard E. A.; Bloomberg, Jacob J.

    2006-01-01

    Horizontal head movements were studied in six subjects as they made rapid horizontal gaze adjustments while walking. The aim of the present research was to determine if gait-cycle events alter the head movement response to a visual target acquisition task. Gaze shifts of approximately 40deg were elicited by a step change in the position of a visual target from a central location to a second location in the left or right horizontal periphery. The timing of the target position change was constrained to occur at 25,50,75 and 100% of the stride cycle. The trials were randomly presented as the subjects walked on a treadmill at their preferred speed (range: 1.25 to 1.48 m/s, mean: 1.39 +/- 0.09 m/s ) . Analyses focused on the movement onset latencies of the head and eyes and on the peak velocity and saccade amplitude of the head movement response. A comparison of the group means indicated that the head movement onset lagged the eye onset (262 ms versus 252 ms). The head and eye movement onset latencies were not affected by either the direction of the target change nor the point in the gait cycle during which the target relocation occurred. However, the presence of an interaction between the gait cycle events and the direction of the visual target shift indicates that the peak head saccade velocity and head saccade amplitude are affected by the natural head oscillations that occur while walking.

  1. Direct Energy Conversion Fission Reactor, Gaseous Core Reactor with Magnetohydrodynamic (MHD) Generator; Final Report - Part I and Part II

    SciTech Connect

    Samim Anghaie; Blair Smith; Travis Knight

    2002-11-12

    This report focuses on the power conversion cycle and efficiency. The technical issues involving the ionization mechanisms, the power management and distribution and radiation shielding and safety will be discussed in future reports.

  2. Effects of cycle-frequency and temperature on the performance of anaerobic sequencing batch reactors (ASBRs) treating swine waste.

    PubMed

    Ndegwa, P M; Hamilton, D W; Lalman, J A; Cumba, H J

    2008-04-01

    Anaerobic digestion of animal waste is a technically viable process for the abatement of adverse environmental impacts caused by animal wastes; however, widespread acceptance has been plagued by poor economics. This situation is dismal if the technology is adapted for treating low strength animal slurries because of large digester-volume requirements and a corresponding high energy input. A possible technology to address these constraints is the anaerobic sequencing batch reactor (ASBR). The ASBR technology has demonstrated remarkable potential to improve the economics of treating dilute animal waste effluents. This paper presents preliminary data on the effects of temperature and frequency-cycle on the operation of an ASBR at a fixed hydraulic retention time (HRT). The results suggest that within the parameter range under consideration, temperature did not affect the biogas yield significantly, however, higher cycle-frequency had a negative effect. The biogas quality (%CH(4)) was not significantly affected by temperature nor by the cycle-frequency. The operating principle of the ASBR follows four phases: feed, react, settle, and decant in a cyclic mode. To improve the biogas production in an ASBR, one long react-phase was preferable compared to three shorter react-phases. Treatment of dilute manure slurries in an ASBR at 20 degrees C was more effective than at 35 degrees C; similarly more bio-stable effluents were obtained at low cycle-frequency. The treatment of dilute swine slurries in an ASBR at the lower temperature (20 degrees C) and lower cycle-frequency is, therefore, recommended for the bio-stabilization of dilute swine wastewaters. The results also indicate that significantly higher VFA degradation occurred at 20 degrees C than at 35 degrees C, suggesting that the treatment of dilute swine slurries in ASBRs for odor control might be more favorable at the lower than at the higher temperatures examined in this study. Volatile fatty acid reduction at the two

  3. Space reactor/Stirling cycle systems for high power Lunar applications

    SciTech Connect

    Schmitz, P.D.; Mason, L.S.

    1994-09-01

    NASA`s Space Exploration Initiative (SEI) has proposed the use of high power nuclear power systems on the lunar surface as a necessary alternative to solar power. Because of the long lunar night ({approximately} 14 earth days) solar powered systems with the requisite energy storage in the form of regenerative fuel cells or batteries becomes prohibitively heavy at high power levels ({approximately} 100 kWe). At these high power levels nuclear power systems become an enabling technology for variety of missions. One way of producing power on the lunar surface is with an SP-100 class reactor coupled with Stirling power converters. In this study, analysis and characterization of the SP-100 class reactor coupled with Free Piston Stirling Power Conversion (FPSPC) system will be performed. Comparison of results with previous studies of other systems, particularly Brayton and Thermionic, are made.

  4. Space reactor/Stirling cycle systems for high power lunar application

    NASA Technical Reports Server (NTRS)

    Schmitz, Paul C.; Mason, Lee S.

    1991-01-01

    An analysis is performed to mathematically model a 550 kWe lunar base power supply which uses a SP-100 reactor coupled with Stirling converters. The reactor is placed in an excavation to keep activated coolant in the hole and to allow maintenance of the components outside the hole. Two technology levels are considered. They are 1050 and 1300 K heater head Stirling converts. It is found that for a 1050 K converter the total mass which provided 1000 volts DC at 250 m is 14,366 kg while the 1300 K system mass is 12,104 kg. The radiation area of the 1050 and 1300 K systems are 641 and 356 sq m respectively. Comparisons are made with Brayton and thermionic systems with both near term and advanced technology considered.

  5. Space reactor/Stirling cycle systems for high power lunar applications

    NASA Technical Reports Server (NTRS)

    Schmitz, Paul C.; Mason, Lee S.

    1991-01-01

    An analysis is performed to mathematically model a 550 kWe lunar base power supply which uses a SP-100 reactor coupled with Stirling converters. The reactor is placed in an excavation to keep activated coolant in the hole and to allow maintance of the components outside the hole. Two technology levels are considered. They are 1050 and 1300 K heater head Stirling converts. It is found that for a 1050 K converter the total mass which provided 1000 volts dc at 250 m is 14,366 kg while the 1300 K system mass is 12,104 kg. The radiation area of the 1050 and 1300 K systems are 641 and 356 sq m respectively. Comparisons are made with Brayton and thermionic systems with both near term and advanced technology considered.

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

    NASA Technical Reports Server (NTRS)

    Whitmarsh, C. L., Jr.

    1972-01-01

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

  7. Direct positive selection for improved nitroreductase variants using SOS triggering of bacteriophage lambda lytic cycle.

    PubMed

    Guise, C P; Grove, J I; Hyde, E I; Searle, P F

    2007-04-01

    Expression of prodrug-activating enzymes that convert non-toxic substrates to cytotoxic derivatives is a promising strategy for cancer gene therapy. However, their catalytic activity with unnatural, prodrug substrates is often suboptimal. Efforts to improve these enzymes have been limited by the inability to select directly for increased prodrug activation. We have focussed on developing variants of Escherichia coli (E. coli) nitroreductase (NTR) with improved ability to activate the prodrug 5-(aziridin-1-yl)-2,4-dinitrobenzamide (CB1954), and describe here a novel, direct, positive selection for improved enzymes that exploits the alternative life cycles of bacteriophage lambda. In lambda lysogens of E. coli, the activation of the prodrug CB1954 by NTR triggers the SOS response to DNA damage, switching integrated lambda prophages into lytic cycle. This provides a direct, positive selection for phages encoding improved NTR variants, as, upon limiting exposure of lysogenized E. coli to CB1954, only those encoding the most active enzyme variants are triggered into lytic cycle, allowing their selective recovery. We exemplify the selection by isolating highly improved 'turbo-NTR' variants from a library of 6.8 x 10(5) clones, conferring up to 50-fold greater sensitivity to CB1954 than the wild type. Carcinoma cells infected with adenovirus expressing T41Q/N71S/F124T-NTR were sensitized to CB1954 concentrations 40- to 80-fold lower than required with WT-NTR. PMID:17301844

  8. Materials considerations for the coupling of thermochemical hydrogen cycles to tandem mirror reactors

    SciTech Connect

    Krikorian, O.H.

    1980-10-10

    Candidate materials are discussed and initial choices made for the critical elements in a liquid Li-Na Cauldron Tandem Mirror blanket and the General Atomic Sulfur-Iodine Cycle for thermochemical hydrogen production. V and Ti alloys provide low neutron activation, good radiation damage resistance, and good chemical compatibility for the Cauldron design. Aluminide coated In-800H and siliconized SiC are materials choices for heat exchanger components in the thermochemical cycle interface.

  9. Biological phosphorus and nitrogen removal in sequencing batch reactors: effects of cycle length, dissolved oxygen concentration and influent particulate matter.

    PubMed

    Ginige, Maneesha P; Kayaalp, Ahmet S; Cheng, Ka Yu; Wylie, Jason; Kaksonen, Anna H

    2013-01-01

    Removal of phosphorus (P) and nitrogen (N) from municipal wastewaters is required to mitigate eutrophication of receiving water bodies. While most treatment plants achieve good N removal using influent carbon (C), the use of influent C to facilitate enhanced biological phosphorus removal (EBPR) is poorly explored. A number of operational parameters can facilitate optimum use of influent C and this study investigated the effects of cycle length, dissolved oxygen (DO) concentration during aerobic period and influent solids on biological P and N removal in sequencing batch reactors (SRBs) using municipal wastewaters. Increasing cycle length from 3 to 6 h increased P removal efficiency, which was attributed to larger portion of N being removed via nitrite pathway and more biodegradable organic C becoming available for EBPR. Further increasing cycle length from 6 to 8 h decreased P removal efficiencies as the demand for biodegradable organic C for denitrification increased as a result of complete nitrification. Decreasing DO concentration in the aerobic period from 2 to 0.8 mg L(-1) increased P removal efficiency but decreased nitrification rates possibly due to oxygen limitation. Further, sedimented wastewater was proved to be a better influent stream than non-sedimented wastewater possibility due to the detrimental effect of particulate matter on biological nutrient removal.

  10. Preliminary reactor cavity melt dispersal model for direct containment heating scenarios

    SciTech Connect

    Ginsberg, T.; Tutu, N.K.

    1989-01-01

    This paper presents the results of a series of experiments performed to study the effect of initial pressure vessel conditions on the extent of melt dispersal from scaled reactor cavities and describes progress in development of a mathematical model which is designed to predict the melt mass dispersed from reactor cavities as a function of reactor vessel initial conditions and on the vessel breach area. The model, which is being developed to also characterize the heat transfer and chemical reaction phenomena which would take place within the reactor cavity, is designed to be incorporated into a lumped-parameter containment analysis computer code.

  11. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle description

    SciTech Connect

    Not Available

    1980-06-01

    The Nonproliferation Alterntive Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates.

  12. Cycle-Based Cluster Variational Method for Direct and Inverse Inference

    NASA Astrophysics Data System (ADS)

    Furtlehner, Cyril; Decelle, Aurélien

    2016-08-01

    Large scale inference problems of practical interest can often be addressed with help of Markov random fields. This requires to solve in principle two related problems: the first one is to find offline the parameters of the MRF from empirical data (inverse problem); the second one (direct problem) is to set up the inference algorithm to make it as precise, robust and efficient as possible. In this work we address both the direct and inverse problem with mean-field methods of statistical physics, going beyond the Bethe approximation and associated belief propagation algorithm. We elaborate on the idea that loop corrections to belief propagation can be dealt with in a systematic way on pairwise Markov random fields, by using the elements of a cycle basis to define regions in a generalized belief propagation setting. For the direct problem, the region graph is specified in such a way as to avoid feed-back loops as much as possible by selecting a minimal cycle basis. Following this line we are led to propose a two-level algorithm, where a belief propagation algorithm is run alternatively at the level of each cycle and at the inter-region level. Next we observe that the inverse problem can be addressed region by region independently, with one small inverse problem per region to be solved. It turns out that each elementary inverse problem on the loop geometry can be solved efficiently. In particular in the random Ising context we propose two complementary methods based respectively on fixed point equations and on a one-parameter log likelihood function minimization. Numerical experiments confirm the effectiveness of this approach both for the direct and inverse MRF inference. Heterogeneous problems of size up to 10^5 are addressed in a reasonable computational time, notably with better convergence properties than ordinary belief propagation.

  13. 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.

  14. Variation in the statistical properties of IMF direction fluctuations during the 22-year solar magnetic cycle

    NASA Astrophysics Data System (ADS)

    Erofeev, D. V.

    2014-12-01

    The variation in the IMF direction distribution during the 22-year solar magnetic cycle has been studied. Data obtained in near-Earth orbits and measurements in the heliospheric regions located far from the Earth, performed with the Helios and Ulysses spacecraft devices, have been analyzed. It has been found that the correlation between the azimuth and magnetic field fluctuations is statistically significant in the low-latitude heliospheric region at heliocentric distances of 0.3-5.4 AU, and the sign of this correlation reverses at a change in the polar solar magnetic field orientation. In the polar zones of the heliosphere outside the latitudinal extension of the heliospheric current sheet, the angle correlation coefficient rapidly decreases with increasing heliographic latitude. The angle correlation sign reversal during the 22-year cycle is accompanied by a change of the asymmetry sign of the magnetic field inclination distribution.

  15. Reactor design considerations in the hot filament/direct current plasma synthesis of carbon nanofibers

    NASA Astrophysics Data System (ADS)

    Cruden, Brett A.; Cassell, Alan M.; Ye, Qi; Meyyappan, M.

    2003-09-01

    A combined hot filament/direct current (dc) plasma approach to chemical vapor deposition of carbon nanofibers (CNFs) using an acetylene/ammonia feedstock has been explored. As a part of the study, the impact of filament usage and substrate holder design has been examined by scanning electron microscopy imaging of deposition products and monitoring of downstream products by residual gas analysis (RGA). It is demonstrated that the filament wire is important primarily in the pretreatment of the substrate, improving CNF growth quality. However, the filament has a more minor impact when combined with the dc plasma, increasing growth rate but reducing growth quality. The substrate holder is modified by introducing a graphite spacer into the electrode. By varying the size of the spacer, the effective surface area of the cathode is modified, allowing control over the power input to the reactor while holding the voltage constant. This allows for some independent control of physicochemical processes that are typically inseparable in plasma processing, including gas phase chemistry, substrate heating and etching by ion bombardment, and growth alignment effects due to the electric field. This work demonstrates how separating these processes allows for better control over the desired growth product.

  16. Direct Characterization of Methanogens in Two High-Rate Anaerobic Biological Reactors

    PubMed Central

    Kobayashi, Hester A.; de Macario, Everly Conway; Williams, Regan S.; Macario, Alberto J. L.

    1988-01-01

    The methanogenic flora from two types of turbulent, high-rate reactors was studied by immunologic methods as well as by phase-contrast, fluorescence, and scanning electron microscopy. The reactors were a fluidized sand-bed biofilm ANITRON reactor and an ultrafiltration membrane-associated suspended growth MARS reactor (both trademarks of Air Products and Chemicals, Inc., Allentown, Pa.). Conventional microscopic methods revealed complex mixtures of microbes of a range of sizes and shapes, among which morphotypes resembling Methanothrix spp. and Methanosarcina spp. were noticed. Precise identification of these and other methanogens was accomplished by antigenic fingerprinting with a comprehensive panel of calibrated antibody probes of predefined specificity spectra. The methanogens identified showed morphotypes and antigenic fingerprints indicating their close similarity with the following reference organisms: Methanobacterium formicicum MF and Methanosarcina barkeri W in the ANITRON reactor only; Methanosarcina barkeri R1M3, M. mazei S6, Methanogenium cariaci JR1, and Methanobrevibacter arboriphilus AZ in the MARS reactor only; and Methanobrevibacter smithii ALI and Methanothrix soehngenii Opfikon in both reactors. Species diversity and distribution appeared to be, at least in part, dependent on the degree of turbulence inside the reactor. Images PMID:16347581

  17. A description of the demonstration Integral Fast Reactor fuel cycle facility.

    PubMed

    Courtney, J C; Carnes, M D; Dwight, C C; Forrester, R J

    1991-10-01

    A fuel examination facility at the Idaho National Engineering Laboratory is being converted into a facility that will electrochemically process spent fuel. This is an important step in the demonstration of the Integral Fast Reactor concept being developed by Argonne National Laboratory. Renovations are designed to bring the facility up to current health and safety and environmental standards and to support its new mission. Improvements include the addition of high-reliability earthquake hardened off-gas and electrical power systems, the upgrading of radiological instrumentation, and the incorporation of advances in contamination control. A major task is the construction of a new equipment repair and decontamination facility in the basement of the building to support operations.

  18. Fuel cycle facility control system for the Integral Fast Reactor Program

    SciTech Connect

    Benedict, R.W.; Tate, D.A.

    1993-09-01

    As part of the Integral Fast Reactor (IFR) Fuel Demonstration, a new distributed control system designed, implemented and installed. The Fuel processes are a combination of chemical and machining processes operated remotely. To meet this special requirement, the new control system provides complete sequential logic control motion and positioning control and continuous PID loop control. Also, a centralized computer system provides near-real time nuclear material tracking, product quality control data archiving and a centralized reporting function. The control system was configured to use programmable logic controllers, small logic controllers, personal computers with touch screens, engineering work stations and interconnecting networks. By following a structured software development method the operator interface was standardized. The system has been installed and is presently being tested for operations.

  19. A SAT Based Effective Algorithm for the Directed Hamiltonian Cycle Problem

    NASA Astrophysics Data System (ADS)

    Jäger, Gerold; Zhang, Weixiong

    The Hamiltonian cycle problem (HCP) is an important combinatorial problem with applications in many areas. While thorough theoretical and experimental analyses have been made on the HCP in undirected graphs, little is known for the HCP in directed graphs (DHCP). The contribution of this work is an effective algorithm for the DHCP. Our algorithm explores and exploits the close relationship between the DHCP and the Assignment Problem (AP) and utilizes a technique based on Boolean satisfiability (SAT). By combining effective algorithms for the AP and SAT, our algorithm significantly outperforms previous exact DHCP algorithms including an algorithm based on the award-winning Concorde TSP algorithm.

  20. Sub-Cycle Quantum Optics: Direct Access to Electric Field Vacuum Fluctuations

    NASA Astrophysics Data System (ADS)

    Seletskiy, Denis; Riek, Claudius; Moskalenko, Andrey; Schmidt, Jan; Krauspe, Philipp; Eckart, Sebastian; Eggert, Stefan; Burkard, Guido; Leitenstorfer, Alfred

    Vacuum fluctuations are fundamental to a variety of physical aspects ranging from spontaneous photon emission via the Casimir force all the way to cosmology. Study and manipulation of the ground state of the radiation field is a central subject in quantum optics. In common approaches, such as for example homodyne detection, the information is averaged over multiple cycles of light and amplification to finite intensity is mandatory. Usually, ultrashort pulses are applied for quantum measurements within a slowly-varying envelope approximation. We demonstrate direct detection of the vacuum fluctuations of the local electric field amplitude in free space. Broadband electro-optic sampling with sub-6 femtosecond gate pulses enables quantum-statistic readout. Distinction from the detector shot noise is achieved by modification of the sampled space-time volume. Measuring with a bandwidth matching the 70 THz center frequency maximizes the vacuum amplitude since the ground-state energy approaches half a photon per optical cycle. Our findings open up a new avenue to quantum analysis and manipulation of light working in the time domain and with sub-cycle access to the electric field quadrature.

  1. Direct numerical simulation of multiple cycles in a valve/piston assembly

    NASA Astrophysics Data System (ADS)

    Schmitt, Martin; Frouzakis, Christos E.; Tomboulides, Ananias G.; Wright, Yuri M.; Boulouchos, Konstantinos

    2014-03-01

    The dynamics and multiple-cycle evolution of the incompressible flow induced by a moving piston through the open valve of a motored piston-cylinder assembly was investigated using direct numerical simulation. A spectral element solver, adapted for moving geometries using an Arbitrary Lagrange/Eulerian formulation, was employed. Eight cycles were simulated and the ensemble- and azimuthally-averaged data were found to be in good agreement with experimentally determined means and fluctuations at all measured points and times. During the first half of the intake stroke the flow field is dominated by the dynamics of the incoming jet and the vortex rings it creates. With decreasing piston speed a large central ring becomes the dominant flow feature until the top dead center. The flow field at the end of the previous cycle is found to have a dominant effect on the jet breakup and the vortex ring dynamics below the valve and on the observed significant cyclic variations. Based on statistical averaging, the evolution of the turbulent flow field during the first half of the intake stroke is dominated by the jet breakup process leading to a strongly anisotropic behavior. In the second part of the intake stroke, the decrease of the incoming jet velocity results in a more isotropic behavior.

  2. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    NASA Astrophysics Data System (ADS)

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

    2014-02-01

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is 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 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  3. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    SciTech Connect

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

    2014-02-12

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is 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 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  4. Review of the cost estimate and schedule for the 2240-MWt high-temperature gas-cooled reactor steam-cycle/cogeneration lead plant

    SciTech Connect

    Not Available

    1983-09-01

    This report documents Bechtel's review of the cost estimate and schedule for the 2240 MWt High Temperature Gas-Cooled Reactor Steam Cycle/Cogeneration (HTGR-SC/C) Lead Plant. The overall objective of the review is to verify that the 1982 update of the cost estimate and schedule for the Lead Plant are reasonable and consistent with current power plant experience.

  5. Sensitivity of Advanced Reactor and Fuel Cycle Performance Parameters to Nuclear Data Uncertainties

    NASA Astrophysics Data System (ADS)

    Aliberti, G.; Palmiotti, G.; Salvatores, M.; Kim, T. K.; Taiwo, T. A.; Kodeli, I.; Sartori, E.; Bosq, J. C.; Tommasi, J.

    2006-04-01

    As a contribution to the feasibility assessment of Gen IV and AFCI relevant systems, a sensitivity and uncertainty study has been performed to evaluate the impact of neutron cross section uncertainty on the most significant integral parameters related to the core and fuel cycle. Results of an extensive analysis indicate only a limited number of relevant parameters and do not show any potential major problem due to nuclear data in the assessment of the systems considered. However, the results obtained depend on the uncertainty data used, and it is suggested to focus some future evaluation work on the production of consistent, as far as possible complete and user oriented covariance data.

  6. A multidimensional model of direct-stream heating of newspaper and municipal solid waste in a hydrothermal reactor

    SciTech Connect

    Thorsness, C.B.

    1995-09-28

    Hydrothermal treatment (reaction in a water medium at elevated temperatures) can transform many municipal solid waste (MSW) constituents into a synthetic coal material which is more amenable for use as a fuel or chemical feedstock than the raw MSW. One means of heating the MSW is to use direct high temperature steam injection into a closed reactor and allow the latent heat of the steam to raise the MSW to the desired temperature and at the same time build the pressure necessary to maintain a water phase. This report describes a computer model which can be used to look at details of the steam flow, water evaporation/condensation, thermal evolution, and MSW decomposition in a direct-steam heated MSW hydrothermal reactor. The model treats the system as a packed bed using a Darcy`s law formulation for computing gas flow rates. The model has been applied to a pilot and a commercial scale system. Computations take between 1-6 hours on a HP-9000/730. Initial computations performed with the model indicate that pressure drop and velocities on a pilot scale systems will be small. On the other hand, they indicate that gas velocities inside a commercial scale reactor can reach levels at which entrainment of liquid or solids could occur. In addition, on the commercial scale, model results indicate that in the absence of liquid water flow the thermal coupling between vessel contents and heavy reactor walls should be small thus minimizing unwanted heat loss.

  7. Direct inhibition of Retinoblastoma phosphorylation by Nimbolide causes cell cycle arrest and suppresses glioblastoma growth

    PubMed Central

    Anderson, Jane; Liu, Xiaona; Henry, Heather; Gasilina, Anjelika; Nassar, Nicholas; Ghosh, Jayeeta; Clark, Jason P; Kumar, Ashish; Pauletti, Giovanni M.; Ghosh, Pradip K; Dasgupta, Biplab

    2013-01-01

    Purpose Classical pharmacology allows the use and development of conventional phytomedicine faster and more economically than conventional drugs. This approach should be tested for their efficacy in terms of complementarity and disease control. The purpose of this study was to determine the molecular mechanisms by which nimbolide, a triterpenoid found in the well-known medicinal plant Azadirachta indica controls glioblastoma (GBM) growth. Experimental Design Using in vitro signaling, anchorage-independent growth, kinase assays, and xenograft models, we investigated the mechanisms of its growth inhibition in glioblastoma. Results We show that nimbolide or an ethanol soluble fraction of A. indica leaves (Azt) that contains nimbolide as the principal cytotoxic agent is highly cytotoxic against GBM in vitro and in vivo. Azt caused cell cycle arrest, most prominently at the G1-S stage in GBM cells expressing EGFRvIII, an oncogene present in about 20-25% of GBMs. Azt/nimbolide directly inhibited CDK4/CDK6 kinase activity leading to hypophosphorylation of the retinoblastoma (RB) protein, cell cycle arrest at G1-S and cell death. Independent of RB hypophosphorylation, Azt also significantly reduced proliferative and survival advantage of GBM cells in vitro and in tumor xenografts by downregulating Bcl2 and blocking growth factor induced phosphorylation of Akt, Erk1/2 and STAT3. These effects were specific since Azt did not affect mTOR or other cell cycle regulators. In vivo, Azt completely prevented initiation and inhibited progression of GBM growth. Conclusions Our preclinical findings demonstrate Nimbolide as a potent anti-glioma agent that blocks cell cycle and inhibits glioma growth in vitro and in vivo. PMID:24170547

  8. Thermodynamic Analysis of the Use a Chemical Heat Pump to Link a Supercritical Water-Cooled Nuclear Reactor and a Thermochemical Water-Splitting Cycle for Hydrogen Production

    NASA Astrophysics Data System (ADS)

    Granovskii, Mikhail; Dincer, Ibrahim; Rosen, Marc A.; Pioro, Igor

    Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved “steam” parameters (outlet temperatures up to 625°C and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600°C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the “nuclear” heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted

  9. Development of the DIPRES process for the fast breeder reactor fuel cycle

    SciTech Connect

    Collins, E D; Jackson, M D; Griffin, C W; Rasmussen, D E; Norman, R E

    1984-01-01

    In 1979 the Consolidated Fuel Reprocessing Program (CFRP) at ORNL initiated a program for the development of advanced conversion processes with potential for simplifying and improving the conversion/pellet fabrication flowsheet for recycle plutonium. An evaluation of advanced conversion processes led to the selection of DIPRES (DIrect PREss Spheriodized) for development because it has the largest potential for process and product improvements. DIPRES utilizes a gel sphere conversion process and product to provide a spherical feed material for pellet fabrication. The free-flowing nature of the spherical conversion product allows it to be fed directly to pellet presses (i.e., direct press feed) in place of conventional, mechanically blended powder feed. This is advantageous for remote fabrication. The DIPRES feed is prepared by an internal gelation process.

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

    SciTech Connect

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

    1981-01-01

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

  11. Mathematical modelling and reactor design for multi-cycle bioregeneration of nitrate exhausted ion exchange resin.

    PubMed

    Ebrahimi, Shelir; Roberts, Deborah J

    2016-01-01

    Nitrate contamination is one of the largest issues facing communities worldwide. One of the most common methods for nitrate removal from water is ion exchange using nitrate selective resin. Although these resins have a great capacity for nitrate removal, they are considered non regenerable. The sustainability of nitrate-contaminated water treatment processes can be achieved by regenerating the exhausted resin several times rather than replacing and incineration of exhausted resin. The use of multi-cycle exhaustion/bioregeneration of resin enclosed in a membrane has been shown to be an effective and innovative regeneration method. In this research, the mechanisms for bioregeneration of resin were studied and a mathematical model which incorporated physical desorption process with biological removal kinetics was developed. Regardless of the salt concentration of the solution, this specific resin is a pore-diffusion controlled process (XδD ¯CDr0(5+2α)<1). Also, Thiele modulus was calculated to be between 4 and 12 depending on the temperature and salt concentration. High Thiele modulus (>3) shows that the bioregeneration process is controlled by reaction kinetics and is governed by biological removal of nitrate. The model was validated by comparison to experimental data; the average of R-squared values for cycle 1 to 5 of regeneration was 0.94 ± 0.06 which shows that the developed model predicted the experimental results very well. The model sensitivity for different parameters was evaluated and a model bioreactor design for bioregeneration of highly selective resins was also presented. PMID:26595098

  12. Mathematical modelling and reactor design for multi-cycle bioregeneration of nitrate exhausted ion exchange resin.

    PubMed

    Ebrahimi, Shelir; Roberts, Deborah J

    2016-01-01

    Nitrate contamination is one of the largest issues facing communities worldwide. One of the most common methods for nitrate removal from water is ion exchange using nitrate selective resin. Although these resins have a great capacity for nitrate removal, they are considered non regenerable. The sustainability of nitrate-contaminated water treatment processes can be achieved by regenerating the exhausted resin several times rather than replacing and incineration of exhausted resin. The use of multi-cycle exhaustion/bioregeneration of resin enclosed in a membrane has been shown to be an effective and innovative regeneration method. In this research, the mechanisms for bioregeneration of resin were studied and a mathematical model which incorporated physical desorption process with biological removal kinetics was developed. Regardless of the salt concentration of the solution, this specific resin is a pore-diffusion controlled process (XδD ¯CDr0(5+2α)<1). Also, Thiele modulus was calculated to be between 4 and 12 depending on the temperature and salt concentration. High Thiele modulus (>3) shows that the bioregeneration process is controlled by reaction kinetics and is governed by biological removal of nitrate. The model was validated by comparison to experimental data; the average of R-squared values for cycle 1 to 5 of regeneration was 0.94 ± 0.06 which shows that the developed model predicted the experimental results very well. The model sensitivity for different parameters was evaluated and a model bioreactor design for bioregeneration of highly selective resins was also presented.

  13. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    SciTech Connect

    Bruce G. Schnitzler; Stanley K. Borowski

    2012-07-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine

  14. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    NASA Technical Reports Server (NTRS)

    Schnitzler, Bruce G.; Borowski, Stanley K.

    2012-01-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were

  15. High-Temperature Gas-Cooled Reactor Steam Cycle/Cogeneration Lead Project strategy plan

    SciTech Connect

    1982-03-01

    The strategy for developing the HTGR system and introducing it into the energy marketplace is based on using the most developed technology path to establish a HTGR-Steam Cycle/Cogeneration (SC/C) Lead Project. Given the status of the HTGR-SC/C technology, a Lead Plant could be completed and operational by the mid 1990s. While there is remaining design and technology development that must be accomplished to fulfill technical and licensing requirements for a Lead Project commitment, the major barriers to the realization a HTGR-SC/C Lead Project are institutional in nature, e.g. Project organization and management, vendor/supplier development, cost/risk sharing between the public and private sector, and Project financing. These problems are further exacerbated by the overall pervading issues of economic and regulatory instability that presently confront the utility and nuclear industries. This document addresses the major institutional issues associated with the HTGR-SC/C Lead Project and provides a starting point for discussions between prospective Lead Project participants toward the realization of such a Project.

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

    NASA Technical Reports Server (NTRS)

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

    2007-01-01

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

  17. Single Channel Testing for Characterization of the Direct Gas Cooled Reactor and the SAFE-100 Heat Exchanger

    SciTech Connect

    Bragg-Sitton, S.M.; Kapernick, R.; Godfroy, T.J.

    2004-02-04

    Experiments have been designed to characterize the coolant gas flow in two space reactor concepts that are currently under investigation by NASA Marshall Space Flight Center and Los Alamos National Laboratory: the direct-drive gas-cooled reactor (DDG) and the SAFE-100 heatpipe-cooled reactor (HPR). For the DDG concept, initial tests have been completed to measure pressure drop versus flow rate for a prototypic core flow channel, with gas exiting to atmospheric pressure conditions. The experimental results of the completed DDG tests presented in this paper validate the predicted results to within a reasonable margin of error. These tests have resulted in a re-design of the flow annulus to reduce the pressure drop. Subsequent tests will be conducted with the re-designed flow channel and with the outlet pressure held at 150 psi (1 MPa). Design of a similar test for a nominal flow channel in the HPR heat exchanger (HPR-HX) has been completed and hardware is currently being assembled for testing this channel at 150 psi. When completed, these test programs will provide the data necessary to validate calculated flow performance for these reactor concepts (pressure drop and film temperature rise)

  18. Reactor Meltdown: Critical Zone Processes In Siliciclastics Unlikely To Be Directly Transferable To Carbonates

    NASA Astrophysics Data System (ADS)

    Gulley, J. D.; Cohen, M. J.; Kramer, M. G.; Martin, J. B.; Graham, W. D.

    2013-12-01

    Carbonate terrains cover 20% of Earth's ice-free land and are modified through interactions between rocks, water and biota that couple ecosystems processes to weathering reactions within the critical zone. Weathering in carbonate systems differs from the Critical Zone Reactor model developed for siliciclastic systems because reactions in siliciclastic critical zones largely consist of incongruent weathering (e.g., feldspar to secondary clay minerals) that typically occur in the soil zone within a few meters of the land surface. These incongruent reactions create regolith, which is removed by physical transport mechanisms that drive landscape denudation. In contrast, carbonate critical zones are mostly composed of homogeneous and soluble minerals, which dissolve congruently with the weathering products exported in solution, limiting regolith in the soil mantle to small amounts of insoluble residues. These reactions can extend to depths greater than 2 km below the surface. As water at the land surface drains preferentially through vertical joints and horizontal bedding planes of the carbonate critical zones, it is 'charged' with biologically-derived carbon dioxide, which decreases pH, dissolves carbonate rock, and enlarges subsurface flowpaths through feedbacks between flow and dissolution. Caves are extreme end products of this process and are key morphological features of carbonate critical zones. Caves link surface processes to the deep subsurface and serve as efficient delivery agents for oxygen, carbon and nutrients to zones within the critical zone that are deficient in all three, interrupting vertical and horizontal chemical gradients that would exist if caves were not present. We present select data from air and water-filled caves in the upper Floridan aquifer, Florida, USA, that demonstrate how caves, acting as very large preferential flow paths, alter processes in carbonate relative to siliciclastic critical zones. While caves represent an extreme end

  19. Influence of Observed Diurnal Cycles of Aerosol Optical Depth on Aerosol Direct Radiative Effect

    NASA Technical Reports Server (NTRS)

    Arola, A.; Eck, T. F.; Huttunen, J.; Lehtinen, K. E. J.; Lindfors, A. V.; Myhre, G.; Smirinov, A.; Tripathi, S. N.; Yu, H.

    2013-01-01

    The diurnal variability of aerosol optical depth (AOD) can be significant, depending on location and dominant aerosol type. However, these diurnal cycles have rarely been taken into account in measurement-based estimates of aerosol direct radiative forcing (ADRF) or aerosol direct radiative effect (ADRE). The objective of our study was to estimate the influence of diurnal aerosol variability at the top of the atmosphere ADRE estimates. By including all the possible AERONET sites, we wanted to assess the influence on global ADRE estimates. While focusing also in more detail on some selected sites of strongest impact, our goal was to also see the possible impact regionally.We calculated ADRE with different assumptions about the daily AOD variability: taking the observed daily AOD cycle into account and assuming diurnally constant AOD. Moreover, we estimated the corresponding differences in ADREs, if the single AOD value for the daily mean was taken from the the Moderate Resolution Imaging Spectroradiometer (MODIS) Terra or Aqua overpass times, instead of accounting for the true observed daily variability. The mean impact of diurnal AOD variability on 24 h ADRE estimates, averaged over all AERONET sites, was rather small and it was relatively small even for the cases when AOD was chosen to correspond to the Terra or Aqua overpass time. This was true on average over all AERONET sites, while clearly there can be much stronger impact in individual sites. Examples of some selected sites demonstrated that the strongest observed AOD variability (the strongest morning afternoon contrast) does not typically result in a significant impact on 24 h ADRE. In those cases, the morning and afternoon AOD patterns are opposite and thus the impact on 24 h ADRE, when integrated over all solar zenith angles, is reduced. The most significant effect on daily ADRE was induced by AOD cycles with either maximum or minimum AOD close to local noon. In these cases, the impact on 24 h ADRE was

  20. Performance calculations and research direction for a water enhanced regenerative gas turbine cycle

    NASA Astrophysics Data System (ADS)

    Rogers, L. H.; Archer, D. H.

    A cycle has been conceived that combines compressor cooling, humidification, and regenerative air heating with the added enhancement of direct injection of water into the air flow. In this cycle it is proposed that a fine mist of water be injected into the compressor air stream and a spray or film of water into the regenerator air stream. Water injection into the compressor air flow realizes several benefits: it cools the air flow, reducing the power required for compression and increasing the potential for exhaust heat recovery; it adds mass to the air stream, increasing the power produced by expansion; and it reduces the amount of cooling bleed air required by increasing the specific heat and decreasing the temperature of the cooling air stream. The greatest benefit would be derived from spraying a fine mist of water directly into the existing air flow into or before the compressor so that cooling and compression would occur simultaneously. This may be accomplished by entraining the water droplets in the inlet air flow or by introducing the water in stages during compression. An alternative and less technically challenging approach is to extract the air stream to a saturation chamber and then reintroduce the air stream into the compressor. This approach is not as desirable because it would increase the equipment cost and add a significant pressure drop penalty. The second use of water in this cycle is in water-assisted regeneration. The heat capacity of the hot stream in regenerators is greater than the heat capacity of the cool stream because of the increased mass flow and specific heat of the combustion products. This imbalance leads to a less than ideal exhaust heat recovery since the cool air stream is unable to absorb all of the available heat. If water is injected into the cool stream in the regenerator, some of the available heat is used to vaporize the water, allowing additional heat recovery and also adding mass to the air flow. Also, the effectiveness

  1. Life cycle assessment of hydrogen production from S-I thermochemical process coupled to a high temperature gas reactor

    SciTech Connect

    Giraldi, M. R.; Francois, J. L.; Castro-Uriegas, D.

    2012-07-01

    The purpose of this paper is to quantify the greenhouse gas (GHG) emissions associated to the hydrogen produced by the sulfur-iodine thermochemical process, coupled to a high temperature nuclear reactor, and to compare the results with other life cycle analysis (LCA) studies on hydrogen production technologies, both conventional and emerging. The LCA tool was used to quantify the impacts associated with climate change. The product system was defined by the following steps: (i) extraction and manufacturing of raw materials (upstream flows), (U) external energy supplied to the system, (iii) nuclear power plant, and (iv) hydrogen production plant. Particular attention was focused to those processes where there was limited information from literature about inventory data, as the TRISO fuel manufacture, and the production of iodine. The results show that the electric power, supplied to the hydrogen plant, is a sensitive parameter for GHG emissions. When the nuclear power plant supplied the electrical power, low GHG emissions were obtained. These results improve those reported by conventional hydrogen production methods, such as steam reforming. (authors)

  2. Development of a plant dynamics computer code for analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to a natural circulation lead-cooled fast reactor.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

    STAR-LM is a lead-cooled pool-type fast reactor concept operating under natural circulation of the coolant. The reactor core power is 400 MWt. The open-lattice core consists of fuel pins attached to the core support plate, (the does not consist of removable fuel assemblies). The coolant flows outside of the fuel pins. The fuel is transuranic nitride, fabricated from reprocessed LWR spent fuel. The cladding material is HT-9 stainless steel; the steady-state peak cladding temperature is 650 C. The coolant is single-phase liquid lead under atmospheric pressure; the core inlet and outlet temperatures are 438 C and 578 C, respectively. (The Pb coolant freezing and boiling temperatures are 327 C and 1749 C, respectively). The coolant is contained inside of a reactor vessel. The vessel material is Type 316 stainless steel. The reactor is autonomous meaning that the reactor power is self-regulated based on inherent reactivity feedbacks and no external power control (through control rods) is utilized. The shutdown (scram) control rods are used for startup and shutdown and to stop the fission reaction in case of an emergency. The heat from the reactor is transferred to the S-CO{sub 2} Brayton cycle in in-reactor heat exchangers (IRHX) located inside the reactor vessel. The IRHXs are shell-and-tube type heat exchangers with lead flowing downwards on the shell side and CO{sub 2} flowing upwards on the tube side. No intermediate circuit is utilized. The guard vessel surrounds the reactor vessel to contain the coolant, in the very unlikely event of reactor vessel failure. The Reactor Vessel Auxiliary Cooling System (RVACS) implementing the natural circulation of air flowing upwards over the guard vessel is used to cool the reactor, in the case of loss of normal heat removal through the IRHXs. The RVACS is always in operation. The gap between the vessels is filled with liquid lead-bismuth eutectic (LBE) to enhance the heat removal by air by significantly reducing the thermal

  3. Applications of plasma core reactors to terrestrial energy systems

    NASA Technical Reports Server (NTRS)

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

    1974-01-01

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

  4. Scrap tyre recycling process with molten zinc as direct heat transfer and solids separation fluid: A new reactor concept.

    PubMed

    Riedewald, Frank; Goode, Kieran; Sexton, Aidan; Sousa-Gallagher, Maria J

    2016-01-01

    Every year about 1.5 billion tyres are discarded worldwide representing a large amount of solid waste, but also a largely untapped source of raw materials. The objective of the method was to prove the concept of a novel scrap tyre recycling process which uses molten zinc as the direct heat transfer fluid and, simultaneously, uses this media to separate the solids products (i.e. steel and rCB) in a sink-float separation at an operating temperature of 450-470 °C. This methodology involved: •construction of the laboratory scale batch reactor,•separation of floating rCB from the zinc,•recovery of the steel from the bottom of the reactor following pyrolysis.

  5. Scrap tyre recycling process with molten zinc as direct heat transfer and solids separation fluid: A new reactor concept.

    PubMed

    Riedewald, Frank; Goode, Kieran; Sexton, Aidan; Sousa-Gallagher, Maria J

    2016-01-01

    Every year about 1.5 billion tyres are discarded worldwide representing a large amount of solid waste, but also a largely untapped source of raw materials. The objective of the method was to prove the concept of a novel scrap tyre recycling process which uses molten zinc as the direct heat transfer fluid and, simultaneously, uses this media to separate the solids products (i.e. steel and rCB) in a sink-float separation at an operating temperature of 450-470 °C. This methodology involved: •construction of the laboratory scale batch reactor,•separation of floating rCB from the zinc,•recovery of the steel from the bottom of the reactor following pyrolysis. PMID:27274458

  6. Sensitivity simulations with direct radiative forcing by aeolian dust during glacial cycles

    NASA Astrophysics Data System (ADS)

    Bauer, E.; Ganopolski, A.

    2014-01-01

    Possible feedback effects between aeolian dust, climate and ice sheets are studied for the first time with an Earth system model of intermediate complexity over the late Pleistocene period. Correlations between climate variables and dust deposits suggest that aeolian dust potentially plays an important role for the evolution of glacial cycles. Here climatic effects from the dust direct radiative forcing (DRF) caused by absorption and scattering of solar radiation are investigated. Key factors controlling the dust DRF are the atmospheric dust distribution and the absorption-scattering efficiency of dust aerosols. Effective physical parameters in the description of these factors are varied within uncertainty ranges known from available data and detailed model studies. Although the parameters are reasonably constrained by use of these studies, the simulated dust DRF spans a wide uncertainty range related to nonlinear dependencies. In our simulations, the dust DRF is highly localized. Medium-range parameters result in negative DRF of several W m-2 in regions close to major dust sources and negligible values elsewhere. In case of high absorption efficiency, the local dust DRF can reach positive values and the global mean DRF can be insignificantly small. In case of low absorption efficiency, the dust DRF can produce a significant global cooling in glacial periods which leads to a doubling of the maximum glacial ice volume relative to the case with small dust DRF. DRF-induced temperature and precipitation changes can either be attenuated or amplified through a feedback loop involving the dust cycle. The sensitivity experiments suggest that depending on dust optical parameters the DRF has the potential to either damp or reinforce glacial-interglacial climate changes.

  7. Sensitivity simulations with direct shortwave radiative forcing by aeolian dust during glacial cycles

    NASA Astrophysics Data System (ADS)

    Bauer, E.; Ganopolski, A.

    2014-07-01

    Possible feedback effects between aeolian dust, climate and ice sheets are studied for the first time with an Earth system model of intermediate complexity over the late Pleistocene period. Correlations between climate and dust deposition records suggest that aeolian dust potentially plays an important role for the evolution of glacial cycles. Here climatic effects from the dust direct radiative forcing (DRF) caused by absorption and scattering of solar radiation are investigated. Key elements controlling the dust DRF are the atmospheric dust distribution and the absorption-scattering efficiency of dust aerosols. Effective physical parameters in the description of these elements are varied within uncertainty ranges known from available data and detailed model studies. Although the parameters can be reasonably constrained, the simulated dust DRF spans a~wide uncertainty range related to the strong nonlinearity of the Earth system. In our simulations, the dust DRF is highly localized. Medium-range parameters result in negative DRF of several watts per square metre in regions close to major dust sources and negligible values elsewhere. In the case of high absorption efficiency, the local dust DRF can reach positive values and the global mean DRF can be insignificantly small. In the case of low absorption efficiency, the dust DRF can produce a significant global cooling in glacial periods, which leads to a doubling of the maximum glacial ice volume relative to the case with small dust DRF. DRF-induced temperature and precipitation changes can either be attenuated or amplified through a feedback loop involving the dust cycle. The sensitivity experiments suggest that depending on dust optical parameters, dust DRF has the potential to either damp or reinforce glacial-interglacial climate changes.

  8. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Satya, Octavianus Cakra; Monado, Fiber; Su'ud, Zaki; Sekimoto, Hiroshi

    2016-03-01

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on "Region-8" and "Region-10" core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  9. Short Contact Time Direct Coal Liquefactionn Using a Novel Batch Reactor. Quarterly Report. May 16 - August 15, 1996

    SciTech Connect

    He Huang; Michael T. Klein; William H. Calkins

    1996-08-30

    The objective of this research is to optimize the design and operation of the bench scale batch reactor (SCTBR) for studying direct coal liquefaction at short contact times (.01 to 10 minutes or longer). Additional objectives are to study the kinetics of direct coal liquefaction particularly at short reaction times and to investigate the role of organic oxygen components of coal and their reaction pathways during coal liquefaction. Many of those objectives have already been achieved. This quarterly report discusses further kinetic studies of the liquefaction of Illinois #6 bituminous coal, Wyodak-Anderson subbituminous coal, and Pittsburgh #8 bituminous coal. The thermodynamic characteristics of the extraction stage at the start of the liquefaction process in the liquefaction of Illinois #6 coal is also discussed. Further work has also been done to attempt to clarify the role of the liquefaction solvent in the direct liquefaction process.

  10. The effectiveness of using the combined-cycle technology in a nuclear power plant unit equipped with an SVBR-100 reactor

    NASA Astrophysics Data System (ADS)

    Kasilov, V. F.; Dudolin, A. A.; Gospodchenkov, I. V.

    2015-05-01

    The design of a modular SVBR-100 reactor with a lead-bismuth alloy liquid-metal coolant is described. The basic thermal circuit of a power unit built around the SVBR-100 reactor is presented together with the results of its calculation. The gross electrical efficiency of the turbine unit driven by saturated steam at a pressure of 6.7 MPa is estimated at η{el/gr} = 35.5%. Ways for improving the efficiency of this power unit and increasing its power output by applying gas-turbine and combined-cycle technologies are considered. With implementing a combined-cycle power-generating system comprising two GE-6101FA gas-turbine units with a total capacity of 140 MW, it becomes possible to obtain the efficiency of the combined-cycle plant equipped with the SVBR-100 reactor η{el/gr} = 45.39% and its electrical power output equal to 328 MW. The heat-recovery boiler used as part of this power installation generates superheated steam with a temperature of 560°C, due to which there is no need to use a moisture separator/steam reheater in the turbine unit thermal circuit.

  11. Current Status of the Experiment on Direct Measurement of Neutron-Neutron Scattering Length at the Reactor YAGUAR

    SciTech Connect

    Furman, W. I.; Muzichka, A. Yu.; Lychagin, E. V.; Nekhaev, G. V.; Sharapov, E. I.; Shvetsov, V. N.; Strelkov, A. V.; Crawford, B. E.; Stephenson, S. L.; Howell, C. R.; Tornow, W.; Kandiev, Ya.; Levakov, B. G.; Litvin, V. I.; Lyzhin, A. E.; Tchernukhin, Yu. I.; Mitchell, G. E.

    2009-03-31

    A new experiment was proposed in 2002 to perform the first direct measurement of neutron-neutron scattering on the powerful pulsed reactor YAGUAR located at Snezhinsk, Ural region, Russia. Extensive efforts were made to model the background conditions and to optimize the set-up design. To make the experiment feasible it was necessary to suppress the background from various origins by more than 16 orders of magnitude for thermal neutrons and 14 orders of magnitude for fast neutrons. In 2003 a channel was drilled under the reactor and equipped for time-of-flight measurements. During the next two years at this channel there were carried out a series of test experiments aimed at verifying the accuracy of the background modeling. Good agreement of the measured results with the calculated values enabled us to make the final design of the full scale set-up. During 2005-2006 the experimental system was manufactured. After vacuum tests at JINR the set-up was mounted at the YAGUAR reactor hall. In 2006-2007 calibration measurements with noble gases were performed. The results confirmed the validity of the modeling of the full scale experiment and verified the calibration. The first preliminary experiments for nn-scattering were performed in April 2008. These recent results are discussed.

  12. Conceptual Design for a 2 GW Inertial Fusion Energy (IFE) Direct-Drive Power Reactor Employing Magnetic Intervention

    NASA Astrophysics Data System (ADS)

    Tresemer, K. R.; Gentile, C. A.

    2007-11-01

    Presented is a conceptual design for a 2 GW IFE direct drive fusion power reactor. This design employs a cusp field to deflect IFE-generated ions away from the dry first wall of the target chamber and into specifically designed ion dumps. The reactor operates at 5 Hz, consuming ˜450,000 tritium targets/day, injected at >100 m/s into the target chamber and uniformly illuminated by laser light, stimulating detonation. The resulting fusion energy is collected by equatorial ion dumps equipped with heat exchangers. The reactor will breed and recycle its own fuel through the use of breeder blankets and a fuel recovery system. To minimize target-particle interference, the chamber will be kept at <0.5 mTorr through the use of magnetically levitated turbomolecular pumps (TMPs) and corresponding backing pumps. Under investigation are the principles of magnetohydrodynamics (MHD) which may be applied to attenuate and harness the energy residing in the post detonation ion fields.

  13. Cell cycle and p53 gate the direct conversion of human fibroblasts to dopaminergic neurons.

    PubMed

    Jiang, Houbo; Xu, Zhimin; Zhong, Ping; Ren, Yong; Liang, Gaoyang; Schilling, Haley A; Hu, Zihua; Zhang, Yi; Wang, Xiaomin; Chen, Shengdi; Yan, Zhen; Feng, Jian

    2015-01-01

    The direct conversion of fibroblasts to induced dopaminergic (iDA) neurons and other cell types demonstrates the plasticity of cell fate. The low efficiency of these relatively fast conversions suggests that kinetic barriers exist to safeguard cell-type identity. Here we show that suppression of p53, in conjunction with cell cycle arrest at G1 and appropriate extracellular environment, markedly increase the efficiency in the transdifferentiation of human fibroblasts to iDA neurons by Ascl1, Nurr1, Lmx1a and miR124. The conversion is dependent on Tet1, as G1 arrest, p53 knockdown or expression of the reprogramming factors induces Tet1 synergistically. Tet1 knockdown abolishes the transdifferentiation while its overexpression enhances the conversion. The iDA neurons express markers for midbrain DA neurons and have active dopaminergic transmission. Our results suggest that overcoming these kinetic barriers may enable highly efficient epigenetic reprogramming in general and will generate patient-specific midbrain DA neurons for Parkinson's disease research and therapy. PMID:26639555

  14. Determination of aerobic work and power on a rope-braked cycle ergometer by direct measurement.

    PubMed

    Gordon, Rae S; Franklin, Kathryn L; Baker, Julien S; Davies, Bruce

    2006-08-01

    The purpose of this study was to compare the power and work outputs of a cycle ergometer using the manufacturer's guidelines, with calculations using direct flywheel velocity and brake torque. A further aim was to compare the values obtained with those supplied by the manufacturer. A group of 10 male participants were asked to pedal a Monark 824E ergometer at a constant cadence of 60 r/min for a period of 3 min against a resistive mass of 3 kg. The flywheel velocity was measured using a tachometer. The brake force was determined by measuring the tension in the rope on either side of the flywheel. The calculated mean power was 147.45 +/- 6.5 W compared with the Monark value of 183 +/- 3.7 W. The difference between the methods for power estimation was 18% and was statistically significant (p < 0.01). The mean work done by the participants during the 3 min period was found to be 26 460 +/- 1145 J compared with the Monark value of 33,067 +/- 648 J (p < 0.01). The Monark formulae currently used to determine the power and work done by a participant overestimates the actual values required to overcome the resistance. There findings have far-reaching implications in the physiological assessment of athletic, sedentary, and diseased populations. PMID:16900228

  15. Physiological assessment of isolated running does not directly replicate running capacity after triathlon-specific cycling.

    PubMed

    Etxebarria, Naroa; Hunt, Julie; Ingham, Steve; Ferguson, Richard

    2014-01-01

    Triathlon running is affected by prior cycling and power output during triathlon cycling is variable in nature. We compared constant and triathlon-specific variable power cycling and their effect on subsequent submaximal running physiology. Nine well-trained male triathletes (age 24.6 ± 4.6 years, [Formula: see text] 4.5 ± 0.4 L · min(-1); mean ± SD) performed a submaximal incremental run test, under three conditions: no prior exercise and after a 1 h cycling trial at 65% of maximal aerobic power with either a constant or a variable power profile. The variable power protocol involved multiple 10-90 s intermittent efforts at 40-140% maximal aerobic power. During cycling, pulmonary ventilation (22%, ± 14%; mean; ± 90% confidence limits), blood lactate (179%, ± 48%) and rating of perceived exertion (7.3%, ± 10.2%) were all substantially higher during variable than during constant power cycling. At the start of the run, blood lactate was 64%, ± 61% higher after variable compared to constant power cycling, which decreased running velocity at 4 mM lactate threshold by 0.6, ± 0.9 km · h(-1). Physiological responses to incremental running are negatively affected by prior cycling and, to a greater extent, by variable compared to even-paced cycling. Testing and training of triathletes should account foe higher physiological cost of triathlon-specific cycling and its effect on subsequent running.

  16. Direct radioimmunoassay of urinary estrogen and pregnanediol glucuronides during the menstrual cycle

    SciTech Connect

    Stanczyk, F.Z.; Miyakawa, I.; Goebelsmann, U.

    1980-06-15

    Assays measuring immunoreactive estrone glucuronide (E/sub 1/G), estradiol-3-glucuronide (E/sub 2/-3G), estradiol-17..beta..-glucuronide (E/sub 2/-17G), estriol-3-glucuronide (E/sub 3/-3G), estriol-16..cap alpha..-glucuronide (E/sub 3/-16G), and pregnanediol-3..cap alpha..-glucuronide (Pd-3G) directly in diluted urine were developed and validated. These estrogen and pregnanediol glucuronide fractions were measured in aliquots of 24-hour and overnight samples of urine collected daily from seven women for one menstrual cycle. Urinary hormone excretion was correlated with daily serum estradiol (E/sub 2/), progesterone (P), and lutenizing hormonee (LH) levels. A sharp midcycle LH peak preceded by a preovulatory rise in serum E/sub 2/ and followed by luteal phase serum P levels were noted in each of the seven apparently ovulatory cycles. Twenty-four-hour and overnight urinary excretion patterns of estrogen glucuronides were similar to those of serum E/sub 2/. Of the five estrogen glucuronide fractions tested, excretion of E/sub 2/-17G exhibited the earliest and steepest ascending slope of the preovulatory estrogen surge and correlated best with serum E/sub 2/ levels. Urinary excretion of E/sub 1/-G, E/sub 2/-3G, and E/sub 3/-16G also showed an early and steep preovulatory rise and preceded that of E/sub 3/-3G, whereas urinary excretion of E/sub 3/-3G exhibited the poorest correlation with serum E/sub 2/ concentrations. The urinary excretion of Pd-3G rose parallel to serum P levels and was markedly elevated 2 to 3 days after the midcycle LH peak in both 24-hour and overnight collections of urine. These results indicate that among the urinary estrogen conjugate fractions tested, E/sub 2/-17G is the one that most suitably predicts ovulation.

  17. Direct visualization by electron microscopy of the weakly bound intermediates in the actomyosin adenosine triphosphatase cycle.

    PubMed Central

    Pollard, T D; Bhandari, D; Maupin, P; Wachsstock, D; Weeds, A G; Zot, H G

    1993-01-01

    We used a novel stopped-flow/rapid-freezing machine to prepare the transient intermediates in the actin-myosin adenosine triphosphatase (ATPase) cycle for direct observation by electron microscopy. We focused on the low affinity complexes of myosin-adenosine triphosphate (ATP) and myosin-adenosine diphosphate (ADP)-Pi with actin filaments since the transition from these states to the high affinity actin-myosin-ADP and actin-myosin states is postulated to generate the molecular motion that drives muscle contraction and other types of cellular movements. After rapid freezing and metal replication of mixtures of myosin subfragment-1, actin filaments, and ATP, the structure of the weakly bound intermediates is indistinguishable from nucleotide-free rigor complexes. In particular, the average angle of attachment of the myosin head to the actin filament is approximately 40 degrees in both cases. At all stages in the ATPase cycle, the configuration of most of the myosin heads bound to actin filaments is similar, and the part of the myosin head preserved in freeze-fracture replicas does not tilt by more than a few degrees during the transition from the low affinity to high affinity states. In contrast, myosin heads chemically cross-linked to actin filaments differ in their attachment angles from ordered at 40 degrees without ATP to nearly random in the presence of ATP when viewed by negative staining (Craig, R., L.E. Greene, and E. Eisenberg. 1985. Proc. Natl. Acad. Sci. USA. 82:3247-3251, and confirmed here), freezing in vitreous ice (Applegate, D., and P. Flicker. 1987. J. Biol. Chem. 262:6856-6863), and in replicas of rapidly frozen samples. This suggests that many of the cross-linked heads in these preparations are dissociated from but tethered to the actin filaments in the presence of ATP. These observations suggest that the molecular motion produced by myosin and actin takes place with the myosin head at a point some distance from the actin binding site or does not

  18. Investigation of plant control strategies for the supercritical C0{sub 2}Brayton cycle for a sodium-cooled fast reactor using the plant dynamics code.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J.

    2011-04-12

    The development of a control strategy for the supercritical CO{sub 2} (S-CO{sub 2}) Brayton cycle has been extended to the investigation of alternate control strategies for a Sodium-Cooled Fast Reactor (SFR) nuclear power plant incorporating a S-CO{sub 2} Brayton cycle power converter. The SFR assumed is the 400 MWe (1000 MWt) ABR-1000 preconceptual design incorporating metallic fuel. Three alternative idealized schemes for controlling the reactor side of the plant in combination with the existing automatic control strategy for the S-CO{sub 2} Brayton cycle are explored using the ANL Plant Dynamics Code together with the SAS4A/SASSYS-1 Liquid Metal Reactor (LMR) Analysis Code System coupled together using the iterative coupling formulation previously developed and implemented into the Plant Dynamics Code. The first option assumes that the reactor side can be ideally controlled through movement of control rods and changing the speeds of both the primary and intermediate coolant system sodium pumps such that the intermediate sodium flow rate and inlet temperature to the sodium-to-CO{sub 2} heat exchanger (RHX) remain unvarying while the intermediate sodium outlet temperature changes as the load demand from the electric grid changes and the S-CO{sub 2} cycle conditions adjust according to the S-CO{sub 2} cycle control strategy. For this option, the reactor plant follows an assumed change in load demand from 100 to 0 % nominal at 5 % reduction per minute in a suitable fashion. The second option allows the reactor core power and primary and intermediate coolant system sodium pump flow rates to change autonomously in response to the strong reactivity feedbacks of the metallic fueled core and assumed constant pump torques representing unchanging output from the pump electric motors. The plant behavior to the assumed load demand reduction is surprising close to that calculated for the first option. The only negative result observed is a slight increase in the intermediate

  19. Direct photodetachment of F- by mid-infrared few-cycle femtosecond laser pulses

    NASA Astrophysics Data System (ADS)

    Shearer, S. F. C.; Monteith, M. R.

    2013-09-01

    The recent adiabatic saddle-point approach of Shearer [Phys. Rev. APLRAAN1050-294710.1103/PhysRevA.84.033409 84, 033409 (2011)] is extended to multiphoton detachment of negative ions with outer p-state electrons. This theory is applied to investigate the strong-field photodetachment dynamics of F- ions exposed to few-cycle femtosecond laser pulses, without taking into account the rescattering mechanism. Numerical calculations are considered for mid-infrared laser wavelengths of 1300 and 1800 nm at laser intensities of 7.7 × 1012, 1.1 × 1013, and 1.3 × 1013 W/cm2. Two-dimensional momenta saddle-point spectra exhibit a distinct distribution in the shape of a “smile” in the complex-time plane. Electron momentum distribution maps of direct electrons are investigated. These produce a distinct pattern of above-threshold detachment (ATD) concentric rings due to constructive and destructive quantum interference of electrons detached from their parent ions. Probability detachment distributions presented, capturing the influence of saturation effects that are found to become more significant with increasing laser intensity at a fixed wavelength. ATD photoangular distributions as functions of laser intensity and wavelength near channel closings are also investigated and found to be sensitive to initial-state symmetry. Nonmonotonic structures observed in the ejected photoelectron energy spectra are attributed to interference effects from coherent electronic wave packets. Additionally the profiles of all the photoelectron emission spectra show strong dependence on the carrier-envelope phase, indicating that it is a reliable parameter for characterizing the wave form of the pulse.

  20. Direct chromatin PCR (DC-PCR): hypotonic conditions allow differentiation of chromatin states during thermal cycling.

    PubMed

    Vatolin, Sergei; Khan, Shahper N; Reu, Frederic J

    2012-01-01

    Current methods to study chromatin configuration are not well suited for high throughput drug screening since they require large cell numbers and multiple experimental steps that include centrifugation for isolation of nuclei or DNA. Here we show that site specific chromatin analysis can be achieved in one step by simply performing direct chromatin PCR (DC-PCR) on cells. The basic underlying observation was that standard hypotonic PCR buffers prevent global cellular chromatin solubilization during thermal cycling while more loosely organized chromatin can be amplified. Despite repeated heating to >90 °C, 41 of 61 tested 5' sequences of silenced genes (CDKN2A, PU.1, IRF4, FOSB, CD34) were not amplifiable while 47 could be amplified from expressing cells. Two gene regions (IRF4, FOSB) even required pre-heating of cells in isotonic media to allow this differentiation; otherwise none of 19 assayed sequences yielded PCR products. Cells with baseline expression or epigenetic reactivation gave similar DC-PCR results. Silencing during differentiation of CD34 positive cord blood cells closed respective chromatin while treatment of myeloma cells with an IRF4 transcriptional inhibitor opened a site to DC-PCR that was occupied by RNA polymerase II and NFκB as determined by ChIP. Translation into real-time PCR can not be achieved with commercial real-time PCR buffers which potently open chromatin, but even with simple ethidium bromide addition to standard PCR mastermix we were able to identify hits in small molecules screens that suppressed IRF4 expression or reactivated CDKN2A in myeloma cells using densitometry or visual inspection of PCR plates under UV light. While need in drug development inspired this work, application to genome-wide analysis appears feasible using phi29 for selective amplification of open cellular chromatin followed by library construction from supernatants since such supernatants yielded similar results as gene specific DC-PCR.

  1. Sinusoidal potential cycling operation of a direct ethanol fuel cell to improving carbon dioxide yields

    NASA Astrophysics Data System (ADS)

    Majidi, Pasha; Pickup, Peter G.

    2014-12-01

    A direct ethanol fuel cell has been operated under sinusoidal (AC) potential cycling conditions in order to increase the yield of carbon dioxide and thereby increase cell efficiency relative to operation at a fixed potential. At 80 °C, faradaic yields of CO2 as high as 25% have been achieved with a PtRu anode catalyst, while the maximum CO2 production at constant potential was 13%. The increased yields under cycling conditions have been attributed to periodic oxidative stripping of adsorbed CO. These results will be important in the optimization of operating conditions for direct ethanol fuel cells, where the benefits of potential cycling are projected to increase as catalysts that produce CO2 more efficiently are implemented.

  2. Structure and thermal cycling stability of a hafnium monocarbide reinforced directionally solidified cobalt-base eutectic alloy

    NASA Technical Reports Server (NTRS)

    Kim, Y. G.

    1975-01-01

    A nominal composition of Co-15Cr-2ONi-10.5 Hf-0.7 C (NASA-HAFCO-11) was directionally solidified at 0.8 cm/hr growth rate to produce aligned HfC in a cobalt matrix alloy. The aligned HfC fibers were present as rod and plate types. The diameter of the aligned fibers was about 1 micron, with volume fraction in the range of 11 to 15 percent. The growth direction of the fibers was parallel to the 100. The NASA-HAFCO-11 alloy was subjected to thermal cycling between 425 deg and 1100 C, using a 2.5 minute cycle. No microstructural degradation of the HfC fibers in the alloy was observed after 2500 cycles.

  3. Influences of Scholarship Aid on the Social Exchange Cycle: A Qualitative Exploration of Scholarship Recipients and Direct Reciprocity

    ERIC Educational Resources Information Center

    Forrest, Jeannie

    2010-01-01

    Social exchange theory asserts that individuals who receive a gift will be pressed by an internal sense of obligation to give back in turn (Mauss, 2002). While there is a great deal of literature devoted to giving, there is little literature about the receiving end of the exchange cycle. Deeply impacted by the effects of direct reciprocity are…

  4. Efficient Silicon Reactor

    NASA Technical Reports Server (NTRS)

    Bates, H. E.; Hill, D. M.; Jewett, D. N.

    1983-01-01

    High-purity silicon efficiently produced and transferred by continuous two-cycle reactor. New reactor operates in relatively-narrow temperature rate and uses large surfaces area to minimize heat expenditure and processing time in producing silicon by hydrogen reduction of trichlorosilane. Two cycles of reactor consists of silicon production and removal.

  5. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  6. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  7. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  8. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  9. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  10. Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK™

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Sanchez, Travis

    2005-02-01

    The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK™ (Simulink, 2004). SIMULINK™ is a development environment packaged with MatLab™ (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion components such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK™ models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK™ modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator).

  11. Evaluation of neutron background in cryogenic Germanium target for WIMP direct detection when using reactor neutrino detector as neutron veto

    NASA Astrophysics Data System (ADS)

    Xu, Ye; Lan, Jieqin; Bai, Ying; Gao, Weiwei

    2016-09-01

    A direct WIMP (Weakly Interacting Massive Particle) detector with a neutron veto system is designed to better reject neutrons. An experimental configuration is studied in the present paper: 984 Ge modules are placed inside a reactor neutrino detector. In order to discriminate between nuclear and electron recoil, both ionization and heat signatures are measured using cryogenic germanium detectors in this detection. The neutrino detector is used as a neutron veto device. The neutron background for the experimental design has been estimated using the Geant4 simulation. The results show that the neutron background can decrease to O(0.01) events per year per tonne of high purity Germanium. We calculate the sensitivity to spin-independent WIMP-nucleon elastic scattering. An exposure of one tonne × year could reach a cross-section of about 2×10-11 pb.

  12. Short contact time direct coal liquefaction using a novel batch reactor. Quarterly progress report, January 1--May 15, 1995

    SciTech Connect

    Klein, M.T.; Calkins, W.H.

    1995-05-31

    The objective of this research is to optimize the design and operation of the bench scale batch reactor for coal liquefaction at short contact times (0.01 to 10 minutes or longer). Additional objectives are to study the kinetics of direct coal liquefaction particularly at short reaction times, and to investigate the role of the organic oxygen components of coal and their reaction pathways during liquefaction. Experimental progress is reported for uncatalyzed liquefactions, catalyzed liquefactions, liquefaction in the presence of solvents other than tetralin, and kinetics of gas formation during coal liquefaction. Analytical methods were developed for the determination of the boiling range of coal liquids by thermogravimetric analysis and the determination of phenolic hydroxyl in coal, coal liquids, and coal residues.

  13. The 22-Year Hale Cycle in Cosmic Ray Flux - Evidence for Direct Heliospheric Modulation

    NASA Astrophysics Data System (ADS)

    Thomas, Simon; Owens, Mathew; Lockwood, Mike

    2013-04-01

    The ability to predict times of greater fluxes of galactic cosmic rays is important for reducing the hazards caused by these energetic particles on satellite communications, aviation and astronauts. During the 22-year Hale cycle, we see a difference in shape from a 'flat topped' to a 'spiked topped' peak in cosmic ray flux time series. It is thought that differing drift patterns for when the northern solar pole is predominantly positive (qA>0) to when the northern pole is negative (qA<0) cause this difference in cosmic ray modulation. Here, we demonstrate a link between cosmic ray modulation and properties of the large-scale heliospheric magnetic field during the declining phase of the solar cycle, when the difference between qA>0 and qA<0 cycles is most apparent. The results suggest that drift affects may not be the sole mechanism responsible for the Hale Cycle in cosmic ray flux at Earth. Further to this it is suggested that the Hale cycle in cosmic ray flux may be primarily limited to the grand solar maximum of the space-age.

  14. Involvement of condensin-directed gene associations in the organization and regulation of chromosome territories during the cell cycle

    PubMed Central

    Iwasaki, Osamu; Corcoran, Christopher J.; Noma, Ken-ichi

    2016-01-01

    Chromosomes are not randomly disposed in the nucleus but instead occupy discrete sub-nuclear domains, referred to as chromosome territories. The molecular mechanisms that underlie the formation of chromosome territories and how they are regulated during the cell cycle remain largely unknown. Here, we have developed two different chromosome-painting approaches to address how chromosome territories are organized in the fission yeast model organism. We show that condensin frequently associates RNA polymerase III-transcribed genes (tRNA and 5S rRNA) that are present on the same chromosomes, and that the disruption of these associations by condensin mutations significantly compromises the chromosome territory arrangement. We also find that condensin-dependent intra-chromosomal gene associations and chromosome territories are co-regulated during the cell cycle. For example, condensin-directed gene associations occur to the least degree during S phase, with the chromosomal overlap becoming largest. In clear contrast, condensin-directed gene associations become tighter in other cell-cycle phases, especially during mitosis, with the overlap between the different chromosomes being smaller. This study suggests that condensin-driven intra-chromosomal gene associations contribute to the organization and regulation of chromosome territories during the cell cycle. PMID:26704981

  15. The seasonal cycle of the Atlantic Jet dynamics in the Alboran Sea: direct atmospheric forcing versus Mediterranean thermohaline circulation

    NASA Astrophysics Data System (ADS)

    Macias, Diego; Garcia-Gorriz, Elisa; Stips, Adolf

    2016-02-01

    The Atlantic Jet (AJ) is the inflow of Atlantic surface waters into the Mediterranean Sea. This geostrophically adjusted jet fluctuates in a wide range of temporal scales from tidal to subinertial, seasonal, and interannual modifying its velocity and direction within the Alboran Sea. At seasonal scale, a clearly defined cycle has been previously described, with the jet being stronger and flowing towards the northeast during the first half of the year and weakening and flowing more southwardly towards the end of the year. Different hypothesis have been proposed to explain this fluctuation pattern but, up to now, no quantitative assessment of the importance of the different forcings for this seasonality has been provided. Here, we use a 3D hydrodynamic model of the entire Mediterranean Sea forced at the surface with realistic atmospheric conditions to study and quantify the importance of the different meteorological forcings on the velocity and direction of the AJ at seasonal time scale. We find that the direct effects of local zonal wind variations are much more important to explain extreme collapse events when the jet dramatically veers southward than to the seasonal cycle itself while sea level pressure variations over the Mediterranean seem to have very little direct effect on the AJ behavior at monthly and longer time scales. Further model results indicate that the annual cycle of the thermohaline circulation is the main driver of the seasonality of the AJ dynamics in the model simulations. The annual cycles in local wind forcing and SLP variations over the Mediterranean have no causal relationship with the AJ seasonality.

  16. Spectrally efficient terabit optical transmission with Nyquist 64-QAM half-cycle subcarrier modulation and direct detection.

    PubMed

    Zou, Kaiheng; Zhu, Yixiao; Zhang, Fan; Chen, Zhangyuan

    2016-06-15

    We demonstrate 1.728  Tb/s(16×108  Gb/s) direct-detection wavelength division multiplexing (WDM) transmission over 80 km standard single mode fiber (SSMF) with Nyquist 64-ary quadrature amplitude modulation (64-QAM) and half-cycle subcarrier modulation. Each channel carries single sideband 18 GBaud 64-QAM signal and the channel spacing is 27 GHz. Considering 20% soft-decision forward error correction and frame redundancy, a net spectral efficiency record of 3.25 b/s/Hz is achieved for 100 G single polarization direct-detection WDM transmission.

  17. Breeding of {sup 233}U in the thorium–uranium fuel cycle in VVER reactors using heavy water

    SciTech Connect

    Marshalkin, V. E. Povyshev, V. M.

    2015-12-15

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the {sup 233}U–{sup 232}Th oxide fuel of water-moderated reactors with variable water composition (D{sub 2}O, H{sub 2}O) that ensures breeding of the {sup 233}U and {sup 235}U isotopes. The method is comparatively simple to implement.

  18. Breeding of 233U in the thorium-uranium fuel cycle in VVER reactors using heavy water

    NASA Astrophysics Data System (ADS)

    Marshalkin, V. E.; Povyshev, V. M.

    2015-12-01

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the 233U-232Th oxide fuel of water-moderated reactors with variable water composition (D2O, H2O) that ensures breeding of the 233U and 235U isotopes. The method is comparatively simple to implement.

  19. Opportunities to reduce consumption of natural uranium in reactor SVBR-75/100 when changing over to the closed fuel cycle

    SciTech Connect

    Toshinsky, G.I.; Komlev, O.G.; Mel'nikov, K.G.; Novikova, N.N.

    2007-07-01

    The design of reactor SVBR-75/100 allows it to operate using different types of fuel and in different fuel cycles without changing its design and deteriorating its safety characteristics. Fuel-at-once refueling adopted in the design (lack of partial refueling) makes it possible to change the core content at each refueling by using the type of fuel that is the most economically effective at the current stage of nuclear power (NP) development. In the nearest future use of mastered oxide uranium fuel and operating in the opened fuel cycle with postponed reprocessing will be the most economically effective. Changeover to the mixed uranium-plutonium fuel and closed nuclear fuel cycle (NFC) will be economically effective in an event of increase of natural uranium costs when the expenditures for construction of the enterprises on reprocessing the spent nuclear fuel (SNF), re-fabrication of new fuel with plutonium and their operating are less than the corresponding costs of natural uranium, its enrichment costs, the costs of manufacturing fresh uranium fuel and long temporary storage of SNF. At this, it is possible to use both MOX fuel with weapon or reactor plutonium and mixed nitride fuel in case its usage is more profitable. As fast reactors (FR) using uranium fuel and operating in the opened NFC consume much more natural uranium in comparison with thermal reactors (TR), and at the expected high paces of NP development the cheap resources of natural uranium will be exhausted prior to the middle of the century that will cause increase in the uranium cost, the period of FRs operating in the opened NFC must be maximally reduced. However, it should be mentioned that it is difficult to forecast reliably the date when because of the increased cost of natural uranium the NP will lose its competitiveness with electric power using fossil fuel. This is conditioned by the fact that the cost of the NPP produced electricity is less sensitive to the cost of natural uranium in

  20. The effect of thermal cycling on the structure and properties of a Co, Cr, Ni-TaC directionally solidified eutectic composite

    NASA Technical Reports Server (NTRS)

    Dunlevey, F. M.; Wallace, J. F.

    1973-01-01

    The effect of thermal cycling on the structure and properties of a cobalt, chromium, nickel, tantalum carbide directionally solidified eutectic composite is reported. It was determined that the stress rupture properties of the alloy were decreased by the thermal cycling. The loss in stress rupture properties varied with the number of cycles with the loss in properties after about 200 cycles being relatively high. The formation of serrations and the resulting changes in the mechanical properties of the material are discussed.

  1. Fast Thorium Molten Salt Reactors Started with Plutonium

    SciTech Connect

    Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Brissot, R.; Liatard, E.; Meplan, O.; Nuttin, A.

    2006-07-01

    One of the pending questions concerning Molten Salt Reactors based on the {sup 232}Th/{sup 233}U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since {sup 233}U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing {sup 233}U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce {sup 233}U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/{sup 233}U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into {sup 233}U. A particular reactor configuration is used, called 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum, allowing Plutonium to be used as fissile matter. The conversion capacities of such Molten Salt Reactors are excellent. For Molten Salt Reactors only started with Plutonium, the assets of the Thorium fuel cycle turn out to be quickly recovered and the reactor's characteristics turn out to be equivalent to Molten Salt Reactors operated with {sup 233}U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with {sup 233}U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)

  2. Direct measurements of the coordination of lever arm swing and the catalytic cycle in myosin V.

    PubMed

    Trivedi, Darshan V; Muretta, Joseph M; Swenson, Anja M; Davis, Jonathon P; Thomas, David D; Yengo, Christopher M

    2015-11-24

    Myosins use a conserved structural mechanism to convert the energy from ATP hydrolysis into a large swing of the force-generating lever arm. The precise timing of the lever arm movement with respect to the steps in the actomyosin ATPase cycle has not been determined. We have developed a FRET system in myosin V that uses three donor-acceptor pairs to examine the kinetics of lever arm swing during the recovery and power stroke phases of the ATPase cycle. During the recovery stroke the lever arm swing is tightly coupled to priming the active site for ATP hydrolysis. The lever arm swing during the power stroke occurs in two steps, a fast step that occurs before phosphate release and a slow step that occurs before ADP release. Time-resolved FRET demonstrates a 20-Å change in distance between the pre- and postpower stroke states and shows that the lever arm is more dynamic in the postpower stroke state. Our results suggest myosin binding to actin in the ADP.Pi complex triggers a rapid power stroke that gates the release of phosphate, whereas a second slower power stroke may be important for mediating strain sensitivity.

  3. Direct measurements of the coordination of lever arm swing and the catalytic cycle in myosin V

    PubMed Central

    Trivedi, Darshan V.; Muretta, Joseph M.; Swenson, Anja M.; Davis, Jonathon P.; Thomas, David D.; Yengo, Christopher M.

    2015-01-01

    Myosins use a conserved structural mechanism to convert the energy from ATP hydrolysis into a large swing of the force-generating lever arm. The precise timing of the lever arm movement with respect to the steps in the actomyosin ATPase cycle has not been determined. We have developed a FRET system in myosin V that uses three donor–acceptor pairs to examine the kinetics of lever arm swing during the recovery and power stroke phases of the ATPase cycle. During the recovery stroke the lever arm swing is tightly coupled to priming the active site for ATP hydrolysis. The lever arm swing during the power stroke occurs in two steps, a fast step that occurs before phosphate release and a slow step that occurs before ADP release. Time-resolved FRET demonstrates a 20-Å change in distance between the pre- and postpower stroke states and shows that the lever arm is more dynamic in the postpower stroke state. Our results suggest myosin binding to actin in the ADP.Pi complex triggers a rapid power stroke that gates the release of phosphate, whereas a second slower power stroke may be important for mediating strain sensitivity. PMID:26553992

  4. Influence of the cycle length on the production of PHA and polyglucose from glycerol by bacterial enrichments in sequencing batch reactors.

    PubMed

    Moralejo-Gárate, Helena; Palmeiro-Sánchez, Tania; Kleerebezem, Robbert; Mosquera-Corral, Anuska; Campos, José Luis; van Loosdrecht, Mark C M

    2013-12-01

    PHA, a naturally occurring biopolymer produced by a wide range of microorganisms, is known for its applications as bioplastic. In recent years the use of agro-industrial wastewater as substrate for PHA production by bacterial enrichments has attracted considerable research attention. Crude glycerol as generated during biodiesel production is a waste stream that due to its high organic matter content and low price could be an interesting substrate for PHA production. Previously we have demonstrated that when glycerol is used as substrate in a feast-famine regime, PHA and polyglucose are simultaneously produced as storage polymers. The work described in this paper aimed at understanding the effect of the cycle length on the bacterial enrichment process with emphasis on the distribution of glycerol towards PHA and polyglucose. Two sequencing batch reactors where operated with the same hydraulic and biomass retention time. A short cycle length (6 h) favored polyglucose production over PHA, whereas at long cycle length (24 h) PHA was more favored. In both communities the same microorganism appeared dominating, suggesting a metabolic rather than a microbial competition response. Moreover, the presence of ammonium during polymer accumulation did not influence the maximum amount of PHA that was attained.

  5. High-temperature, high-pressure testing of zinc titanate in a bench-scale fluidized-bed reactor for 100 cycles

    SciTech Connect

    Gupta, R.P.; Gangwal, S.K.

    1993-06-01

    Integrated gasification combined cycle (IGCC) power plants are being advanced worldwide to produce electricity from coal owing to their potential for superior environmental performance, economics, and efficiency in comparison to conventional coal-based power plants. A key component of these plants is a hot-gas desulfurization system employing efficient regenerable mixed-metal oxide sorbents. Leading sorbent candidates include zinc ferrite and zinc titanate. These sorbents can remove hydrogen sulfide (H{sub 2}S) in the fuel gas down to very low levels (typically <20 ppmv) at 500 to 750{degree}C and can be readily regenerated for multicycle operation with air. To this end, the Research Triangle Institute (RTI) has formulated and tested a series of zinc titanate sorbents in a high-temperature, high- pressure HTHP fluidized-bed bench-scale reactor. Multicycle HTHP bench-scale testing of these sorbents under a variety of conditions culminated in the development of a ZT-4 sorbent that exhibited the best overall performance in terms of chemical reactivity, sulfur capacity, regenerability, structural properties, and attrition resistance. Following this parametric study, a life-cycle test consisting of 100 sulfidation-regeneration cycles was carried out with ZT-4 in the bench unit.

  6. Influence of the cycle length on the production of PHA and polyglucose from glycerol by bacterial enrichments in sequencing batch reactors.

    PubMed

    Moralejo-Gárate, Helena; Palmeiro-Sánchez, Tania; Kleerebezem, Robbert; Mosquera-Corral, Anuska; Campos, José Luis; van Loosdrecht, Mark C M

    2013-12-01

    PHA, a naturally occurring biopolymer produced by a wide range of microorganisms, is known for its applications as bioplastic. In recent years the use of agro-industrial wastewater as substrate for PHA production by bacterial enrichments has attracted considerable research attention. Crude glycerol as generated during biodiesel production is a waste stream that due to its high organic matter content and low price could be an interesting substrate for PHA production. Previously we have demonstrated that when glycerol is used as substrate in a feast-famine regime, PHA and polyglucose are simultaneously produced as storage polymers. The work described in this paper aimed at understanding the effect of the cycle length on the bacterial enrichment process with emphasis on the distribution of glycerol towards PHA and polyglucose. Two sequencing batch reactors where operated with the same hydraulic and biomass retention time. A short cycle length (6 h) favored polyglucose production over PHA, whereas at long cycle length (24 h) PHA was more favored. In both communities the same microorganism appeared dominating, suggesting a metabolic rather than a microbial competition response. Moreover, the presence of ammonium during polymer accumulation did not influence the maximum amount of PHA that was attained. PMID:23835920

  7. Unprecedented inhibition of tubulin polymerization directed by gold nanoparticles inducing cell cycle arrest and apoptosis

    NASA Astrophysics Data System (ADS)

    Choudhury, Diptiman; Xavier, Paulrajpillai Lourdu; Chaudhari, Kamalesh; John, Robin; Dasgupta, Anjan Kumar; Pradeep, Thalappil; Chakrabarti, Gopal

    2013-05-01

    The effect of gold nanoparticles (AuNPs) on the polymerization of tubulin has not been examined till now. We report that interaction of weakly protected AuNPs with microtubules (MTs) could cause inhibition of polymerization and aggregation in the cell free system. We estimate that single citrate capped AuNPs could cause aggregation of ~105 tubulin heterodimers. Investigation of the nature of inhibition of polymerization and aggregation by Raman and Fourier transform-infrared (FTIR) spectroscopies indicated partial conformational changes of tubulin and microtubules, thus revealing that AuNP-induced conformational change is the driving force behind the observed phenomenon. Cell culture experiments were carried out to check whether this can happen inside a cell. Dark field microscopy (DFM) combined with hyperspectral imaging (HSI) along with flow cytometric (FC) and confocal laser scanning microscopic (CLSM) analyses suggested that AuNPs entered the cell, caused aggregation of the MTs of A549 cells, leading to cell cycle arrest at the G0/G1 phase and concomitant apoptosis. Further, Western blot analysis indicated the upregulation of mitochondrial apoptosis proteins such as Bax and p53, down regulation of Bcl-2 and cleavage of poly(ADP-ribose) polymerase (PARP) confirming mitochondrial apoptosis. Western blot run after cold-depolymerization revealed an increase in the aggregated insoluble intracellular tubulin while the control and actin did not aggregate, suggesting microtubule damage induced cell cycle arrest and apoptosis. The observed polymerization inhibition and cytotoxic effects were dependent on the size and concentration of the AuNPs used and also on the incubation time. As microtubules are important cellular structures and target for anti-cancer drugs, this first observation of nanoparticles-induced protein's conformational change-based aggregation of the tubulin-MT system is of high importance, and would be useful in the understanding of cancer therapeutics

  8. ASSESSMENT OF POSSIBLE CYCLE LENGTHS FOR FULLY-CERAMIC MICRO-ENCAPSULATED FUEL-BASED LIGHT WATER REACTOR CONCEPTS

    SciTech Connect

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal Pasamehmetoglu; Francesco Venneri

    2012-04-01

    The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO{sub 2} results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rates of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO{sub 2}-loaded TRISO particles-based fuel with enrichment at or below 20 w/o.

  9. Summary of the radiological assessment of the fuel cycle for a thorium-uranium carbide-fueled fast breeder reactor

    SciTech Connect

    Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Meyer, H.R.; Morse, L.E.; Till, J.E.; Yalcintas, M.G.

    1980-01-01

    A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

  10. Direct observations of the full Dungey convection cycle in the polar ionosphere for southward interplanetary magnetic field conditions

    NASA Astrophysics Data System (ADS)

    Zhang, Q.-H.; Lockwood, M.; Foster, J. C.; Zhang, S.-R.; Zhang, B.-C.; McCrea, I. W.; Moen, J.; Lester, M.; Ruohoniemi, J. M.

    2015-06-01

    Tracking the formation and full evolution of polar cap ionization patches in the polar ionosphere, we directly observe the full Dungey convection cycle for southward interplanetary magnetic field (IMF) conditions. This enables us to study how the Dungey cycle influences the patches' evolution. The patches were initially segmented from the dayside storm enhanced density plume at the equatorward edge of the cusp, by the expansion and contraction of the polar cap boundary due to pulsed dayside magnetopause reconnection, as indicated by in situ Time History of Events and Macroscale Interactions during Substorms (THEMIS) observations. Convection led to the patches entering the polar cap and being transported antisunward, while being continuously monitored by the globally distributed arrays of GPS receivers and Super Dual Auroral Radar Network radars. Changes in convection over time resulted in the patches following a range of trajectories, each of which differed somewhat from the classical twin-cell convection streamlines. Pulsed nightside reconnection, occurring as part of the magnetospheric substorm cycle, modulated the exit of the patches from the polar cap, as confirmed by coordinated observations of the magnetometer at Tromsø and European Incoherent Scatter Tromsø UHF radar. After exiting the polar cap, the patches broke up into a number of plasma blobs and returned sunward in the auroral return flow of the dawn and/or dusk convection cell. The full circulation time was about 3 h.

  11. Optically heated ultra-fast-cycling gas chromatography module for separation of direct sampling and online monitoring applications.

    PubMed

    Fischer, Michael; Wohlfahrt, Sebastian; Varga, Janos; Matuschek, Georg; Saraji-Bozorgzad, Mohammad R; Denner, Thomas; Walte, Andreas; Zimmermann, Ralf

    2015-09-01

    This work describes an ultrafast-cycling gas chromatography module (fast-GC module) for direct-sampling gas chromatography/mass spectrometry (GC-MS). The sample can be introduced into the fast-GC module using a common GC injector or any GC × GC modulator. The new fast-GC module offers the possibility to conduct a complete temperature cycle within 30 s. Its thermal mass is minimized by using a specially developed home-built fused silica capillary column stack and a halogen lamp for heat generation, both placed inside a gold-coated quartz glass cylinder. A high airflow blower enables rapid cooling. The new device is highly flexible concerning the used separation column, the applied temperature program, and the integration into existing systems. An application of the fast-GC module is shown in this work by thermal analysis coupled to gas chromatography-mass spectrometry (TA-GC-MS). The continuously evolving gases of the TA are modulated by a liquid CO2 modulator. Because of the rapid cycling of the fast-GC module, it is possible to obtain the best separation while maintaining the online character of the TA. Restrictions in separation and retention time shifting, known from isothermal and normal ramped fast-GC systems, are overcome. PMID:26226397

  12. Unprecedented inhibition of tubulin polymerization directed by gold nanoparticles inducing cell cycle arrest and apoptosis

    NASA Astrophysics Data System (ADS)

    Choudhury, Diptiman; Xavier, Paulrajpillai Lourdu; Chaudhari, Kamalesh; John, Robin; Dasgupta, Anjan Kumar; Pradeep, Thalappil; Chakrabarti, Gopal

    2013-05-01

    The effect of gold nanoparticles (AuNPs) on the polymerization of tubulin has not been examined till now. We report that interaction of weakly protected AuNPs with microtubules (MTs) could cause inhibition of polymerization and aggregation in the cell free system. We estimate that single citrate capped AuNPs could cause aggregation of ~105 tubulin heterodimers. Investigation of the nature of inhibition of polymerization and aggregation by Raman and Fourier transform-infrared (FTIR) spectroscopies indicated partial conformational changes of tubulin and microtubules, thus revealing that AuNP-induced conformational change is the driving force behind the observed phenomenon. Cell culture experiments were carried out to check whether this can happen inside a cell. Dark field microscopy (DFM) combined with hyperspectral imaging (HSI) along with flow cytometric (FC) and confocal laser scanning microscopic (CLSM) analyses suggested that AuNPs entered the cell, caused aggregation of the MTs of A549 cells, leading to cell cycle arrest at the G0/G1 phase and concomitant apoptosis. Further, Western blot analysis indicated the upregulation of mitochondrial apoptosis proteins such as Bax and p53, down regulation of Bcl-2 and cleavage of poly(ADP-ribose) polymerase (PARP) confirming mitochondrial apoptosis. Western blot run after cold-depolymerization revealed an increase in the aggregated insoluble intracellular tubulin while the control and actin did not aggregate, suggesting microtubule damage induced cell cycle arrest and apoptosis. The observed polymerization inhibition and cytotoxic effects were dependent on the size and concentration of the AuNPs used and also on the incubation time. As microtubules are important cellular structures and target for anti-cancer drugs, this first observation of nanoparticles-induced protein's conformational change-based aggregation of the tubulin-MT system is of high importance, and would be useful in the understanding of cancer therapeutics

  13. High efficiency direct fuel cell hybrid power cycle for near term application

    SciTech Connect

    Steinfeld, G.; Maru, H.C.; Sanderson, R.A.

    1996-12-31

    Direct carbonate fuel cells being developed by Energy Research Corporation can generate power at an efficiency approaching 60% LHV. This unique fuel cell technology can consume natural gas and other hydrocarbon based fuels directly without requiring an external reformer, thus providing a simpler and inherently efficient power generation system. A 2 MW power plant demonstration of this technology has been initiated at an installation in the city of Santa Clara in California. A 2.85 MW commercial configuration shown in Figure 1 is presently being developed. The complete plant includes the carbonate fuel cell modules, an inverter, transformer and switchgear, a heat recovery unit and supporting instrument air and water treatment systems. The emission levels for this 2.85 MW plant are projected to be orders of magnitude below existing or proposed standards. The 30 year levelized cost of electricity, without inflation, is projected to be approximately 5{cents}/kW-h assuming capital cost for the carbonate fuel cell system of $1000/kW.

  14. Short contact time direct coal liquefaction using a novel batch reactor. Progress report, January 1, 1994--May 15, 1994

    SciTech Connect

    Klein, M.T.; Calkins, W.H.

    1994-05-31

    The objective for this research is to optimize the design and operation of the bench scale batch reactor (SCTBR) for coal liquefaction at short contact times (0.01 to 10 minutes or longer). This reactor is simple enough and low enough in cost to serve as a suitable replacement for the traditional tubing-bomb reactors for many coal liquefaction and other high-pressure, high-temperature reaction studies. The liquefaction of selected Argonne Premium coals and the role of organic oxygen components of the coal and their reaction pathways at very low conversions are being investigated.

  15. Conceptual design study FY 1981: synfuels from fusion - using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

    SciTech Connect

    Krikorian, O.H.

    1982-02-09

    This report represents the second year's effort of a scoping and conceptual design study being conducted for the express purpose of evaluating the engineering potential of producing hydrogen by thermochemical cycles using a tandem mirror fusion driver. The hydrogen thus produced may then be used as a feedstock to produce fuels such as methane, methanol, or gasoline. The main objective of this second year's study has been to obtain some approximate cost figures for hydrogen production through a conceptual design study.

  16. Biomass Direct Liquefaction Options. TechnoEconomic and Life Cycle Assessment

    SciTech Connect

    Tews, Iva J.; Zhu, Yunhua; Drennan, Corinne; Elliott, Douglas C.; Snowden-Swan, Lesley J.; Onarheim, Kristin; Solantausta, Yrjo; Beckman, David

    2014-07-31

    The purpose of this work was to assess the competitiveness of two biomass to transportation fuel processing routes, which were under development in Finland, the U.S. and elsewhere. Concepts included fast pyrolysis (FP), and hydrothermal liquefaction (HTL), both followed by hydrodeoxygenation, and final product refining. This work was carried out as a collaboration between VTT (Finland), and PNNL (USA). The public funding agents for the work were Tekes in Finland and the Bioenergy Technologies Office of the U.S. Department of Energy. The effort was proposed as an update of the earlier comparative technoeconomic assessment performed by the IEA Bioenergy Direct Biomass Liquefaction Task in the 1980s. New developments in HTL and the upgrading of the HTL biocrude product triggered the interest in reinvestigating this comparison of these biomass liquefaction processes. In addition, developments in FP bio-oil upgrading had provided additional definition of this process option, which could provide an interesting comparison.

  17. Short contact time direct coal liquefaction using a novel batch reactor. Progress report, September 27, 1993--December 31, 1993

    SciTech Connect

    Klein, M.T.; Calkins, W.H.

    1994-01-19

    The objective for this research is to optimize the design and operation of the bench scale batch reactor (STBR) for coat liquefaction at short contact times (0.01 to 10 minutes). This reactor is simple and low enough in cost to serve as a suitable replacement for the traditional tubing-bomb reactors for coal liquefaction and other high-pressure, high-temperature reaction studies. The details of the reactor system are shown in Figure 2. The heating bath used is a Techne IFB-52 industrial fluidized sand bath, which maintains a reaction temperature of {plus_minus}2{degrees}C. The 30 cm{sup 3} reactor is capable of containing up to 17 MPa (2500 psi) pressure at temperatures up to 550{degrees}C. The tubing used for preheater and precooler was 1/4in. 316 stainless steel with wall thickness of 0.035in. The lengths of the preheater and precooler are selected based on the particular process being studied. Since a gas (e.g. hydrogen or nitrogen) is bubbled through the reaction mixture under pressure and out through a letdown valve, a small water cooled condenser above the reactor before the let-down valve is added to avoid loss of solvent or other low boiling components. Coal liquefaction runs are made by preparing slurries of coal in reagent grade tetralin. Various ratios of tetralin to coal are used, and in some cases, a catalyst such as Ni/Mo on alumina is added.

  18. Effect of transient thermal cycles in a supercritical water-cooled reactor on the microstructure and properties of ferritic martensitic steels

    NASA Astrophysics Data System (ADS)

    Totemeier, T. C.; Clark, D. E.

    2006-09-01

    Microstructural and mechanical property changes in modified 9Cr-1Mo and HCM12A ferritic-martensitic steels resulting from short-duration thermal transients that occur during loss of feedwater flow events in a supercritical water reactor (SCWR) were studied. Specimen blanks were exposed to reference transients with 810 and 840 °C maximum temperatures using a thermal cycle simulator, and the subsequent microstructure, hardness, and creep-rupture strength were evaluated. Exposure to five consecutive cycles at either temperature resulted in no significant changes - only very slight indications of overtempering. Subsequent study of a wider variety of transient conditions showed that significant ferrite-to-austenite transformation occurred during thermal transients whose maximum temperature exceeded 860 °C, or during transients with holds exceeding 10 s at 840 °C maximum temperature. The subsequent presence of untempered martensite in the microstructure, coupled with severe overtempering, resulted in an order of magnitude decrease in creep-rupture strength at 600 °C. The findings were consistent with measured Ac1 temperatures for the two steels and the dependence of Ac1 on heating rate.

  19. Direct identification of all oncogenic mutants in KRAS exon 1 by cycling temperature capillary electrophoresis.

    PubMed

    Bjørheim, Jens; Gaudernack, Gustav; Giercksky, Karl-Erik; Ekstrøm, Per O

    2003-01-01

    Over the past few decades, advances in genetics and molecular biology have revolutionized our understanding of cancer initiation and progression. Molecular progression models outlining genetic events have been developed for many solid tumors, including colon cancer. Previous reports in the literature have shown a relationship between different KRAS mutations and prognosis and response to medical treatment in colon cancer patients. Furthermore, the presence of a mutated KRAS has been correlated with different clinicopathological variables including age and gender of patients and tumor location. To our knowledge, few institutions screen for KRAS mutations on regular basis in colon cancer patients despite such evidence that knowledge of KRAS exon 1 status is informative. Here, we report on a mutation analysis method adapted to a 96-capillary electrophoresis instrument that allows identification of all 12 oncogenic mutations in KRAS exon 1 under denaturing conditions. To determine the optimal parameters, a series of DNA constructs generated by site-directed mutagenesis was analyzed and the migration times of all mutant peaks were measured. A classification tree was then made based on the differences in migration time between the mutants and an internal standard. A randomized series of 500 samples constructed with mutagenesis as well as 60 blind samples from sporadic colon carcinomas was analyzed to test the method. No wild-type samples were scored as mutants and all mutants were correctly identified. Post polymerase chain reaction (PCR) analysis time of 96 samples was performed within 40 min. PMID:12652573

  20. Direct view on the phase evolution in individual LiFePO4 nanoparticles during Li-ion battery cycling

    PubMed Central

    Zhang, Xiaoyu; van Hulzen, Martijn; Singh, Deepak P.; Brownrigg, Alex; Wright, Jonathan P.; van Dijk, Niels H.; Wagemaker, Marnix

    2015-01-01

    Phase transitions in Li-ion electrode materials during (dis)charge are decisive for battery performance, limiting high-rate capabilities and playing a crucial role in the cycle life of Li-ion batteries. However, the difficulty to probe the phase nucleation and growth in individual grains is hindering fundamental understanding and progress. Here we use synchrotron microbeam diffraction to disclose the cycling rate-dependent phase transition mechanism within individual particles of LiFePO4, a key Li-ion electrode material. At low (dis)charge rates well-defined nanometer thin plate-shaped domains co-exist and transform much slower and concurrent as compared with the commonly assumed mosaic transformation mechanism. As the (dis)charge rate increases phase boundaries become diffuse speeding up the transformation rates of individual grains. Direct observation of the transformation of individual grains reveals that local current densities significantly differ from what has previously been assumed, giving new insights in the working of Li-ion battery electrodes and their potential improvements. PMID:26395323

  1. THE FIRST GROUND LEVEL ENHANCEMENT EVENT OF SOLAR CYCLE 24: DIRECT OBSERVATION OF SHOCK FORMATION AND PARTICLE RELEASE HEIGHTS

    SciTech Connect

    Gopalswamy, N.; Xie, H.; Akiyama, S.; Yashiro, S.; Davila, J. M.; Usoskin, I. G.

    2013-03-10

    We report on the 2012 May 17 ground level enhancement (GLE) event, which is the first of its kind in solar cycle 24. This is the first GLE event to be fully observed close to the surface by the Solar Terrestrial Relations Observatory (STEREO) mission. We determine the coronal mass ejection (CME) height at the start of the associated metric type II radio burst (i.e., shock formation height) as 1.38 Rs (from the Sun center). The CME height at the time of GLE particle release was directly measured from a STEREO image as 2.32 Rs, which agrees well with the estimation from CME kinematics. These heights are consistent with those obtained for cycle-23 GLEs using back-extrapolation. By contrasting the 2012 May 17 GLE with six other non-GLE eruptions from well-connected regions with similar or larger flare sizes and CME speeds, we find that the latitudinal distance from the ecliptic is rather large for the non-GLE events due to a combination of non-radial CME motion and unfavorable solar B0 angle, making the connectivity to Earth poorer. We also find that the coronal environment may play a role in deciding the shock strength.

  2. Direct-contact condensers for open-cycle OTEC applications: Model validation with fresh water experiments for structured packings

    NASA Astrophysics Data System (ADS)

    Bharathan, D.; Parsons, B. K.; Althof, J. A.

    1988-10-01

    The objective of the reported work was to develop analytical methods for evaluating the design and performance of advanced high-performance heat exchangers for use in open-cycle thermal energy conversion (OC-OTEC) systems. This report describes the progress made on validating a one-dimensional, steady-state analytical computer of fresh water experiments. The condenser model represents the state of the art in direct-contact heat exchange for condensation for OC-OTEC applications. This is expected to provide a basis for optimizing OC-OTEC plant configurations. Using the model, we examined two condenser geometries, a cocurrent and a countercurrent configuration. This report provides detailed validation results for important condenser parameters for cocurrent and countercurrent flows. Based on the comparisons and uncertainty overlap between the experimental data and predictions, the model is shown to predict critical condenser performance parameters with an uncertainty acceptable for general engineering design and performance evaluations.

  3. Direct-contact condensers for open-cycle OTEC applications: Model validation with fresh water experiments for structured packings

    SciTech Connect

    Bharathan, D.; Parsons, B.K.; Althof, J.A.

    1988-10-01

    The objective of the reported work was to develop analytical methods for evaluating the design and performance of advanced high-performance heat exchangers for use in open-cycle thermal energy conversion (OC-OTEC) systems. This report describes the progress made on validating a one-dimensional, steady-state analytical computer of fresh water experiments. The condenser model represents the state of the art in direct-contact heat exchange for condensation for OC-OTEC applications. This is expected to provide a basis for optimizing OC-OTEC plant configurations. Using the model, we examined two condenser geometries, a cocurrent and a countercurrent configuration. This report provides detailed validation results for important condenser parameters for cocurrent and countercurrent flows. Based on the comparisons and uncertainty overlap between the experimental data and predictions, the model is shown to predict critical condenser performance parameters with an uncertainty acceptable for general engineering design and performance evaluations. 33 refs., 69 figs., 38 tabs.

  4. Carrier-envelope phase dependence of the directional fragmentation and hydrogen migration in toluene in few-cycle laser fields

    PubMed Central

    Li, Hui; Kling, Nora G.; Förg, Benjamin; Stierle, Johannes; Kessel, Alexander; Trushin, Sergei A.; Kling, Matthias F.; Kaziannis, Spyros

    2016-01-01

    The dissociative ionization of toluene initiated by a few-cycle laser pulse as a function of the carrier envelope phase (CEP) is investigated using single-shot velocity map imaging. Several ionic fragments, CH3+, H2+, and H3+, originating from multiply charged toluene ions present a CEP-dependent directional emission. The formation of H2+ and H3+ involves breaking C-H bonds and forming new bonds between the hydrogen atoms within the transient structure of the multiply charged precursor. We observe appreciable intensity-dependent CEP-offsets. The experimental data are interpreted with a mechanism that involves laser-induced coupling of vibrational states, which has been found to play a role in the CEP-control of molecular processes in hydrocarbon molecules, and appears to be of general importance for such complex molecules. PMID:26958589

  5. Ozone generation by negative direct current corona discharges in dry air fed coaxial wire-cylinder reactors

    SciTech Connect

    Yehia, Ashraf; Mizuno, Akira

    2013-05-14

    An analytical study was made in this paper for calculating the ozone generation by negative dc corona discharges. The corona discharges were formed in a coaxial wire-cylinder reactor. The reactor was fed by dry air flowing with constant rates at atmospheric pressure and room temperature, and stressed by a negative dc voltage. The current-voltage characteristics of the negative dc corona discharges formed inside the reactor were measured in parallel with concentration of the generated ozone under different operating conditions. An empirical equation was derived from the experimental results for calculating the ozone concentration generated inside the reactor. The results, that have been recalculated by using the derived equation, have agreed with the experimental results over the whole range of the investigated parameters, except in the saturation range for the ozone concentration. Therefore, the derived equation represents a suitable criterion for expecting the ozone concentration generated by negative dc corona discharges in dry air fed coaxial wire-cylinder reactors under any operating conditions in range of the investigated parameters.

  6. 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.

  7. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities, Sections 15-19

    SciTech Connect

    Schneider, K.J.

    1982-09-01

    Information is presented under the following section headings: fuel reprocessing; spent fuel and high-level and transuranic waste storage; spent fuel and high-level and transuranic waste disposal; low-level and intermediate-level waste disposal; and, transportation of radioactive materials in the nuclear fuel cycle. In each of the first three sections a description is given on the mainline process, effluent processing and waste management systems, plant layout, and alternative process schemes. Safety information and a summary are also included in each. The section on transport of radioactive materials includes information on the transportation of uranium ore, uranium ore concentrate, UF/sub 6/, PuO/sub 2/ powder, unirradiated uranium and mixed-oxide fuel assemblies, spent fuel, solidified high-level waste, contact-handled transuranic waste, remote-handled transuranic waste, and low and intermediate level nontransuranic waste. A glossary is included. (JGB)

  8. Notch stimulates growth by direct regulation of genes involved in the control of glycolysis and the tricarboxylic acid cycle

    PubMed Central

    Slaninova, Vera; Krafcikova, Michaela; Perez-Gomez, Raquel; Steffal, Pavel; Trantirek, Lukas; Bray, Sarah J.

    2016-01-01

    Glycolytic shift is a characteristic feature of rapidly proliferating cells, such as cells during development and during immune response or cancer cells, as well as of stem cells. It results in increased glycolysis uncoupled from mitochondrial respiration, also known as the Warburg effect. Notch signalling is active in contexts where cells undergo glycolytic shift. We decided to test whether metabolic genes are direct transcriptional targets of Notch signalling and whether upregulation of metabolic genes can help Notch to induce tissue growth under physiological conditions and in conditions of Notch-induced hyperplasia. We show that genes mediating cellular metabolic changes towards the Warburg effect are direct transcriptional targets of Notch signalling. They include genes encoding proteins involved in glucose uptake, glycolysis, lactate to pyruvate conversion and repression of the tricarboxylic acid cycle. The direct transcriptional upregulation of metabolic genes is PI3K/Akt independent and occurs not only in cells with overactivated Notch but also in cells with endogenous levels of Notch signalling and in vivo. Even a short pulse of Notch activity is able to elicit long-lasting metabolic changes resembling the Warburg effect. Loss of Notch signalling in Drosophila wing discs as well as in human microvascular cells leads to downregulation of glycolytic genes. Notch-driven tissue overgrowth can be rescued by downregulation of genes for glucose metabolism. Notch activity is able to support growth of wing during nutrient-deprivation conditions, independent of the growth of the rest of the body. Notch is active in situations that involve metabolic reprogramming, and the direct regulation of metabolic genes may be a common mechanism that helps Notch to exert its effects in target tissues. PMID:26887408

  9. A green approach to ethyl acetate: quantitative conversion of ethanol through direct dehydrogenation in a Pd-Ag membrane reactor.

    PubMed

    Zeng, Gaofeng; Chen, Tao; He, Lipeng; Pinnau, Ingo; Lai, Zhiping; Huang, Kuo-Wei

    2012-12-01

    Pincers do the trick: The conversion of ethanol to ethyl acetate and hydrogen was achieved using a pincer-Ru catalyst in a Pd-Ag membrane reactor. Near quantitative conversions and yields could be achieved without the need for acid or base promoters or hydrogen acceptors (see scheme). PMID:23136053

  10. Potential of direct metal deposition technology for manufacturing thick functionally graded coatings and parts for reactors components

    NASA Astrophysics Data System (ADS)

    Thivillon, L.; Bertrand, Ph.; Laget, B.; Smurov, I.

    2009-03-01

    Direct metal deposition (DMD) is an automated 3D deposition process arising from laser cladding technology with co-axial powder injection to refine or refurbish parts. Recently DMD has been extended to manufacture large-size near-net-shape components. When applied for manufacturing new parts (or their refinement), DMD can provide tailored thermal properties, high corrosion resistance, tailored tribology, multifunctional performance and cost savings due to smart material combinations. In repair (refurbishment) operations, DMD can be applied for parts with a wide variety of geometries and sizes. In contrast to the current tool repair techniques such as tungsten inert gas (TIG), metal inert gas (MIG) and plasma welding, laser cladding technology by DMD offers a well-controlled heat-treated zone due to the high energy density of the laser beam. In addition, this technology may be used for preventative maintenance and design changes/up-grading. One of the advantages of DMD is the possibility to build functionally graded coatings (from 1 mm thickness and higher) and 3D multi-material objects (for example, 100 mm-sized monolithic rectangular) in a single-step manufacturing cycle by using up to 4-channel powder feeder. Approved materials are: Fe (including stainless steel), Ni and Co alloys, (Cu,Ni 10%), WC compounds, TiC compounds. The developed coatings/parts are characterized by low porosity (<1%), fine microstructure, and their microhardness is close to the benchmark value of wrought alloys after thermal treatment (Co-based alloy Stellite, Inox 316L, stainless steel 17-4PH). The intended applications concern cooling elements with complex geometry, friction joints under high temperature and load, light-weight mechanical support structures, hermetic joints, tubes with complex geometry, and tailored inside and outside surface properties, etc.

  11. Investigation of the Performance of D2O-Cooled High-Conversion Reactors for Fuel Cycle Calculations

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    Th-U fueled cores are identified by evaluating the sensitivity coefficients of keff, mass balance, and void coefficient. The effect of advanced iron alloy cladding (i.e., FeCrAl) on the performance of Pu conversion in MOX fueled cores is studied instead of using standard stainless-steel cladding. Variations in clad thickness and coolant-to-fuel volume ratio are also exercised. The use of FeCrAl instead of SS as a cladding alloy reduces the required Pu enrichment and improves the Pu conversion rate primarily due to the absence of nickel in the cladding alloy that results in the reduction of the neutron absorption. Also the difference in void coefficients between SS and FeCrAl alloys is nearly 500 pcm over the entire burnup range. The report also shows sensitivity and uncertainty analyses in order to characterize D2O cooled HCPWRs from different aspects. The uncertainties of integral parameters (keff and void coefficient) for selected reactor cores are evaluated at different burnup points in order to find similarities and trends respect to D2O-HCPWR.

  12. Safe new reactor for radionuclide production

    SciTech Connect

    Gray, P.L.

    1995-02-15

    In late 1995, DOE is schedule to announce a new tritium production unit. Near the end of the last NPR (New Production Reactors) program, work was directed towards eliminating risks in current designs and reducing effects of accidents. In the Heavy Water Reactor Program at Savannah River, the coolant was changed from heavy to light water. An alternative, passively safe concept uses a heavy-water-filled, zircaloy reactor calandria near the bottom of a swimming pool; the calandria is supported on a light-water-coolant inlet plenum and has upflow through assemblies in the calandria tubes. The reactor concept eliminates or reduces significantly most design basis and severe accidents that plague other deigns. The proven, current SRS tritium cycle remains intact; production within the US of medical isotopes such as Mo-99 would also be possible.

  13. Effect of intermittent aeration cycle on nutrient removal and microbial community in a fluidized bed reactor-membrane bioreactor combo system.

    PubMed

    Guadie, Awoke; Xia, Siqing; Zhang, Zhiqiang; Zeleke, Jemaneh; Guo, Wenshan; Ngo, Huu Hao; Hermanowicz, Slawomir W

    2014-03-01

    Effect of intermittent aeration cycle (IAC=15/45-60/60min) on nutrient removal and microbial community structure was investigated using a novel fluidized bed reactor-membrane bioreactor (FBR-MBR) combo system. FBR alone was found more efficient for removing PO4-P (>85%) than NH4-N (<40%) and chemical oxygen demand (COD<35%). However, in the combo system, COD and NH4-N removals were almost complete (>98%). Efficient nitrification, stable mixed liquor suspended solid and reduced transmembrane pressure was also achieved. Quantitative real-time polymerase chain reaction results of total bacteria 16S rRNA gene copies per mL of mixed-liquor varied from (2.48±0.42)×10(9) initial to (2.74±0.10)×10(8), (6.27±0.16)×10(9) and (9.17±1.78)×10(9) for 15/45, 45/15 and 60/60min of IACs, respectively. The results of clone library analysis revealed that Proteobacteria (59%), Firmicutes (12%) and Bacteroidetes (11%) were the dominant bacterial group in all samples. Overall, the combo system performs optimum nutrient removal and host stable microbial communities at 45/15min of IAC. PMID:24508900

  14. Commercial-Scale Performance Predictions for High-Temperature Electrolysis Plants Coupled to Three Advanced Reactor Types

    SciTech Connect

    M. G. McKellar; J. E. O'Brien; J. S. Herring

    2007-09-01

    This report presents results of system analyses that have been developed to assess the hydrogen production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor – power-cycle combinations: a high-temperature helium cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to- hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable production rates with the high-temperature helium cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor.

  15. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  16. Continuous preparation of carbon-nanotube-supported platinum catalysts in a flow reactor directly heated by electric current

    PubMed Central

    dos Santos, Antonio Rodolfo; Kunz, Ulrich; Turek, Thomas

    2011-01-01

    Summary In this contribution we present for the first time a continuous process for the production of highly active Pt catalysts supported by carbon nanotubes by use of an electrically heated tubular reactor. The synthesized catalysts show a high degree of dispersion and narrow distributions of cluster sizes. In comparison to catalysts synthesized by the conventional oil-bath method a significantly higher electrocatalytic activity was reached, which can be attributed to the higher metal loading and smaller and more uniformly distributed Pt particles on the carbon support. Our approach introduces a simple, time-saving and cost-efficient method for fuel cell catalyst preparation in a flow reactor which could be used at a large scale. PMID:22043252

  17. Evaluation of technical feasibility of closed-cycle non-equilibrium MHD power generation with direct coal firing. Final report, Task I

    SciTech Connect

    Not Available

    1981-11-01

    Program accomplishments in a continuing effort to demonstrate the feasibility of direct coal-fired, closed-cycle MHD power generation are reported. This volume contains the following appendices: (A) user's manual for 2-dimensional MHD generator code (2DEM); (B) performance estimates for a nominal 30 MW argon segmented heater; (C) the feedwater cooled Brayton cycle; (D) application of CCMHD in an industrial cogeneration environment; (E) preliminary design for shell and tube primary heat exchanger; and (F) plant efficiency as a function of output power for open and closed cycle MHD power plants. (WHK)

  18. A microphysical interpretation of the rate-and-state friction direct effect: implications for the seismic cycle

    NASA Astrophysics Data System (ADS)

    van den Ende, Martijn; Niemeijer, André; Spiers, Christopher

    2015-04-01

    small direct effect, which can not be accounted for by dilatation alone. Localisation of deformation decreases the magnitude of the direct effect, which is in agreement with observations by Marone et al. (1990). We are in the process of developing a microphysical model to explain the observed behaviour and to allow for extrapolation to natural conditions. Our microphysically based and experimentally verified model will help gain a better understanding of the seismic cycle. References: Beeler N. M., T. E. Tullis, A. K. Kronenberg, and L. A. Reinen (2007), Instantaneous rate dependence in low temperature laboratory rock friction and rock deformation experiments, J. Geophys. Res., 112, B07310, doi:10.1029/2005JB003772 Marone C., C. B. Raleigh, and C. H. Scholz (1990), Frictional behavior and constitutive modeling of simulated fault gouge, J. Geophys. Res., 95 (B5), 7007-7025, doi:10.1029/JB095iB05p07007

  19. Fluctuation-driven directional flow in biochemical cycle: further study of electric activation of Na,K pumps.

    PubMed Central

    Xie, T D; Chen, Y; Marszalek, P; Tsong, T Y

    1997-01-01

    Directional flow of information and energies is characteristic of many types of biochemical reactions, for instance, ion transport, energy coupling during ATP synthesis, and muscle contraction. Can a fluctuating force field, or a noise, induce such a directional flux? Previous work has shown that Na,K-ATPase of human erythrocyte can absorb free energy from an externally applied random-telegraph-noise (RTN) electric field to pump Rb+ up its concentration gradient. However, the RTN field used in these experiments was constant in amplitude and would not mimic fluctuating electric fields of a cell membrane. Here we show that electric fields which fluctuate both in life time and in amplitude, and thus, better mimicking the transmembrane electric fields of a cell, can also induce Rb+ pumping by Na,K-ATPase. A Gaussian-RTN-electric field, or a field with amplitude fluctuating according to the Gaussian distribution, with varied standard deviation (sigma), induced active pumping of Rb+ in human erythrocyte, which was completely inhibited by ouabain. Increased values for sigma led to a nonmonotonic reduction in pumping efficiency. A general formula for calculating the ion transport in a biochemical cycle induced by fluctuating electric field has been derived and applied to a simple four-state electroconformational coupling (ECC) model. It was found that the calculated efficiency in the energy coupling decreased with increasing sigma value, and this effect was relatively small and monotonic, whereas experimental data were more complex: monotonic under certain sets of conditions but nonmonotonic under different sets. The agreement in general features but disagreement in some fine features suggest that there are other properties of the electric activation process for Na,K-ATPase that cannot be adequately described by the simple ECC model, and further refinement of the ECC model is required. PMID:9168026

  20. Development of pyro-processing technology for thorium-fuelled molten salt reactor

    SciTech Connect

    Uhlir, J.; Straka, M.; Szatmary, L.

    2012-07-01

    The Molten Salt Reactor (MSR) is classified as the non-classical nuclear reactor type based on the specific features coming out from the use of liquid fuel circulating in the MSR primary circuit. Other uniqueness of the reactor type is based on the fact that the primary circuit of the reactor is directly connected with the on-line reprocessing technology, necessary for keeping the reactor in operation for a long run. MSR is the only reactor system, which can be effectively operated within the {sup 232}Th- {sup 233}U fuel cycle as thorium breeder with the breeding factor significantly higher than one. The fuel cycle technologies proposed as ford the fresh thorium fuel processing as for the primary circuit fuel reprocessing are pyrochemical and mainly fluoride. Although these pyrochemical processes were never previously fully verified, the present-day development anticipates an assumption for the successful future deployment of the thorium-fuelled MSR technology. (authors)

  1. Are there statistical links between the direction of European weather systems and ENSO, the solar cycle or stratospheric aerosols?

    PubMed Central

    2016-01-01

    The Hess Brezowsky Großwetterlagen (HBGWL) European weather classification system, accumulated over a long period (more than 130 years), provides a rare opportunity to examine the impact of various factors on regional atmospheric flow. We have used these data to examine changes in the frequency (days/month) of given weather systems direction (WSD) during peak phases in the North Atlantic Oscillation (NAO), El Niño Southern Oscillation (ENSO), solar cycle (SC) and peaks in stratospheric aerosol optical depth (AOD) with superposed epoch analysis and Monte Carlo significance testing. We found highly significant responses to the NAO consistent with expectations: this signal confirmed the utility of the HBGWL data for this type of analysis and provided a benchmark of a clear response. WSD changes associated with ENSO, SC and AOD were generally within the ranges expected from random samples. When seasonal restrictions were added the results were similar, however, we found one clearly significant result: an increase in southerly flow of 2.6±0.8 days/month (p=1.9×10−4) during boreal summertime in association with El Niño. This result supports the existence of a robust teleconnection between the ENSO and European weather. PMID:26998314

  2. Characterization study and five-cycle tests in a fixed-bed reactor of titania-supported nickel oxide as oxygen carriers for the chemical-looping combustion of methane.

    PubMed

    Corbella, Beatriz M; de Diego, Luis F; García-Labiano, Francisco; Adánez, Juan; Palaciost, José M

    2005-08-01

    Recent investigations have shown that in the combustion of carbonaceous compounds CO2 and NOx emissions to the atmosphere can be substantially reduced by using a two stage chemical-looping process. In this process, the reduction stage is undertaken in a first reactor in which the framework oxygen of a reducible inorganic oxide is used, instead of the usual atmospheric oxygen, for the combustion of a carbonaceous compound, for instance, methane. The outlet gas from this reactor is mostly composed of CO2 and steam as reaction products and further separation of these two components can be carried out easily by simple condensation of steam. Then, the oxygen carrier found in a reduced state is transported to a second reactor in which carrier regeneration with air takes place at relatively low temperatures, consequently preventing the formation of thermal NOx. Afterward, the regenerated carrier is carried to the first reactor to reinitiate a new cycle and so on for a number of repetitive cycles, while the carrier is able to withstand the severe chemical and thermal stresses involved in every cycle. In this paper, the performance of titania-supported nickel oxides has been investigated in a fixed-bed reactor as oxygen carriers for chemical-looping combustion of methane. Samples with different nickel oxide contents were prepared by successive incipient wet impregnations, and their performance as oxygen carriers was investigated at 900 degrees C and atmospheric pressure in five-cycle fixed-bed reactor tests using pure methane and pure air for the respective reduction and regeneration stages. The evolution of the outlet gas composition in each stage was followed by gas chromatography, and the involved chemical, structural, and textural changes of the carrier in the reactor bed were studied by using different characterization techniques. From the study, it is deduced that the reactivity of these nickel-based oxygen carriers is in the two involved stages and almost independent

  3. Evaluation of the neutron background in an HPGe target for WIMP direct detection when using a reactor neutrino detector as a neutron veto system

    SciTech Connect

    Ji, Xiangpan; Xu, Ye Lin, Junsong; Feng, Yulong; Li, Haolin

    2013-11-15

    A direct WIMP (weakly interacting massive particle) detector with a neutron veto system is designed to better reject neutrons. The experimental configuration is studied in this paper involves 984 Ge modules placed inside a reactor-neutrino detector. The neutrino detector is used as a neutron veto device. The neutron background for the experimental design is estimated using the Geant4 simulation. The results show that the neutron background can decrease to O(0.01) events per year per tonne of high-purity germanium and it can be ignored in comparison with electron recoils.

  4. Short contact time direct coal liquefaction using a novel batch reactor. Quarterly technical progress report, September 15, 1995--January 15, 1996

    SciTech Connect

    Klein, M.T.; Calkins, W.H.; Huang, He

    1996-01-26

    The objective of this research is to optimize the design and operation of the bench scale batch reactor (SCTBR) f or coal liquefaction at short contact times (0.01 to 10 minutes or longer). Additional objectives are to study the kinetics of direct coal liquefaction particularly at short reaction times, and to investigate the role of the organic oxygen components of coal and their reaction pathways during liquefaction. Many of those objectives have already been achieved and others are still in progress. This quarterly report covers further progress toward those objectives.

  5. Plant heat cycles, vessel internal arrangement, and auxiliary systems. Volume five

    SciTech Connect

    Not Available

    1986-01-01

    This volume covers nuclear power plant heat cycles (type of nuclear power cycles, power cycle refinements, BWR/PWR power cycle, BWR/PWR reactor coolant system), reactor vessel internal arrangement (reactor vessel features, BWR/PWR reactor vessel and internals, BWR/PWR reactor core), reactor auxiliary systems (purpose of reactor auxiliary systems, PWR and BWR reactor auxiliary systems, PWR and BWR control rod drive mechanisms).

  6. NEUTRONIC REACTOR POWER PLANT

    DOEpatents

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

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

    SciTech Connect

    Chang Oh; Richard Moore; Robert Barner

    2005-05-01

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

  8. A study of a zone approach to IAEA (International Atomic Energy Agency) safeguards: The low-enriched-uranium zone of a light-water-reactor fuel cycle

    SciTech Connect

    Fishbone, L.G.; Higinbotham, W.A.

    1986-06-01

    At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the conclusions regarding the effectiveness of safeguards for the individual facilities within a state. In this study it was convenient to define three zones in a state with a closed light-water-reactor nuclear fuel cycle. Each zone contains those facilities or parts thereof which use or process nuclear materials of the same safeguards significance: low-enriched uranium, radioactive spent fuel, or recovered plutonium. The possibility that each zone might be treated as an extended material balance area for safeguards purposes is under investigation. The approach includes defining the relevant features of the facilities in the three zones and listing the safeguards activities which are now practiced. This study has focussed on the fresh-fuel zone, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. There are a number of possible safeguards approaches which fall between the two extremes. The intention is to develop a rational approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the approach involving the zone as a material balance area, and for some reasonable intermediate safeguards approaches.

  9. REACTOR CONTROL

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  10. Modeling the impacts of solar radiation partitioning into direct and diffuse fractions for the global water cycle

    NASA Astrophysics Data System (ADS)

    Oliveira, Paulo J. C.; Davin, Edouard L.; Seneviratne, Sonia I.

    2010-05-01

    Incident solar radiation at the Earth's surface affects plant photosynthesis and evapotranspiration, and consequently the global water budget. Observations from 1960-1990's across the Northern Hemisphere suggest that increased aerosol loadings from industrialization led not only to a decline in the intensity of solar radiation at the surface (global dimming), but also to a higher fraction of scattered light, which enhanced plant photosynthesis and the land carbon sink, with probable concurrent impacts on the water cycle. Thus, we used the NCAR Community Land Model (version 3.5) to perform global offline simulations and study the effects of the imposition of changes to radiation partitioning in diffuse and direct fractions on trends in evapotranspiration and runoff. We find that most modeled land surface variables respond to an increased-diffuse simulation where the relative fraction of radiation is changed globally at a high rate of increased diffuse as reported by some observation stations. Increased-diffuse partitioning causes a rise in total ET in all regions, an effect of opposite sign but smaller absolute value than that resulting from global dimming. Evapotranspiration rises by over 0.5 watt/m2 per decade in the tropics, due to increased shaded leaf stomatal conductance, with an opposite effect noted elsewhere due to lower ground evaporation. In the eastern U.S.A. and the Amazon basin, decadal trend anomalies in evapotranspiration for increased-diffuse radiation change reach 25-30% the absolute magnitude of those caused by dimming. Reductions to river runoff are modest nearly everywhere outside the Amazon. Understanding the mechanisms behind the interactions between solar radiation and the various land-surface components will help the development of climate models, improving predictions, in particular regarding changes in terrestrial hydrologic resources.

  11. Potential enhancement of direct interspecies electron transfer for syntrophic metabolism of propionate and butyrate with biochar in up-flow anaerobic sludge blanket reactors.

    PubMed

    Zhao, Zhiqiang; Zhang, Yaobin; Holmes, Dawn E; Dang, Yan; Woodard, Trevor L; Nevin, Kelly P; Lovley, Derek R

    2016-06-01

    Promoting direct interspecies electron transfer (DIET) to enhance syntrophic metabolism may be a strategy for accelerating the conversion of organic wastes to methane, but microorganisms capable of metabolizing propionate and butyrate via DIET under methanogenic conditions have yet to be identified. In an attempt to establish methanogenic communities metabolizing propionate or butyrate with DIET, enrichments were initiated with up-flow anaerobic sludge blanket (UASB), similar to those that were previously reported to support communities that metabolized ethanol with DIET that relied on direct biological electrical connections. In the absence of any amendments, microbial communities enriched were dominated by microorganisms closely related to pure cultures that are known to metabolize propionate or butyrate to acetate with production of H2. When biochar was added to the reactors there was a substantial enrichment on the biochar surface of 16S rRNA gene sequences closely related to Geobacter and Methanosaeta species known to participate in DIET. PMID:26967338

  12. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    SciTech Connect

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

  13. Perspectives on research reactor utilization

    NASA Astrophysics Data System (ADS)

    Dodd, Brian; Dolan, Thomas J.; Laraia, Michele; Ritchie, Iain

    2002-01-01

    The current state of research reactors around the world is summarized using information from the Research Reactor Database. Some current trends of research reactors in advanced and developing countries are described. The need for strategic planning is emphasized, and elements of a typical strategic plan are presented. The problems of reactor lifetime extension, nuclear fuel cycle issues, and decommissioning are briefly discussed. It is concluded that research reactors will continue to be vital elements of the nuclear infrastructures in many countries, and that the IAEA can help countries solve their problems of utilization, safety, lifetime extension, fuel cycle, and decommissioning.

  14. Fast Breeder Reactor studies

    SciTech Connect

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  15. Some properties of a granular activated carbon-sequencing batch reactor (GAC-SBR) system for treatment of textile wastewater containing direct dyes.

    PubMed

    Sirianuntapiboon, Suntud; Sadahiro, Ohmomo; Salee, Paneeta

    2007-10-01

    Resting (living) bio-sludge from a domestic wastewater treatment plant was used as an adsorbent of both direct dyes and organic matter in a sequencing batch reactor (SBR) system. The dye adsorption capacity of the bio-sludge was not increased by acclimatization with direct dyes. The adsorption of Direct Red 23 and Direct Blue 201 onto the bio-sludge was almost the same. The resting bio-sludge showed higher adsorption capacity than the autoclaved bio-sludge. The resting bio-sludge that was acclimatized with synthetic textile wastewater (STWW) without direct dyes showed the highest Direct Blue 201, COD, and BOD(5) removal capacities of 16.1+/-0.4, 453+/-7, and 293+/-9 mg/g of bio-sludge, respectively. After reuse, the dye adsorption ability of deteriorated bio-sludge was recovered by washing with 0.1% sodium dodecyl sulfate (SDS) solution. The direct dyes in the STWW were also easily removed by a GAC-SBR system. The dye removal efficiencies were higher than 80%, even when the system was operated under a high organic loading of 0.36kgBOD(5)/m(3)-d. The GAC-SBR system, however, showed a low direct dye removal efficiency of only 57+/-2.1% with raw textile wastewater (TWW) even though the system was operated with an organic loading of only 0.083kgBOD(5)/m(3)-d. The dyes, COD, BOD(5), and total kjeldalh nitrogen removal efficiencies increased up to 76.0+/-2.8%, 86.2+/-0.5%, 84.2+/-0.7%, and 68.2+/-2.1%, respectively, when 0.89 g/L glucose (organic loading of 0.17kgBOD(5)/m(3)-d) was supplemented into the TWW.

  16. Direct Carbon Conversion: Review of Production and Electrochemical Conversion of Reactive Carbons, Economics and Potential Impact on the Carbon Cycle

    SciTech Connect

    Cooper, J F; Cherepy, N; Upadhye, R; Pasternak, A; Steinberg, M

    2000-12-12

    Concerns over global warning have motivated the search for more efficient technologies for electric power generation from fossil fuels. Today, 90% of electric power is produced from coal, petroleum or natural gas. Higher efficiency reduces the carbon dioxide emissions per unit of electric energy. Exercising an option of deep geologic or ocean sequestration for the CO{sub 2} byproduct would reduce emissions further and partially forestall global warming. We introduce an innovative concept for conversion of fossil fuels to electricity at efficiencies in the range of 70-85% (based on standard enthalpy of the combustion reaction). These levels exceed the performance of common utility plants by up to a factor of two. These levels are also in excess of the efficiencies of combined cycle plants and of advanced fuel cells now operated on the pilot scale. The core of the concept is direct carbon conversion a process that is similar to that a fuel cell but differs in that synthesized forms of carbon, not hydrogen, are used as fuel. The cell sustains the reaction, C + O{sub 2} = CO{sub 2} (E {approx} 1.0 V, T = 800 C). The fuel is in the form of fine particulates ({approx}100 nm) distributed by entrainment in a flow of CO{sub 2} to the cells to form a slurry of carbon in the melt. The byproduct stream of CO{sub 2} is pure. It affords the option of sequestration without additional separation costs, or can be reused in secondary oil or gas recovery. Our experimental program has discovered carbon materials with orders of magnitude spreads in anode reactivity reflected in cell power density. One class of materials yields energy at about 1 kW/m{sup 2} sufficiently high to make practical the use of the cell in electric utility applications. The carbons used in such cells are highly disordered on the nanometer scale (2-30 nm), relative to graphite. Such disordered or turbostratic carbons can be produced by controlled pyrolysis (thermal decomposition) of hydrocarbons extracted from

  17. An x-ray method for direct determination of the strain state and strain relaxation in micron-scale passivated metallization lines during thermal cycling

    SciTech Connect

    Besser, P.R. ); Brennan, S. ); Bravman, J.C. )

    1994-01-01

    We describe a method for directly determining the strain state of passivated metal lines. Synchrotron radiation in the grazing incidence geometry is used to directly measure the in-plane interplanar spacing along the length and width of the lines, while the strain normal to the surface of the line is measured using conventional diffraction methods. The entire strain state is thereby defined. Previous work has measured out-of-plane reflections, fit them to a straight line as a trigonometric function of the angle of orientation, and extrapolated to determine the principal strains. The equivalence of the two x-ray methods on the same sample is demonstrated at room temperature before and after thermal cycling. For short time strain relaxation experiments during thermal cycling, measurement of the three principal strains leads to the direct calculation of the stress relaxation. We apply the strain determination technique to Al--0.5%Cu lines passivated with Si[sub 3]N[sub 4] as the lines are thermally cycled from room temperature to 450 [degree]C and back. The strain state, stress state, and strain relaxation of the lines are calculated at several temperatures during thermal cycling.

  18. Direct Observation of Secondary Organic Aerosol Formation during Cloud Condensation-Evaporation Cycles (SOAaq) in Simulation Chamber Experiments

    NASA Astrophysics Data System (ADS)

    Doussin, J. F.; Bregonzio-Rozier, L.; Giorio, C.; Siekmann, F.; Gratien, A.; Temime-Roussel, B.; Ravier, S.; Pangui, E.; Tapparo, A.; Kalberer, M.; Monod, A.

    2014-12-01

    Biogenic volatile organic compounds (BVOCs) undergo many reactions in the atmosphere and form a wide range of oxidised and water-soluble compounds. These compounds can partition into atmospheric water droplets, and react within the aqueous phase producing higher molecular weight and/or less volatile compounds which can remain in the particle phase after water evaporation and thus increase the organic aerosol mass (Ervens et al., 2011; Altieri et al., 2008; Couvidat et al., 2013). While this hypothesis is frequently discussed in the literature, so far, almost no direct observations of such a process have been provided.The aim of the present work is to study SOA formation from isoprene photooxidation during cloud condensation-evaporation cycles.The experiments were performed during the CUMULUS project (CloUd MULtiphase chemistry of organic compoUndS in the troposphere), in the CESAM simulation chamber located at LISA. CESAM is a 4.2 m3 stainless steel chamber equipped with realistic irradiation sources and temperature and relative humidity (RH) controls (Wang et al., 2011). In each experiment, isoprene was allowed to oxidize during several hours in the presence on nitrogen oxides under dry conditions. Gas phase compounds were analyzed on-line by a Proton Transfer Reaction Time of Flight Mass Spectrometer (PTR-ToF-MS), a Fourier Transform Infrared Spectrometer (FTIR), NOx and O3 analyzers. SOA formation was monitored on-line with a Scanning Mobility Particle Sizer (SMPS) and an Aerodyne High Resolution Time-of-Flight Aerosol Mass Spectrometer (HR-ToF-AMS). The experimental protocol was optimised to generate cloud events in the simulation chamber, which allowed us to generate clouds lasting for ca. 10 minutes in the presence of light.In all experiments, we observed that during cloud formation, water-soluble gas-phase oxidation products (e.g., methylglyoxal, hydroxyacetone, acetaldehyde, formic acid, acetic acid and glycolaldehyde) readily partitioned into cloud

  19. The effect of thermal cycling to 1100 degree C on the alpha (Mo) phase in directionally solidified gamma/gamma prime-alpha alloys

    NASA Technical Reports Server (NTRS)

    Harf, F. H.

    1981-01-01

    In gamma/gamma prime - alpha eutectic alloys (Ni-Mo-Al), the resistance of the alpha phase to morphological changes during thermal cycling was found to be dependent on the structure formed during directional solidification. Fine, smooth alpha fibers survived up to 1000 five minute cycles to 1100 C with minor microstructural contour changes, while coarser and irregularly shaped alpha fibers tended to spheroidize. A mechanism to explain this phenomenon is proposed. It is suggested that on heating to 1100 C, the alpha phase is likely to undergo morphological changes, until differential thermal expansion creates a stress free interface between the alpha phase and the gamma/gamma prime matrix.

  20. Treatability studies with granular activated carbon (GAC) and sequencing batch reactor (SBR) system for textile wastewater containing direct dyes.

    PubMed

    Sirianuntapiboon, Suntud; Sansak, Jutarat

    2008-11-30

    The GAC-SBR efficiency was decreased with the increase of dyestuff concentration or the decrease of bio-sludge concentration. The system showed the highest removal efficiency with synthetic textile wastewater (STWW) containing 40 mg/L direct red 23 or direct blue 201 under MLSS of 3,000 mg/L and hydraulic retention time (HRT) of 7.5 days. But, the effluent NO(3)(-) was higher than that of the influent. Direct red 23 was more effective than direct blue 201 to repress the GAC-SBR system efficiency. The dyes removal efficiency of the system with STWW containing direct red 23 was reduced by 30% with the increase of direct red 23 from 40 mg/L to 160 mg/L. The system with raw textile wastewater (TWW) showed quite low BOD(5) TKN and dye removal efficiencies of only 64.7+/-4.9% and 50.2+/-6.9%, respectively. But its' efficiencies could be increased by adding carbon sources (BOD(5)). The dye removal efficiency with TWW was increased by 30% and 20% by adding glucose (TWW+glucose) or Thai rice noodle wastewater (TWW+TRNWW), respectively. SRT of the systems were 28+/-1 days and 31+/-2 days with TWW+glucose and TWW+TRNWW, respectively.

  1. 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.

  2. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  3. A fuel-cell reactor for the direct synthesis of hydrogen peroxide alkaline solutions from H(2) and O(2).

    PubMed

    Yamanaka, Ichiro; Onisawa, Takeshi; Hashimoto, Toshikazu; Murayama, Toru

    2011-04-18

    The effects of the type of fuel-cell reactors (undivided or divided by cation- and anion-exchange membranes), alkaline electrolytes (LiOH, NaOH, KOH), vapor-grown carbon fiber (VGCF) cathode components (additives: none, activated carbon, Valcan XC72, Black Pearls 2000, Seast-6, and Ketjen Black), and the flow rates of anolyte (0, 1.5, 12 mL h(-1)) and catholyte (0, 12 mL h(-1)) on the formation of hydrogen peroxide were studied. A divided fuel-cell system, O(2) (g)|VGCF-XC72 cathode|2 M NaOH catholyte|cation-exchange membrane (Nafion-117)|Pt/XC72-VGCF anode|2 M NaOH anolyte at 12 mL h(-1) flow|H(2) (g), was effective for the selective formation of hydrogen peroxide, with 130 mA cm(-2) , a 2 M aqueous solution of H(2)O(2)/NaOH, and a current efficiency of 95 % at atmospheric pressure and 298 K. The current and formation rate gradually decreased over a long period of time. The cause of the slow decrease in electrocatalytic performance was revealed and the decrease was stopped by a flow of catholyte. Cyclic voltammetry studies at the VGCF-XC72 electrode indicated that fast diffusion of O(2) from the gas phase to the electrode, and quick desorption of hydrogen peroxide from the electrode to the electrolyte were essential for the efficient formation of solutions of H(2)O(2)/NaOH.

  4. Evaluation of technical feasibility of closed-cycle non-equilibrium MHD power generation with direct coal firing. Final report, Task 1

    SciTech Connect

    Not Available

    1981-11-01

    Program accomplishments in a continuing effort to demonstrate the feasibility of direct coal fired, closed cycle, magnetohydrodynamic power generation are detailed. These accomplishments relate to all system aspects of a CCMHD power generation system including coal combustion, heat transfer to the MHD working fluid, MHD power generation, heat and cesium seed recovery and overall systems analysis. Direct coal firing of the combined cycle has been under laboratory development in the form of a high slag rejection, regeneratively air cooled cyclone coal combustor concept, originated within this program. A hot bottom ceramic regenerative heat exchanger system was assembled and test fired with coal for the purposes of evaluating the catalytic effect of alumina on NO/sub x/ emission reduction and operability of the refractory dome support system. Design, procurement, fabrication and partial installation of a heat and seed recovery flow apparatus was accomplished and was based on a stream tube model of the full scale system using full scale temperatures, tube sizes, rates of temperature change and tube geometry. Systems analysis capability was substantially upgraded by the incorporation of a revised systems code, with emphasis on ease of operator interaction as well as separability of component subroutines. The updated code was used in the development of a new plant configuration, the Feedwater Cooled (FCB) Brayton Cycle, which is superior to the CCMHD/Steam cycle both in performance and cost. (WHK)

  5. Regaining Focus in Irish Junior Cycle Science: Potential New Directions for Curriculum and Assessment on Nature of Science

    ERIC Educational Resources Information Center

    Erduran, Sibel; Dagher, Zoubeida R.

    2014-01-01

    The Irish national discourse on curriculum and assessment reform at the Junior Cycle level has been fraught with controversy in the past two years. The introduction of the new curriculum and assessment framework in 2012 by the then Minister of Education, Ruairi Quinn has led to significant media coverage and teacher union response. In this paper,…

  6. Corrigendum to "Sinusoidal potential cycling operation of a direct ethanol fuel cell to improving carbon dioxide yields" [J. Power Sources 268 (5 December 2014) 439-442

    NASA Astrophysics Data System (ADS)

    Majidi, Pasha; Pickup, Peter G.

    2016-09-01

    The authors regret that Equation (5) is incorrect and has resulted in errors in Fig. 4 and the efficiencies stated on p. 442. The corrected equation, figure and text are presented below. In addition, the title should be 'Sinusoidal potential cycling operation of a direct ethanol fuel cell to improve carbon dioxide yields', and the reversible cell potential quoted on p. 441 should be 1.14 V. The authors would like to apologise for any inconvenience caused.

  7. REACTOR MONITORING

    DOEpatents

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  8. NEUTRONIC REACTOR

    DOEpatents

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  9. Paper 8775 - Integrating Natural Resources and Ecological Science into the Disaster Risk CYCLE: Lessons Learned and Future Directions

    NASA Astrophysics Data System (ADS)

    Brosnan, D. M.

    2014-12-01

    Familiar to disaster risk reduction (DRR) scientists and professionals, the disaster cycle is an adaptive approach that involves planning, response and learning for the next event. It has proven effective in saving lives and helping communities around the world deal with natural and other hazards. But it has rarely been applied to natural resource and ecological science, despite the fact that many communities are dependent on these resources. This presentation will include lessons learned from applying science to tackle ecological consequences in several disasters in the US and globally, including the Colorado Floods, the SE Asia tsunami, the Montserrat volcanic eruption, and US SAFRR tsunami scenario. The presentation discusses the role that science and scientists can play at each phase of the disaster cycle. The consequences of not including disaster cycles in the management of natural systems leaves these resources and the huge investments made to protect highly vulnerable. The presentation discusses how The presentation discusses how science can help government and communities in planning and responding to these events. It concludes with a set of lessons learned and guidlines for moving forward.

  10. Direct evidence of estrogen modulation of pituitary sensitivity to luteinizing hormone-releasing factor during the menstrual cycle.

    PubMed Central

    Wang, C F; Yen, S S

    1975-01-01

    To delineate the role of estradiol in the augmented pituitary gonadotropin responsiveness to synthetic luteinizing hormone releasing factor (LRF) seen during high-estrogen phases of the ovulatory cycles (late follicular and midluteal phases), the anti-estrogenic effect of clomiphene citrate (Clomid) on pituitary response to LRF was evaluated during different phases of the ovulatory cycle. Clomid administration (100 mg/day times 5 days) completely negates the augmented gonadotropin responses to LRF (150 mug) during late follicular and midluteal phases observed during the control studies. Thus, a quantitatively and qualitatively similar pituitary sensitivity to LRF during three distinct phases of the menstrual cycle was induced by Clomid treatment that resembles the LRF responsiveness of themale pituitary. The present study demonstrates the pituitary component of the estrogen-induced changes in the sensitivity to LRF. From this and previous data, we conclude that the increases of estradiol secretion associated with the follicular maturation and corpus luteum formation represent a major component of the feedback signal in the modulation of cyclic gonadotropin release occasioned in a large measure by the augmented pituitary sensitivity to LRF. PMID:1088908

  11. Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

    SciTech Connect

    Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi

    2014-09-30

    Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided.

  12. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    SciTech Connect

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  13. Effect of thermal cycling in a Mach 0.3 burner rig on properties and structure of directionally solidified gamma/gamma prime-delta eutectic

    NASA Technical Reports Server (NTRS)

    Gray, H. R.; Sanders, W. A.

    1976-01-01

    An experimental study was carried out to evaluate the effect of cyclic thermal exposures on the mechanical properties of a gamma/gamma prime-delta eutectic alloy parallel to the growth direction. The alloy had a nominal composition by weight of Ni-20 Nb-6 Cr-2.5 Al and was directionally solidified at 3 cm/hr in a furnace with a thermal gradient of at least 200 C/cm. Bars of the alloy were exposed in a Mach 0.3 burner rig and cycled 300 times between 1100 and 425 C. Oxidation-erosion characteristics of the alloy were determined by weight loss measurements at 300-cycle intervals. After cyclic exposure, stress rupture and tensile tests were performed at both 760 and 1040 C. Microstructural changes from cyclic exposure were determined. Thermal cycling resulted in gamma prime coarsening and Widmanstaetten delta precipitation in the gamma phase. An unidentified precipitate, presumably gamma prime, was observed within the delta phase. These microstructural changes did not affect the mechanical properties of the eutectic. High oxidation-erosion weight loss rate was observed.

  14. A localized nucleolar DNA damage response facilitates recruitment of the homology-directed repair machinery independent of cell cycle stage

    PubMed Central

    van Sluis, Marjolein; McStay, Brian

    2015-01-01

    DNA double-strand breaks (DSBs) are repaired by two main pathways: nonhomologous end-joining and homologous recombination (HR). Repair pathway choice is thought to be determined by cell cycle timing and chromatin context. Nucleoli, prominent nuclear subdomains and sites of ribosome biogenesis, form around nucleolar organizer regions (NORs) that contain rDNA arrays located on human acrocentric chromosome p-arms. Actively transcribed rDNA repeats are positioned within the interior of the nucleolus, whereas sequences proximal and distal to NORs are packaged as heterochromatin located at the nucleolar periphery. NORs provide an opportunity to investigate the DSB response at highly transcribed, repetitive, and essential loci. Targeted introduction of DSBs into rDNA, but not abutting sequences, results in ATM-dependent inhibition of their transcription by RNA polymerase I. This is coupled with movement of rDNA from the nucleolar interior to anchoring points at the periphery. Reorganization renders rDNA accessible to repair factors normally excluded from nucleoli. Importantly, DSBs within rDNA recruit the HR machinery throughout the cell cycle. Additionally, unscheduled DNA synthesis, consistent with HR at damaged NORs, can be observed in G1 cells. These results suggest that HR can be templated in cis and suggest a role for chromosomal context in the maintenance of NOR genomic stability. PMID:26019174

  15. Status of French reactors

    SciTech Connect

    Ballagny, A.

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  16. Moving bed reactor for solar thermochemical fuel production

    SciTech Connect

    Ermanoski, Ivan

    2013-04-16

    Reactors and methods for solar thermochemical reactions are disclosed. Embodiments of reactors include at least two distinct reactor chambers between which there is at least a pressure differential. In embodiments, reactive particles are exchanged between chambers during a reaction cycle to thermally reduce the particles at first conditions and oxidize the particles at second conditions to produce chemical work from heat. In embodiments, chambers of a reactor are coupled to a heat exchanger to pre-heat the reactive particles prior to direct exposure to thermal energy with heat transferred from reduced reactive particles as the particles are oppositely conveyed between the thermal reduction chamber and the fuel production chamber. In an embodiment, particle conveyance is in part provided by an elevator which may further function as a heat exchanger.

  17. Systems efficiency and specific mass estimates for direct and indirect solar-pumped closed-cycle high-energy lasers in space

    NASA Technical Reports Server (NTRS)

    Monson, D. J.

    1978-01-01

    Based on expected advances in technology, the maximum system efficiency and minimum specific mass have been calculated for closed-cycle CO and CO2 electric-discharge lasers (EDL's) and a direct solar-pumped laser in space. The efficiency calculations take into account losses from excitation gas heating, ducting frictional and turning losses, and the compressor efficiency. The mass calculations include the power source, radiator, compressor, fluids, ducting, laser channel, optics, and heat exchanger for all of the systems; and in addition the power conditioner for the EDL's and a focusing mirror for the solar-pumped laser. The results show the major component masses in each system, show which is the lightest system, and provide the necessary criteria for solar-pumped lasers to be lighter than the EDL's. Finally, the masses are compared with results from other studies for a closed-cycle CO2 gasdynamic laser (GDL) and the proposed microwave satellite solar power station (SSPS).

  18. Upper ocean carbon cycling inferred from direct pH observations made by profiling floats and estimated alkalinity

    NASA Astrophysics Data System (ADS)

    Johnson, K. S.; Plant, J. N.; Jannasch, H. W.; Coletti, L. J.; Elrod, V.; Sakamoto, C.; Riser, S.

    2015-12-01

    The annual cycle of dissolved inorganic carbon (DIC) is a key tracer of net community production and carbon export in the upper ocean. In particular, the DIC concentration is much less sensitive to air-sea gas exchange, when compared to oxygen, another key tracer of upper ocean metabolism. However, the annual DIC cycle is observed with a seasonal resolution at only a few time-series stations in the open ocean. Here, we consider the annual carbon cycle that has been observed using profiling floats equipped with pH sensors. Deep-Sea DuraFET pH sensors have been deployed on profiling floats for over three years and they can provide temporal and spatial resolution of 5 to 10 days and 5 to 10 m in the upper ocean over multi-year periods. In addition to pH, a second carbon system parameter is required to compute DIC. Total alkalinity can be derived from the float observations of temperature, salinity and oxygen using equations in these variables that are fitted to shipboard observations of alkalinity obtained in the global repeat hydrography programs (e.g., Juranek et al., GRL, doi:10.1029/2011GL048580, 2011), as the relationships should be stable in time in the open ocean. Profiling floats with pH have been deployed from Hawaii Ocean Time-series (HOT) cruises since late 2012 and an array of floats with pH have been deployed since early 2014 in the Southern Ocean as part of the SOCCOM program. The SOCCOM array should grow to nearly 200 floats over the next 5 years. The sensor data was quality controlled and adjusted by comparing observations at 1500 m depth to the deep climatology of pH (derived from DIC and alkalinity) computed with the GLODAP data set. After adjustment, the surface DIC concentrations were calculated from pH and alkalinity. This yields a data set that is used to examine annual net community production in the oligotrophic North Pacific and in the South Pacific near 150 West from 40 South to 65 South.

  19. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  20. Benchmarking NSP Reactors with CORETRAN-01

    SciTech Connect

    Hines, Donald D.; Grow, Rodney L.; Agee, Lance J

    2004-10-15

    As part of an overall verification and validation effort, the Electric Power Research Institute's (EPRIs) CORETRAN-01 has been benchmarked against Northern States Power's Prairie Island and Monticello reactors through 12 cycles of operation. The two Prairie Island reactors are Westinghouse 2-loop units with 121 asymmetric 14 x 14 lattice assemblies utilizing up to 8 wt% gadolinium while Monticello is a General Electric 484 bundle boiling water reactor. All reactor cases were executed in full core utilizing 24 axial nodes per assembly in the fuel with 1 additional reflector node above, below, and around the perimeter of the core. Cross-section sets used in this benchmark effort were generated by EPRI's CPM-3 as well as Studsvik's CASMO-3 and CASMO-4 to allow for separation of the lattice calculation effect from the nodal simulation method. These cases exercised the depletion-shuffle-depletion sequence through four cycles for each unit using plant data to follow actual operations. Flux map calculations were performed for comparison to corresponding measurement statepoints. Additionally, start-up physics testing cases were used to predict cycle physics parameters for comparison to existing plant methods and measurements.These benchmark results agreed well with both current analysis methods and plant measurements, indicating that CORETRAN-01 may be appropriate for steady-state physics calculations of both the Prairie Island and Monticello reactors. However, only the Prairie Island results are discussed in this paper since Monticello results were of similar quality and agreement. No attempt was made in this work to investigate CORETRAN-01 kinetics capability by analyzing plant transients, but these steady-state results form a good foundation for moving in that direction.

  1. Cell cycle-dependent adaptor complex for ClpXP-mediated proteolysis directly integrates phosphorylation and second messenger signals.

    PubMed

    Smith, Stephen C; Joshi, Kamal K; Zik, Justin J; Trinh, Katherine; Kamajaya, Aron; Chien, Peter; Ryan, Kathleen R

    2014-09-30

    The cell-division cycle of Caulobacter crescentus depends on periodic activation and deactivation of the essential response regulator CtrA. Although CtrA is critical for transcription during some parts of the cell cycle, its activity must be eliminated before chromosome replication because CtrA also blocks the initiation of DNA replication. CtrA activity is down-regulated both by dephosphorylation and by proteolysis, mediated by the ubiquitous ATP-dependent protease ClpXP. Here we demonstrate that proteins needed for rapid CtrA proteolysis in vivo form a phosphorylation-dependent and cyclic diguanylate (cdG)-dependent adaptor complex that accelerates CtrA degradation in vitro by ClpXP. The adaptor complex includes CpdR, a single-domain response regulator; PopA, a cdG-binding protein; and RcdA, a protein whose activity cannot be predicted. When CpdR is unphosphorylated and when PopA is bound to cdG, they work together with RcdA in an all-or-none manner to reduce the Km of CtrA proteolysis 10-fold. We further identified a set of amino acids in the receiver domain of CtrA that modulate its adaptor-mediated degradation in vitro and in vivo. Complex formation between PopA and CtrA depends on these amino acids, which reside on alpha-helix 1 of the CtrA receiver domain, and on cdG binding by PopA. These results reveal that each accessory factor plays an essential biochemical role in the regulated proteolysis of CtrA and demonstrate, to our knowledge, the first example of a multiprotein, cdG-dependent proteolytic adaptor.

  2. Cell cycle-dependent adaptor complex for ClpXP-mediated proteolysis directly integrates phosphorylation and second messenger signals

    PubMed Central

    Smith, Stephen C.; Joshi, Kamal K.; Zik, Justin J.; Trinh, Katherine; Kamajaya, Aron; Chien, Peter; Ryan, Kathleen R.

    2014-01-01

    The cell-division cycle of Caulobacter crescentus depends on periodic activation and deactivation of the essential response regulator CtrA. Although CtrA is critical for transcription during some parts of the cell cycle, its activity must be eliminated before chromosome replication because CtrA also blocks the initiation of DNA replication. CtrA activity is down-regulated both by dephosphorylation and by proteolysis, mediated by the ubiquitous ATP-dependent protease ClpXP. Here we demonstrate that proteins needed for rapid CtrA proteolysis in vivo form a phosphorylation-dependent and cyclic diguanylate (cdG)-dependent adaptor complex that accelerates CtrA degradation in vitro by ClpXP. The adaptor complex includes CpdR, a single-domain response regulator; PopA, a cdG-binding protein; and RcdA, a protein whose activity cannot be predicted. When CpdR is unphosphorylated and when PopA is bound to cdG, they work together with RcdA in an all-or-none manner to reduce the Km of CtrA proteolysis 10-fold. We further identified a set of amino acids in the receiver domain of CtrA that modulate its adaptor-mediated degradation in vitro and in vivo. Complex formation between PopA and CtrA depends on these amino acids, which reside on alpha-helix 1 of the CtrA receiver domain, and on cdG binding by PopA. These results reveal that each accessory factor plays an essential biochemical role in the regulated proteolysis of CtrA and demonstrate, to our knowledge, the first example of a multiprotein, cdG-dependent proteolytic adaptor. PMID:25197043

  3. Safeguards Considerations for Thorium Fuel Cycles

    DOE PAGES

    Worrall, Louise G.; Worrall, Andrew; Flanagan, George F.; Croft, Steven

    2016-04-21

    We report that by around 2025, thorium-based fuel cycles are likely to be deployed internationally. States such as China and India are pursuing research, development, and deployment pathways toward a number of commercial-scale thorium fuel cycles, and they are already building test reactors and the associated fuel cycle infrastructure. In the future, the potential exists for these emerging programs to sell, export, and deploy thorium fuel cycle technology in other states. Without technically adequate international safeguards protocols and measures in place, any future potential clandestine misuse of these fuel cycles could go undetected, compromising the deterrent value of these protocolsmore » and measures. The development of safeguards approaches for thorium-based fuel cycles is therefore a matter of some urgency. Yet, the focus of the international safeguards community remains mainly on safeguarding conventional 235U- and 239Pu-based fuel cycles while the safeguards challenges of thorium-uranium fuel cycles remain largely uninvestigated. This raises the following question: Is the International Atomic Energy Agency and international safeguards system ready for thorium fuel cycles? Furthermore, is the safeguards technology of today sufficiently mature to meet the verification challenges posed by thorium-based fuel cycles? In defining these and other related research questions, the objectives of this paper are to identify key safeguards considerations for thorium-based fuel cycles and to call for an early dialogue between the international safeguards and the nuclear fuel cycle communities to prepare for the potential safeguards challenges associated with these fuel cycles. In this paper, it is concluded that directed research and development programs are required to meet the identified safeguards challenges and to take timely action in preparation for the international deployment of thorium fuel cycles.« less

  4. Effect of Drive Cycle and Gasoline Particulate Filter on the Size and Morphology of Soot Particles Emitted from a Gasoline-Direct-Injection Vehicle.

    PubMed

    Saffaripour, Meghdad; Chan, Tak W; Liu, Fengshan; Thomson, Kevin A; Smallwood, Gregory J; Kubsh, Joseph; Brezny, Rasto

    2015-10-01

    The size and morphology of particulate matter emitted from a light-duty gasoline-direct-injection (GDI) vehicle, over the FTP-75 and US06 transient drive cycles, have been characterized by transmission-electron-microscope (TEM) image analysis. To investigate the impact of gasoline particulate filters on particulate-matter emission, the results for the stock-GDI vehicle, that is, the vehicle in its original configuration, have been compared to the results for the same vehicle equipped with a catalyzed gasoline particulate filter (GPF). The stock-GDI vehicle emits graphitized fractal-like aggregates over all driving conditions. The mean projected area-equivalent diameter of these aggregates is in the 78.4-88.4 nm range and the mean diameter of primary particles varies between 24.6 and 26.6 nm. Post-GPF particles emitted over the US06 cycle appear to have an amorphous structure, and a large number of nucleation-mode particles, depicted as low-contrast ultrafine droplets, are observed in TEM images. This indicates the emission of a substantial amount of semivolatile material during the US06 cycle, most likely generated by the incomplete combustion of accumulated soot in the GPF during regeneration. The size of primary particles and soot aggregates does not vary significantly by implementing the GPF over the FTP-75 cycle; however, particles emitted by the GPF-equipped vehicle over the US06 cycle are about 20% larger than those emitted by the stock-GDI vehicle. This may be attributed to condensation of large amounts of organic material on soot aggregates. High-contrast spots, most likely solid nonvolatile cores, are observed within many of the nucleation-mode particles emitted over the US06 cycle by the GPF-equipped vehicle. These cores are either generated inside the engine or depict incipient soot particles which are partially carbonized in the exhaust line. The effect of drive cycle and the GPF on the fractal parameters of particles, such as fractal dimension and

  5. Effect of Drive Cycle and Gasoline Particulate Filter on the Size and Morphology of Soot Particles Emitted from a Gasoline-Direct-Injection Vehicle.

    PubMed

    Saffaripour, Meghdad; Chan, Tak W; Liu, Fengshan; Thomson, Kevin A; Smallwood, Gregory J; Kubsh, Joseph; Brezny, Rasto

    2015-10-01

    The size and morphology of particulate matter emitted from a light-duty gasoline-direct-injection (GDI) vehicle, over the FTP-75 and US06 transient drive cycles, have been characterized by transmission-electron-microscope (TEM) image analysis. To investigate the impact of gasoline particulate filters on particulate-matter emission, the results for the stock-GDI vehicle, that is, the vehicle in its original configuration, have been compared to the results for the same vehicle equipped with a catalyzed gasoline particulate filter (GPF). The stock-GDI vehicle emits graphitized fractal-like aggregates over all driving conditions. The mean projected area-equivalent diameter of these aggregates is in the 78.4-88.4 nm range and the mean diameter of primary particles varies between 24.6 and 26.6 nm. Post-GPF particles emitted over the US06 cycle appear to have an amorphous structure, and a large number of nucleation-mode particles, depicted as low-contrast ultrafine droplets, are observed in TEM images. This indicates the emission of a substantial amount of semivolatile material during the US06 cycle, most likely generated by the incomplete combustion of accumulated soot in the GPF during regeneration. The size of primary particles and soot aggregates does not vary significantly by implementing the GPF over the FTP-75 cycle; however, particles emitted by the GPF-equipped vehicle over the US06 cycle are about 20% larger than those emitted by the stock-GDI vehicle. This may be attributed to condensation of large amounts of organic material on soot aggregates. High-contrast spots, most likely solid nonvolatile cores, are observed within many of the nucleation-mode particles emitted over the US06 cycle by the GPF-equipped vehicle. These cores are either generated inside the engine or depict incipient soot particles which are partially carbonized in the exhaust line. The effect of drive cycle and the GPF on the fractal parameters of particles, such as fractal dimension and

  6. Fast reactors and nuclear nonproliferation

    SciTech Connect

    Avrorin, E.N.; Rachkov, V.I.; Chebeskov, A.N.

    2013-07-01

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  7. Open-cycle magnetohydrodynamic power plant based upon direct-contact closed-loop high-temperature heat exchanger

    DOEpatents

    Berry, Gregory F.; Minkov, Vladimir; Petrick, Michael

    1988-01-05

    A magnetohydrodynamic (MHD) power generating system in which ionized combustion gases with slag and seed are discharged from an MHD combustor and pressurized high temperature inlet air is introduced into the combustor for supporting fuel combustion at high temperatures necessary to ionize the combustion gases, and including a heat exchanger in the form of a continuous loop with a circulating heat transfer liquid such as copper oxide. The heat exchanger has an upper horizontal channel for providing direct contact between the heat transfer liquid and the combustion gases to cool the gases and condense the slag which thereupon floats on the heat transfer liquid and can be removed from the channel, and a lower horizontal channel for providing direct contact between the heat transfer liquid and pressurized air for preheating the inlet air. The system further includes a seed separator downstream of the heat exchanger.

  8. Open-cycle magnetohydrodynamic power plant based upon direct-contact closed-loop high-temperature heat exchanger

    DOEpatents

    Berry, Gregory F.; Minkov, Vladimir; Petrick, Michael

    1988-01-01

    A magnetohydrodynamic (MHD) power generating system in which ionized combustion gases with slag and seed are discharged from an MHD combustor and pressurized high temperature inlet air is introduced into the combustor for supporting fuel combustion at high temperatures necessary to ionize the combustion gases, and including a heat exchanger in the form of a continuous loop with a circulating heat transfer liquid such as copper oxide. The heat exchanger has an upper horizontal channel for providing direct contact between the heat transfer liquid and the combustion gases to cool the gases and condense the slag which thereupon floats on the heat transfer liquid and can be removed from the channel, and a lower horizontal channel for providing direct contact between the heat transfer liquid and pressurized air for preheating the inlet air. The system further includes a seed separator downstream of the heat exchanger.

  9. Open-cycle magnetohydrodynamic power plant based upon direct-contact closed-loop high-temperature heat exchanger

    DOEpatents

    Berry, G.F.; Minkov, V.; Petrick, M.

    1981-11-02

    A magnetohydrodynamic (MHD) power generating system is described in which ionized combustion gases with slag and seed are discharged from an MHD combustor and pressurized high temperature inlet air is introduced into the combustor for supporting fuel combustion at high temperatures necessary to ionize the combustion gases, and including a heat exchanger in the form of a continuous loop with a circulating heat transfer liquid such as copper oxide. The heat exchanger has an upper horizontal channel for providing direct contact between the heat transfer liquid and the combustion gases to cool the gases and condense the slag which thereupon floats on the heat transfer liquid and can be removed from the channel, and a lower horizontal channel for providing direct contact between the heat transfer liquid and pressurized air for preheating the inlet air. The system further includes a seed separator downstream of the heat exchanger.

  10. Experimental Analysis of a Rocket Based Combined Cycle (RBCC) Engine in a Direct-Connect Test Facility

    NASA Technical Reports Server (NTRS)

    Nelson, K.; Hawk, Clark W.

    1997-01-01

    The object of this study is to investigate the operation of a RBCC at ramjet and scramjet flight conditions using a direct-connect test facility. The apparatus being tested is a single strut-rocket within a dual-mode ram/scramjet combustor. The gaseous hydrogen/oxygen, linear strut-rocket was supplied by Aerojet Propulsion Company. The hardware is being tested in the Direct Connect Supersonic Combustion Test Facility at NASA Langley Research Center. The test facilities hydrogen/oxygen vitiated heater is capable of flight total enthalpies to Mach 8. A Mach 2.5 facility nozzle mates the heater to the combustor duct. The rocket ejector will ordinarily operate in a fuel-rich mode. Additional fuel injection is provided by a pair of parallel injectors located at the base of the strut body. Instrumentation on the test apparatus includes a unique, direct thrust measurement system. Performance predictions for the anticipated test conditions have been made using a one-dimensional, thermodynamic analysis code. Results from the code show the dependence of overall thrust and specific impulse on rocket chamber pressure, rocket fuel equivalence ratio, and overall fuel equivalence ratio. Once the experimental test series begins, the inferred combustion efficiency as a function of axial location and the thermal choke region (where applicable) can also be determined using this code. Upon completion of the experimental test series, measurements will be used to calculate thrust, specific impulse, etc. Measured and calculated values will be compared to those found analytically. If appropriate, the code will be tailored to better predict hardware operation. Conclusions will be drawn as to the fuel-rich rocket's overall effect on ramjet and scramjet performance. Also, comparisons will be made between the integrated thrust calculated from the static pressure taps located along the duct and the thrust measured by the direct thrust measurement system.

  11. CONTROL FOR NEUTRONIC REACTOR

    DOEpatents

    Lichtenberger, H.V.; Cameron, R.A.

    1959-03-31

    S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

  12. 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.

  13. Regulation of Cell Proliferation in the Stomatal Lineage by the Arabidopsis MYB FOUR LIPS via Direct Targeting of Core Cell Cycle Genes[W

    PubMed Central

    Xie, Zidian; Lee, EunKyoung; Lucas, Jessica R.; Morohashi, Kengo; Li, Dongmei; Murray, James A.H.; Sack, Fred D.; Grotewold, Erich

    2010-01-01

    Stomata, which are epidermal pores surrounded by two guard cells, develop from a specialized stem cell lineage and function in shoot gas exchange. The Arabidopsis thaliana FOUR LIPS (FLP) and MYB88 genes encode closely related and atypical two-MYB-repeat proteins, which when mutated result in excess divisions and abnormal groups of stomata in contact. Consistent with a role in transcription, we show here that FLP and MYB88 are nuclear proteins with DNA binding preferences distinct from other known MYBs. To identify possible FLP/MYB88 transcriptional targets, we used chromatin immunoprecitation (ChIP) followed by hybridization to Arabidopsis whole genome tiling arrays. These ChIP-chip data indicate that FLP/MYB88 target the upstream regions especially of cell cycle genes, including cyclins, cyclin-dependent kinases (CDKs), and components of the prereplication complex. In particular, we show that FLP represses the expression of the mitosis-inducing factor CDKB1;1, which, along with CDKB1;2, is specifically required both for the last division in the stomatal pathway and for cell overproliferation in flp mutants. We propose that FLP and MYB88 together integrate patterning with the control of cell cycle progression and terminal differentiation through multiple and direct cell cycle targets. FLP recognizes a distinct cis-regulatory element that overlaps with that of the cell cycle activator E2F-DP in the CDKB1;1 promoter, suggesting that these MYBs may also modulate E2F-DP pathways. PMID:20675570

  14. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

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

    NASA Astrophysics Data System (ADS)

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

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

  16. Milestone Report #2: Direct Evaporator Leak and Flammability Analysis Modifications and Optimization of the Organic Rankine Cycle to Improve the Recovery of Waste Heat

    SciTech Connect

    Donna Post Guillen

    2013-09-01

    The direct evaporator is a simplified heat exchange system for an Organic Rankine Cycle (ORC) that generates electricity from a gas turbine exhaust stream. Typically, the heat of the exhaust stream is transferred indirectly to the ORC by means of an intermediate thermal oil loop. In this project, the goal is to design a direct evaporator where the working fluid is evaporated in the exhaust gas heat exchanger. By eliminating one of the heat exchangers and the intermediate oil loop, the overall ORC system cost can be reduced by approximately 15%. However, placing a heat exchanger operating with a flammable hydrocarbon working fluid directly in the hot exhaust gas stream presents potential safety risks. The purpose of the analyses presented in this report is to assess the flammability of the selected working fluid in the hot exhaust gas stream stemming from a potential leak in the evaporator. Ignition delay time for cyclopentane at temperatures and pressure corresponding to direct evaporator operation was obtained for several equivalence ratios. Results of a computational fluid dynamic analysis of a pinhole leak scenario are given.

  17. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, J.

    1996-03-19

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  18. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, Juhani

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  19. Direct measurement of the protein response to an electrostatic perturbation that mimics the catalytic cycle in ketosteroid isomerase.

    PubMed

    Jha, Santosh Kumar; Ji, Minbiao; Gaffney, Kelly J; Boxer, Steven G

    2011-10-01

    Understanding how electric fields and their fluctuations in the active site of enzymes affect efficient catalysis represents a critical objective of biochemical research. We have directly measured the dynamics of the electric field in the active site of a highly proficient enzyme, Δ(5)-3-ketosteroid isomerase (KSI), in response to a sudden electrostatic perturbation that simulates the charge displacement that occurs along the KSI catalytic reaction coordinate. Photoexcitation of a fluorescent analog (coumarin 183) of the reaction intermediate mimics the change in charge distribution that occurs between the reactant and intermediate state in the steroid substrate of KSI. We measured the electrostatic response and angular dynamics of four probe dipoles in the enzyme active site by monitoring the time-resolved changes in the vibrational absorbance (IR) spectrum of a spectator thiocyanate moiety (a quantitative sensor of changes in electric field) placed at four different locations in and around the active site, using polarization-dependent transient vibrational Stark spectroscopy. The four different dipoles in the active site remain immobile and do not align to the changes in the substrate electric field. These results indicate that the active site of KSI is preorganized with respect to functionally relevant changes in electric fields.

  20. The problem of optimizing the water chemistry used in the primary coolant circuit of a nuclear power station equipped with VVER reactors under the conditions of longer fuel cycle campaigns and increased capacity of power units

    NASA Astrophysics Data System (ADS)

    Sharafutdinov, R. B.; Kharitonova, N. L.

    2011-05-01

    It is shown that the optimal water chemistry of the primary coolant circuit must be substantiated while introducing measures aimed at increasing the power output in operating power units and for the project called AES-2006/AES TOI (a typical optimized project of a nuclear power station with enhanced information support). The experience gained from operation of PWR reactors with an elongated fuel cycle at an increased level of power is analyzed. Conditions under which boron compounds are locally concentrated on the fuel rod surfaces (the hideout phenomenon) and axial offset anomaly occurs are enlisted, and the influence of lithium on the hideout in the pores of deposits on the surfaces of fuel assemblies is shown.

  1. An experimental study on the oxidative coupling of methane in a direct current corona discharge reactor over Sr/La{sub 2}O{sub 3} catalyst

    SciTech Connect

    Marafee, A.; Liu, C.; Xu, G.; Mallinson, R.; Lobban, L.

    1997-03-01

    The homogeneous and catalytic oxidative coupling of methane (OCM) for converting methane directly into higher hydrocarbons has been the subject of a large body of research. The present study on conversion of methane in dc corona discharge packed bed reactors may significantly improve the process economics. Experimental investigations have been conducted in which all the reactive gases pass through a catalyst bed which is situated within the corona-induced plasma zone. In this study, a typical OCM catalyst, Sr/La{sub 2}O{sub 3}, was used to investigate experimentally the corona discharge OCM reactions. Experiments were conducted over a wide range of temperatures (823--1,023 K) and input powers (0--6 W) with both positive and negative corona processes. Compared to the catalytic process in the absence of corona discharge, the corona discharge results in higher methane conversion and larger yield of C{sub 2} products even at temperatures at which there is no C{sub 2} activity for the catalyst alone. The methane conversion and C{sub 2} yield increase with O{sub 2} partial pressure during the corona-enhanced catalytic reactions, while the selectivity decreases slightly with increasing O{sub 2} partial pressure. Compared to results obtained in the absence of corona discharges, methane conversion in the presence of the dc corona was nearly five times larger and the selectivity for C{sub 2} over eight times higher at 853 K. A great enhancement in catalytic activity has also been achieved at a temperature at which the catalyst alone shows no C{sub 2} activity. The conversion at higher temperature (more than 953 K) is limited by the poor corona performance and the availability of active oxygen species.

  2. Ignition assist systems for direct-injected, diesel cycle, medium-duty alternative fuel engines: Final report phase 1

    SciTech Connect

    Chan, A.K.

    2000-02-23

    This report is a summary of the results of Phase 1 of this contract. The objective was to evaluate the potential of assist technologies for direct-injected alternative fuel engines vs. glow plug ignition assist. The goal was to demonstrate the feasibility of an ignition system life of 10,000 hours and a system cost of less than 50% of the glow plug system, while meeting or exceeding the engine thermal efficiency obtained with the glow plug system. There were three tasks in Phase 1. Under Task 1, a comprehensive review of feasible ignition options for DING engines was completed. The most promising options are: (1) AC and the ''SmartFire'' spark, which are both long-duration, low-power (LDLP) spark systems; (2) the short-duration, high-power (SDHP) spark system; (3) the micropilot injection ignition; and (4) the stratified charge plasma ignition. Efforts concentrated on investigating the AC spark, SmartFire spark, and short-duration/high-power spark systems. Using proprietary pricing information, the authors predicted that the commercial costs for the AC spark, the short-duration/high-power spark and SmartFire spark systems will be comparable (if not less) to the glow plug system. Task 2 involved designing and performing bench tests to determine the criteria for the ignition system and the prototype spark plug for Task 3. The two most important design criteria are the high voltage output requirement of the ignition system and the minimum electrical insulation requirement for the spark plug. Under Task 3, all the necessary hardware for the one-cylinder engine test was designed. The hardware includes modified 3126 cylinder heads, specially designed prototype spark plugs, ignition system electronics, and parts for the system installation. Two 3126 cylinder heads and the SmartFire ignition system were procured, and testing will begin in Phase 2 of this subcontract.

  3. Physics and Control Assessment of AN 850 Mw(e) - Leu-Candu Reactor.

    NASA Astrophysics Data System (ADS)

    Barbone, Michelangelo

    The physics and control assessment of an 850 MW(e) Low Enriched Uranium CANDU Pressurized Heavy Water (LEU -CANDU-PHW) reactor constitute the major objective of this thesis. The use of Low Enriched Uranium fuel in the present CANDU nuclear power generating stations is recognized as economically beneficial due to reduced fuelling costs. The LEU fuel cycle is also recognized as a stepping stone to transit from the present CANDU-PHW once-through natural Uranium cycle to advanced cycles such as those based on Plutonium recycle, once-through Th + U-235 cycle, Thorium with Uranium recycle and net U-235 feed, Thorium with Uranium recycle and Plutonium feed. However, although the use of Low Enriched Uranium in the present CANDU-PHW reactor has economic advantages, and it would act as a technical bridge between the present cycle and advanced cycles, technical problems in different areas of reactor physics and fuel management were anticipated. The present thesis research work adresses the areas of reactor physics, fuel management, and control (in particular, the spatial control of large CANDU-PHW reactors). The main conclusions that have been drawn following these studies are as follows: (1) The Low Enriched Uranium Cycle is feasible in a CANDU-PHW reactor of present design and provided that: (a) The enrichment is kept relatively low (that is, about 1% instead of 0.711%); (b) the number of bundles to be replaced at every refuelling operation is about one-half that of the natural Uranium fuel case; (c) The channels are refuelled in the same direction as the coolant. (2) The response of an LEU-CANDU-PHW reactor to reactivity perturbation such as single- and two-channel refuelling operation, shim transient, shutdown-start-up transient with enrichment levels of 0.9% and 1.2% is essentially very similar {provided that certain conditions in (1) are respected} to that of the natural uranium reactor core case without any reactor reoptimization. The general behaviour of the reactor

  4. Supercritical Brayton Cycle Nuclear Power System Concepts

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.

    2007-01-01

    Both the NASA and DOE have programs that are investigating advanced power conversion cycles for planetary surface power on the moon or Mars, and for next generation nuclear power plants on earth. The gas Brayton cycle offers many practical solutions for space nuclear power systems and was selected as the nuclear power system of choice for the NASA Prometheus project. An alternative Brayton cycle that offers high efficiency at a lower reactor coolant outlet temperature is the supercritical Brayton cycle (SCBC). The supercritical cycle is a true Brayton cycle because it uses a single phase fluid with a compressor inlet temperature that is just above the critical point of the fluid. This paper describes the use of a supercritical Brayton cycle that achieves a cycle efficiency of 26.6% with a peak coolant temperature of 750 K and for a compressor inlet temperature of 390 K. The working fluid uses a clear odorless, nontoxic refrigerant C318 perflurocarbon (C4F8) that always operates in the gas phase. This coolant was selected because it has a critical temperature and pressure of 388.38 K and 2.777 MPa. The relatively high critical temperature allows for efficient thermal radiation that keeps the radiator mass small. The SCBC achieves high efficiency because the loop design takes advantage of the non-ideal nature of the coolant equation of state just above the critical point. The lower coolant temperature means that metal fuels, uranium oxide fuels, and uranium zirconium hydride fuels with stainless steel, ferretic steel, or superalloy cladding can be used with little mass penalty or reduction in cycle efficiency. The reactor can use liquid-metal coolants and no high temperature heat exchangers need to be developed. Indirect gas cooling or perhaps even direct gas cooling can be used if the C4F8 coolant is found to be sufficiently radiation tolerant. Other fluids can also be used in the supercritical Brayton cycle including Propane (C3H8, Tcritical = 369 K) and Hexane (C6

  5. Supercritical Brayton Cycle Nuclear Power System Concepts

    SciTech Connect

    Wright, Steven A.

    2007-01-30

    Both the NASA and DOE have programs that are investigating advanced power conversion cycles for planetary surface power on the moon or Mars, and for next generation nuclear power plants on earth. The gas Brayton cycle offers many practical solutions for space nuclear power systems and was selected as the nuclear power system of choice for the NASA Prometheus project. An alternative Brayton cycle that offers high efficiency at a lower reactor coolant outlet temperature is the supercritical Brayton cycle (SCBC). The supercritical cycle is a true Brayton cycle because it uses a single phase fluid with a compressor inlet temperature that is just above the critical point of the fluid. This paper describes the use of a supercritical Brayton cycle that achieves a cycle efficiency of 26.6% with a peak coolant temperature of 750 K and for a compressor inlet temperature of 390 K. The working fluid uses a clear odorless, nontoxic refrigerant C318 perflurocarbon (C4F8) that always operates in the gas phase. This coolant was selected because it has a critical temperature and pressure of 388.38 K and 2.777 MPa. The relatively high critical temperature allows for efficient thermal radiation that keeps the radiator mass small. The SCBC achieves high efficiency because the loop design takes advantage of the non-ideal nature of the coolant equation of state just above the critical point. The lower coolant temperature means that metal fuels, uranium oxide fuels, and uranium zirconium hydride fuels with stainless steel, ferretic steel, or superalloy cladding can be used with little mass penalty or reduction in cycle efficiency. The reactor can use liquid-metal coolants and no high temperature heat exchangers need to be developed. Indirect gas cooling or perhaps even direct gas cooling can be used if the C4F8 coolant is found to be sufficiently radiation tolerant. Other fluids can also be used in the supercritical Brayton cycle including Propane (C3H8, Tcritical = 369 K) and Hexane (C6

  6. A saw-less direct conversion long term evolution receiver with 25% duty-cycle LO in 130 nm CMOS technology

    NASA Astrophysics Data System (ADS)

    Siyuan, He; Changhong, Zhang; Liang, Tao; Weifeng, Zhang; Longyue, Zeng; Wei, Lü; Haijun, Wu

    2013-03-01

    A CMOS long-term evolution (LTE) direct convert receiver that eliminates the interstage SAW filter is presented. The receiver consists of a low noise variable gain transconductance amplifier (TCA), a quadrature passive current commutating mixer with a 25% duty-cycle LO, a trans-impedance amplifier (TIA), a 7th-order Chebyshev filter and programmable gain amplifiers (PGAs). A wide dynamic gain range is allocated in the RF and analog parts. A current commutating passive mixer with a 25% duty-cycle LO improves gain, noise, and linearity. An LPF based on a Tow-Thomas biquad suppresses out-of-band interference. Fabricated in a 0.13 μm CMOS process, the receiver chain achieves a 107 dB maximum voltage gain, 2.7 dB DSB NF (from PAD port), -11 dBm IIP3, and > +65 dBm IIP2 after calibration, 96 dB dynamic control range with 1 dB steps, less than 2% error vector magnitude (EVM) from 2.3 to 2.7 GHz. The total receiver (total I Q path) draws 89 mA from a 1.2-V LDO on chip supply.

  7. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1962-12-18

    A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

  8. REACTOR CONTROL SYSTEM

    DOEpatents

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

    1960-05-10

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

  9. Evaluating Environmental, Health and Safety Impacts from Two Nuclear Fuel Cycles: A Comparative Analysis of Once-Through Uranium Use and Plutonium Recycle in Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Smith, Bethan L.

    The work presented in this dissertation represents a systems-level approach to investigate potential net impacts with respect to human health and the environment associated with transitioning to the MOC for the U.S. In Chapter 2, an updated systems-level conceptual model of the OTC is presented to more accurately portray the OTC as currently implemented in the U.S. The conceptual model is the basis for estimating the worker collective doses at each operational stage, and the first demonstration of a quantitative comparative radiological impact assessment from expected normal operations is presented. In the course of evaluating worker collective dose associated with modern OTC practices, it was found that the relative contributions from the two grouped operations (front-end operations for preparing reactor fuel and reactor operations) were substantially different from historical data and conventional wisdom. As a bookend to Chapter 2, a summary is provided that describes the nature of the differences and factors that led to these differences. Detailed information of the work as part of the published journal article based off of this corollary work is included as an Appendix (C). In Chapter 3, the study of worker collective doses from the phased introduction of reprocessing in the MOC scenario, and is presented similarly to the results in Chapter 2. MOC performance was also estimated by evaluating the radioactive waste generated that can be disposed and managed through known disposal practices in shallow-land burial. Relative to the OTC, MOC performance with respect to worker collective dose was not discernibly different; while the volume of radioactive waste generated decreased. It was found that although the sheer volume of radioactive waste avoided is large, the waste disposition pathway is known for the majority of this waste. The radioactive waste that requires disposal at a licensed off-site facility is examined in closer detail. The verification process for

  10. Direct observation of the redistribution of sulfur and polysufides in Li-S batteries during first cycle by in situ X-Ray fluorescence microscopy

    DOE PAGES

    Yu, Xiquian; Pan, Huilin; Zhou, Yongning; Northrup, Paul; Xiao, Jie; Bak, Seongmin; Liu, Mingzhao; Nam, Kyung-Wan; Qu, Deyang; Liu, Jun; et al

    2015-03-25

    ). The applications of these characterization techniques have demonstrated their power in probing the structure changes, morphology evolutions, and coordination of sulfur and polysulfides with the electrolyte in Li–S cells, providing complementary information to each other thus enhancing the understanding in Li–S battery systems. In this communication, in situ X-ray fluorescence (XRF) microscopy was combined with XAS to directly probe the morphology changes of Li–S batteries during first cycle. The morphology changes of the sulfur electrode and the redistribution of sulfur and polysulfides were monitored in real time through the XRF images, while the changes of the sulfur containing compounds were characterized through the XAS spectra simultaneously. In contrast to other studies using ex situ or single characterization technique as reported in the literatures, the in situ technique used in this work has the unique feature of probing the Li–S cell under operating conditions, as well as the combination of XRF imaging with spectroscopy data. By doing this, the morphology evolution and redistribution of specific sulfur particles during cycling can be tracked and identified at certain locations in a real time. In addition, this technique allows us to select the field-of-view (FOV) area from micrometer to centimeter size, providing the capability to study the Li–S reactions not just at the material level, but also at the electrode level. This is very important for both understanding Li–S chemistry and designing effective strategies for Li–S batteries.« less

  11. Direct observation of the redistribution of sulfur and polysufides in Li-S batteries during first cycle by in situ X-Ray fluorescence microscopy

    SciTech Connect

    Yu, Xiquian; Pan, Huilin; Zhou, Yongning; Northrup, Paul; Xiao, Jie; Bak, Seongmin; Liu, Mingzhao; Nam, Kyung-Wan; Qu, Deyang; Liu, Jun; Wu, Tianpin; Yang, Xiao-Qing

    2015-03-25

    ). The applications of these characterization techniques have demonstrated their power in probing the structure changes, morphology evolutions, and coordination of sulfur and polysulfides with the electrolyte in Li–S cells, providing complementary information to each other thus enhancing the understanding in Li–S battery systems. In this communication, in situ X-ray fluorescence (XRF) microscopy was combined with XAS to directly probe the morphology changes of Li–S batteries during first cycle. The morphology changes of the sulfur electrode and the redistribution of sulfur and polysulfides were monitored in real time through the XRF images, while the changes of the sulfur containing compounds were characterized through the XAS spectra simultaneously. In contrast to other studies using ex situ or single characterization technique as reported in the literatures, the in situ technique used in this work has the unique feature of probing the Li–S cell under operating conditions, as well as the combination of XRF imaging with spectroscopy data. By doing this, the morphology evolution and redistribution of specific sulfur particles during cycling can be tracked and identified at certain locations in a real time. In addition, this technique allows us to select the field-of-view (FOV) area from micrometer to centimeter size, providing the capability to study the Li–S reactions not just at the material level, but also at the electrode level. This is very important for both understanding Li–S chemistry and designing effective strategies for Li–S batteries.

  12. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1961-09-01

    A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

  13. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  14. Utilisation of thorium in reactors

    NASA Astrophysics Data System (ADS)

    Anantharaman, K.; Shivakumar, V.; Saha, D.

    2008-12-01

    India's nuclear programme envisages a large-scale utilisation of thorium, as it has limited deposits of uranium but vast deposits of thorium. The large-scale utilisation of thorium requires the adoption of closed fuel cycle. The stable nature of thoria and the radiological issues associated with thoria poses challenges in the adoption of a closed fuel cycle. A thorium fuel based Advanced Heavy Water Reactor (AHWR) is being planned to provide impetus to development of technologies for the closed thorium fuel cycle. Thoria fuel has been loaded in Indian reactors and test irradiations have been carried out with (Th-Pu) MOX fuel. Irradiated thorium assemblies have been reprocessed and the separated 233U fuel has been used for test reactor KAMINI. The paper highlights the Indian experience with the use of thorium and brings out various issues associated with the thorium cycle.

  15. Evaluation and Optimization of a Supercritical Carbon Dioxide Power Conversion Cycle for Nuclear Applications

    SciTech Connect

    Edwin A. Harvego; Michael G. McKellar

    2011-05-01

    There have been a number of studies involving the use of gases operating in the supercritical mode for power production and process heat applications. Supercritical carbon dioxide (CO2) is particularly attractive because it is capable of achieving relatively high power conversion cycle efficiencies in the temperature range between 550°C and 750°C. Therefore, it has the potential for use with any type of high-temperature nuclear reactor concept, assuming reactor core outlet temperatures of at least 550°C. The particular power cycle investigated in this paper is a supercritical CO2 Recompression Brayton Cycle. The CO2 Recompression Brayton Cycle can be used as either a direct or indirect power conversion cycle, depending on the reactor type and reactor outlet temperature. The advantage of this cycle when compared to the helium Brayton Cycle is the lower required operating temperature; 550°C versus 850°C. However, the supercritical CO2 Recompression Brayton Cycle requires an operating pressure in the range of 20 MPa, which is considerably higher than the required helium Brayton cycle operating pressure of 8 MPa. This paper presents results of analyses performed using the UniSim process analyses software to evaluate the performance of the supercritical CO2 Brayton Recompression Cycle for different reactor outlet temperatures. The UniSim model assumed a 600 MWt reactor power source, which provides heat to the power cycle at a maximum temperature of between 550°C and 750°C. The UniSim model used realistic component parameters and operating conditions to model the complete power conversion system. CO2 properties were evaluated, and the operating range for the cycle was adjusted to take advantage of the rapidly changing conditions near the critical point. The UniSim model was then optimized to maximize the power cycle thermal efficiency at the different maximum power cycle operating temperatures. The results of the analyses showed that power cycle thermal

  16. Particle bed reactor modeling

    NASA Technical Reports Server (NTRS)

    Sapyta, Joe; Reid, Hank; Walton, Lew

    1993-01-01

    The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

  17. Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

    PubMed

    Huber, Patrick; Jaffke, Patrick

    2016-03-25

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle. PMID:27058075

  18. Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

    PubMed

    Huber, Patrick; Jaffke, Patrick

    2016-03-25

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  19. Neutron Capture and the Antineutrino Yield from Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Huber, Patrick; Jaffke, Patrick

    2016-03-01

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ˜0.9 % of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  20. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  1. CONVECTION REACTOR

    DOEpatents

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  2. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    SciTech Connect

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  3. Hydrolysis of CuCl{sub 2} in the Cu-Cl thermochemical cycle for hydrogen production : experimental studies using a spray reactor with an ultrasonic atomizer.

    SciTech Connect

    Ferrandon, M. S.; Lewis, M. A.; Alvarez, F.; Shafirovich, E.; Chemical Sciences and Engineering Division; Univ. of Texas at El Paso

    2010-03-01

    The Cu-Cl thermochemical cycle is being developed as a hydrogen production method. Prior proof-of-concept experimental work has shown that the chemistry is viable while preliminary modeling has shown that the efficiency and cost of hydrogen production have the potential to meet DOE's targets. However, the mechanisms of CuCl{sub 2} hydrolysis, an important step in the Cu-Cl cycle, are not fully understood. Although the stoichiometry of the hydrolysis reaction, 2CuCl{sub 2} + H{sub 2}O {leftrightarrow} Cu{sub 2}OCl{sub 2} + 2HCl, indicates a necessary steam-to-CuCl{sub 2} molar ratio of 0.5, a ratio as high as 23 has been typically required to obtain near 100% conversion of the CuCl{sub 2} to the desired products at atmospheric pressure. It is highly desirable to conduct this reaction with less excess steam to improve the process efficiency. Per Le Chatelier's Principle and according to the available equilibrium-based model, the needed amount of steam can be decreased by conducting the hydrolysis reaction at a reduced pressure. In the present work, the experimental setup was modified to allow CuCl{sub 2} hydrolysis in the pressure range of 0.4-1 atm. Chemical and XRD analyses of the product compositions revealed the optimal steam-to-CuCl{sub 2} molar ratio to be 20-23 at 1 atm pressure. The experiments at 0.4 atm and 0.7 atm showed that it is possible to lower the steam-to-CuCl{sub 2} molar ratio to 15, while still obtaining good yields of the desired products. An important effect of running the reaction at reduced pressure is the significant decrease of CuCl concentration in the solid products, which was not predicted by prior modeling. Possible explanations based on kinetics and residence times are suggested.

  4. DAX1, a direct target of EWS/FLI1 oncoprotein, is a principal regulator of cell-cycle progression in Ewing's tumor cells.

    PubMed

    García-Aragoncillo, E; Carrillo, J; Lalli, E; Agra, N; Gómez-López, G; Pestaña, A; Alonso, J

    2008-10-01

    The molecular hallmark of the Ewing's family of tumors is the presence of balanced chromosomal translocations, leading to the formation of chimerical transcription factors (that is, EWS/FLI1) that play a pivotal role in the pathogenesis of Ewing's tumors by deregulating gene expression. We have recently demonstrated that DAX1 (NR0B1), an orphan nuclear receptor that was not previously implicated in cancer, is induced by the EWS/FLI1 oncoprotein and is highly expressed in Ewing's tumors, suggesting that DAX1 is a biologically relevant target of EWS/FLI1-mediated oncogenesis. In this study we demonstrate that DAX1 is a direct transcriptional target of the EWS/FLI1 oncoprotein through its binding to a GGAA-rich region in the DAX1 promoter and show that DAX1 is a key player of EWS/FLI1-mediated oncogenesis. DAX1 silencing using an inducible model of RNA interference induces growth arrest in the A673 Ewing's cell line and severely impairs its capability to grow in semisolid medium and form tumors in immunodeficient mice. Gene expression profile analysis demonstrated that about 10% of the genes regulated by EWS/FLI1 in Ewing's cells are DAX1 targets, confirming the importance of DAX1 in Ewing's oncogenesis. Functional genomic analysis, validated by quantitative RT-PCR, showed that genes implicated in cell-cycle progression, such as CDK2, CDC6, MCM10 or SKP2 were similarly regulated by EWS/FLI1 and DAX1. These findings indicate that DAX1 is important in the pathogenesis of the Ewing's family of tumors, identify new functions for DAX1 as a cell-cycle progression regulator and open the possibility to new therapeutic approaches based on DAX1 function interference.

  5. DAX1, a direct target of EWS/FLI1 oncoprotein, is a principal regulator of cell-cycle progression in Ewing's tumor cells.

    PubMed

    García-Aragoncillo, E; Carrillo, J; Lalli, E; Agra, N; Gómez-López, G; Pestaña, A; Alonso, J

    2008-10-01

    The molecular hallmark of the Ewing's family of tumors is the presence of balanced chromosomal translocations, leading to the formation of chimerical transcription factors (that is, EWS/FLI1) that play a pivotal role in the pathogenesis of Ewing's tumors by deregulating gene expression. We have recently demonstrated that DAX1 (NR0B1), an orphan nuclear receptor that was not previously implicated in cancer, is induced by the EWS/FLI1 oncoprotein and is highly expressed in Ewing's tumors, suggesting that DAX1 is a biologically relevant target of EWS/FLI1-mediated oncogenesis. In this study we demonstrate that DAX1 is a direct transcriptional target of the EWS/FLI1 oncoprotein through its binding to a GGAA-rich region in the DAX1 promoter and show that DAX1 is a key player of EWS/FLI1-mediated oncogenesis. DAX1 silencing using an inducible model of RNA interference induces growth arrest in the A673 Ewing's cell line and severely impairs its capability to grow in semisolid medium and form tumors in immunodeficient mice. Gene expression profile analysis demonstrated that about 10% of the genes regulated by EWS/FLI1 in Ewing's cells are DAX1 targets, confirming the importance of DAX1 in Ewing's oncogenesis. Functional genomic analysis, validated by quantitative RT-PCR, showed that genes implicated in cell-cycle progression, such as CDK2, CDC6, MCM10 or SKP2 were similarly regulated by EWS/FLI1 and DAX1. These findings indicate that DAX1 is important in the pathogenesis of the Ewing's family of tumors, identify new functions for DAX1 as a cell-cycle progression regulator and open the possibility to new therapeutic approaches based on DAX1 function interference. PMID:18591936

  6. Aperiodicity resulting from two-cycle coupling in the Belousov-Zhabotinskii reaction. III. Analysis of a model of the effect of spatial inhomogeneities at the input ports of a continuous-flow, stirred tank reactor

    NASA Astrophysics Data System (ADS)

    Györgyi, László; Field, Richard J.

    1989-11-01

    Deterministic chaos is a well-established phenomenon in continuous-flow, stirred tank reactor (CSTR) experiments with the oscillatory Belousov-Zhabotinskii (BZ) reaction. However, it has not yet been possible to reproduce the experimentally observed, robust chaos in simulations using realistic models of the homogeneous chemical kinetics of the BZ reaction. That it may be necessary to consider spatial inhomogeneities in modeling the BZ chaos is suggested by the existence of strong stirring effects on the aperiodic behavior and by the difficulty of reproducing chaos under the same conditions in reactors of different physical configuration. Such inhomogeneity might result from a lack of perfect mixing in the CSTR, especially near the inlets, or from diffusion of species at low flow rates from the CSTR reaction mixture into the tips of the inlets. The presence of spatial inhomogeneities allows coupling between essentially independent oscillators, a well-known source of chaos. Such a model using a realistic representation of the BZ kinetics leads to an eight-variable set of ordinary differential equations. Numerical analysis of these equations by continuation methods and by numerical integration shows the existence of broad regions of chaos and various hysteresis effects involving limit cycles, a steady state and/or a strange attractor. Tristability was found in calculations in a narrow flow rate range. The computed behavior appears for parameter values closely related to the values used experimentally to obtain similar phenomena, and the visual similarity of the computed and experimental strange attractors is striking. The experimental routes to chaos, period doubling, intermittency, and secondary Hopf bifurcations are all reproduced in the simulations. The computed and experimental structures of periodic windows observed within chaotic regions also are very similar.

  7. UF6 breeder reactor power plants for electric power generation

    NASA Technical Reports Server (NTRS)

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

    1976-01-01

    The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

  8. Nuclear reactor

    DOEpatents

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  9. Physics challenges for advanced fuel cycle assessment

    SciTech Connect

    Giuseppe Palmiotti; Massimo Salvatores; Gerardo Aliberti

    2014-06-01

    Advanced fuel cycles and associated optimized reactor designs will require substantial improvements in key research area to meet new and more challenging requirements. The present paper reviews challenges and issues in the field of reactor and fuel cycle physics. Typical examples are discussed with, in some cases, original results.

  10. Slurry reactor design studies

    SciTech Connect

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  11. Thermal and hydraulic performance tests of a sieve-tray direct-contact heat exchanger vaporizing pure and mixed-hydrocarbon Rankine-cycle working fluids

    SciTech Connect

    Mines, G.L.; Demuth, O.J.; Wiggins, D.J.

    1983-08-01

    Experiments investigating a sieve-tray direct-contact heat exchanger were conducted at the Raft River Geothermal Test Site in southeastern Idaho using the 60-kW Mobile Heat Cycle Research Facility operating in the thermal loop mode (without a turbine). Isobutane, propane, and several hydrocarbon mixtures were heated and boiled in the direct-contact column, which is approx. 12 in. in diameter and 19-1/2 ft. high, using the energy from a 280/sup 0/F geothermal resource. Using pure fluids, isobutane or propane, the column operated much as intended, with 17 trays used for preheating and one or two accomplishing the boiling. For the pure fluids, individual tray efficiencies were found to be 70% or higher for preheating, and close to 100% for boiling; column pinch points were projected to be well under 1/sup 0/F with some runs reaching values as low as approx. 0.02/sup 0/F. Maximum geofluid throughputs for the isobutane tests corresponded roughly to the terminal rise velocity of a 1/32 in. working fluid droplet in geofluid. Boiling was found to occur in as many as 12 trays for the mixtures having the highest concentrations of the minor component, with overall efficiencies in the boiling section estimated on the order of 25 or 30%. Preheating tray efficiencies appeared to be fairly independent of working fluid, with pinch points ranging from as low as approx. 0.03/sup 0/F for a 0.95 isobutane/0.05 hexane mixture to approx. 2.3/sup 0/F for a 0.85 isobutane/0.05 hexane mixture. Column operation was noticeably less stable for the mixtures than for the pure fluids, with maximum throughputs dropping to as low as 40 to 50% of those for the pure fluids.

  12. Thermochemical cycles

    NASA Technical Reports Server (NTRS)

    Funk, J. E.; Soliman, M. A.; Carty, R. H.; Conger, W. L.; Cox, K. E.; Lawson, D.

    1975-01-01

    The thermochemical production of hydrogen is described along with the HYDRGN computer program which attempts to rate the various thermochemical cycles. Specific thermochemical cycles discussed include: iron sulfur cycle; iron chloride cycle; and hybrid sulfuric acid cycle.

  13. Coupled modeling of a directly heated tubular solar receiver for supercritical carbon dioxide Brayton cycle: Optical and thermal-fluid evaluation

    DOE PAGES

    Ortega, Jesus; Khivsara, Sagar; Christian, Joshua; Ho, Clifford; Yellowhair, Julius; Dutta, Pradip

    2016-05-30

    In single phase performance and appealing thermo-physical properties supercritical carbon dioxide (s-CO2) make a good heat transfer fluid candidate for concentrating solar power (CSP) technologies. The development of a solar receiver capable of delivering s-CO2 at outlet temperatures ~973 K is required in order to merge CSP and s-CO2 Brayton cycle technologies. A coupled optical and thermal-fluid modeling effort for a tubular receiver is undertaken to evaluate the direct tubular s-CO2 receiver’s thermal performance when exposed to a concentrated solar power input of ~0.3–0.5 MW. Ray tracing, using SolTrace, is performed to determine the heat flux profiles on the receivermore » and computational fluid dynamics (CFD) determines the thermal performance of the receiver under the specified heating conditions. Moreover, an in-house MATLAB code is developed to couple SolTrace and ANSYS Fluent. CFD modeling is performed using ANSYS Fluent to predict the thermal performance of the receiver by evaluating radiation and convection heat loss mechanisms. Understanding the effects of variation in heliostat aiming strategy and flow configurations on the thermal performance of the receiver was achieved through parametric analyses. Finally, a receiver thermal efficiency ~85% was predicted and the surface temperatures were observed to be within the allowable limit for the materials under consideration.« less

  14. NEUTRONIC REACTOR

    DOEpatents

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  15. REACTOR COOLING

    DOEpatents

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  16. Solar hydrogen by thermochemical water splitting cycles: design, modeling, and demonstration of a novel receiver/reactor for the high temperature decomposition of zno using concentrated sunlight

    NASA Astrophysics Data System (ADS)

    Kaiser, Zachary David Epping

    Documenting the presence of rare bat species can be difficult. The current summer survey protocol for the federally endangered Indiana bat ( Myotis sodalis) requires passive acoustic sampling with directional microphones (e.g., Anabats), but there are still questions about best practices for choosing survey sites and appropriate detector models. Indiana bats are capable of foraging in an array of cover types, including structurally-complex, interior forests. Further, data acquisition among different commercially available bat detectors is likely highly variable, due to the use of proprietary microphones with different frequency responses, sensitivities, and directionality. We paired omnidirectional Wildlife Acoustic SM2BAT+ (SM2) and directional Titley Scientific Anabat SD2 (Anabat) detectors at 71 random points near Indianapolis, Indiana from May-August 2012-2013 to compare data acquisition by phonic group (low, mid, Myotis) and to determine what factors affect probability of detection and site occupancy for Indiana bats when sampling with acoustics near an active maternity colony (0.20--8.39 km away). Weatherproofing for Anabat microphones was 45° angle PVC tubes and for SM2 microphones was their foam shielding; microphones were paired at 2 m and 5 m heights. Habitat and landscape covariates were measured in the field or via ArcGIS. We adjusted file parameters to make SM2 and Anabat data comparable. Files were identified using Bat Call ID software, with visual inspection of Indiana bat calls. The effects of detector type, phonic group, height, and their interactions on mean files recorded per site were assessed using generalized estimating equations and LSD pairwise comparisons. We reduced probability of detection (p) and site occupancy (ψ) model covariates with Pearson's correlation and PCA. We used Presence 6.1 software and Akaike's Information Criteria to assess models for p and ψ. Anabats and SM2s did not perform equally. Anabats recorded more low and

  17. Advanced burner test reactor preconceptual design report.

    SciTech Connect

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  18. Sequences contained within the promoter of the human thymidine kinase gene can direct cell-cycle regulation of heterologous fusion genes.

    PubMed Central

    Kim, Y K; Wells, S; Lau, Y F; Lee, A S

    1988-01-01

    Recent evidence on the transcriptional regulation of the human thymidine kinase (TK) gene raises the possibility that cell-cycle regulatory sequences may be localized within its promoter. A hybrid gene that combines the TK 5' flanking sequence and the coding region of the bacterial neomycin-resistance gene (neo) has been constructed. Upon transfection into a hamster fibroblast cell line K12, the hybrid gene exhibits cell-cycle-dependent expression. Deletion analysis reveals that the region important for cell-cycle regulation is within -441 to -63 nucleotides from the transcriptional initiation site. This region (-441 to -63) also confers cell-cycle regulation to the herpes simplex virus thymidine kinase (HSVtk) promoter, which is not expressed in a cell-cycle manner. We conclude that the -441 to -63 sequence within the human TK promoter is important for cell-cycle-dependent expression. Images PMID:3413063

  19. Effect of thermal cycling in a Mach 0.3 burner rig on properties and structure of directionally solidified gamma/gamma prime - delta eutectic

    NASA Technical Reports Server (NTRS)

    Gray, H. R.; Sanders, W. A.

    1975-01-01

    Tensile and stress rupture properties at 1040 C of a thermally cycled gamma/gamma prime - delta eutectic were essentially equivalent to the as-grown properties. Tensile strength and rupture life at 760 C appeared to decrease slightly by thermal cycling. Thermal cycling resulted in gamma prime coarsening and Widmanstatten delta precipitation in the gamma phase. An unidentified precipitate, presumably gamma prime, was observed within the delta phase. The eutectic alloy exhibited a high rate of oxidation-erosion weight loss during thermal cycling in the Mach 0.3 burner rig.

  20. Calculation of heating values for the high flux isotope reactor

    SciTech Connect

    Peterson, J.; Ilas, G.

    2012-07-01

    Calculating the amount of energy released by a fission reaction (fission Q value) and the heating rate distribution in a nuclear reactor is an important part of the safety analysis. However, these calculations can become very complex. One of the codes that can be used for this type of analyses is the Monte Carlo transport code MCNP5. Currently it is impossible to calculate the Q value and heating rate disposition for delayed beta and delayed gamma particles directly from MCNP5. The purpose of this paper is to outline a rigorous method for indirectly calculating the Q values and heating rates in the High Flux Isotope Reactor (HFIR), based on previous similar studies carried out for very high-temperature reactor configurations. This method has been applied in this study to calculate heating rates for the beginning of cycle (BOC) and end-of-cycle (EOC) states of HFIR. In addition, the BOC results obtained for HFIR are compared with corresponding results for the Advanced Test Reactor. The fission Q value for HFIR was calculated as 200.2 MeV for the BOC and 201.3 MeV for the EOC. It was also determined that 95.1% and 95.4% of the heat was deposited within the HFIR fuel plates for the BOC and EOC models, respectively. This methodology can also be used for heating rate calculations for HFIR experiments. (authors)

  1. Calculation of Heating Values for the High Flux Isotope Reactor

    SciTech Connect

    Peterson, Joshua L; Ilas, Germina

    2012-01-01

    Calculating the amount of energy released by a fission reaction (fission Q value) and the heating rate distribution in a nuclear reactor is an important part of the safety analysis. However, these calculations can become very complex. One of the codes that can be used for this type of analyses is the Monte Carlo transport code MCNP5. Currently it is impossible to calculate the Q value and heating rate disposition for delayed beta and delayed gamma particles directly from MCNP5. The purpose of this paper is to outline a rigorous method for indirectly calculating the Q values and heating rates in the High Flux Isotope Reactor (HFIR), based on previous similar studies carried out for very high-temperature reactor configurations. This method has been applied in this study to calculate heating rates for the beginning of cycle (BOC) and end-of-cycle (EOC) states of HFIR. In addition, the BOC results obtained for HFIR are compared with corresponding results for the Advanced Test Reactor. The fission Q value for HFIR was calculated as 200.2 MeV for the BOC and 201.3 MeV for the EOC. It was also determined that 95.1% and 95.4% of the heat was deposited within the HFIR fuel plates for the BOC and EOC models, respectively. This methodology can also be used for heating rate calculations for HFIR experiments.

  2. C-CAMP, A closed cycle alkali metal power system

    SciTech Connect

    Wichner, R.P.; Hoffman, H.W.

    1988-01-01

    A concept is presented for a Closed-Cycle Alkali Metal (C-CAMP) power systems which utilizes the heat of reaction of an alkali metal and halogen compound to vaporize an alkali metal turbine fluid for a Rankine cycle. Unique features of the concept are (1) direct contact (heat exchange) between the reaction products and turbine fluid, and (2) a flow-through chemical reactor/boiler. The principal feasibility issues of the concept relate to the degree of cross-mixing of product and turbine fluid streams within the reactor-boiler. If proven feasible, the concept may be adapted to a range of fuel and turbine fluids and ultimately lead to thermal efficiencies in excess of 35%.

  3. The role of accelerators in the nuclear fuel cycle

    SciTech Connect

    Takahashi, Hiroshi.

    1990-01-01

    The use of neutrons produced by the medium energy proton accelerator (1 GeV--3 GeV) has considerable potential in reconstructing the nuclear fuel cycle. About 1.5 {approximately} 2.5 ton of fissile material can be produced annually by injecting a 450 MW proton beam directly into fertile materials. A source of neutrons, produced by a proton beam, to supply subcritical reactors could alleviate many of the safety problems associated with critical assemblies, such as positive reactivity coefficients due to coolant voiding. The transient power of the target can be swiftly controlled by controlling the power of the proton beam. Also, the use of a proton beam would allow more flexibility in the choice of fuel and structural materials which otherwise might reduce the reactivity of reactors. This paper discusses the rate of accelerators in the transmutation of radioactive wastes of the nuclear fuel cycles. 34 refs., 17 figs., 9 tabs.

  4. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  5. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.; Johnson, H.W.

    1961-04-01

    BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

  6. Reactor building

    SciTech Connect

    Hista, J. C.

    1984-09-18

    Reactor building comprising a vessel shaft anchored in a slab which is peripherally locked. This reactor building comprises a confinement enclosure within which are positioned internal structures constituted by an internal structure floor, a vessel shaft, a slab being positioned between the general floor and the internal structure floor, the vesse

  7. 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.

  8. Moon base reactor system

    NASA Technical Reports Server (NTRS)

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

    1989-01-01

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

  9. Method and system to directly produce electrical power within the lithium blanket region of a magnetically confined, deuterium-tritium (DT) fueled, thermonuclear fusion reactor

    DOEpatents

    Woolley, Robert D.

    1999-01-01

    A method for integrating liquid metal magnetohydrodynamic power generation with fusion blanket technology to produce electrical power from a thermonuclear fusion reactor located within a confining magnetic field and within a toroidal structure. A hot liquid metal flows from a liquid metal blanket region into a pump duct of an electromagnetic pump which moves the liquid metal to a mixer where a gas of predetermined pressure is mixed with the pressurized liquid metal to form a Froth mixture. Electrical power is generated by flowing the Froth mixture between electrodes in a generator duct. When the Froth mixture exits the generator the gas is separated from the liquid metal and both are recycled.

  10. Method and System to Directly Produce Electrical Power within the Lithium Blanket Region of a Magnetically Confined, Deuterium-Tritium (DT) Fueled, Thermonuclear Fusion Reactor

    SciTech Connect

    Woolley, Robert D.

    1998-09-22

    A method for integrating liquid metal magnetohydrodynamic power generation with fusion blanket technology to produce electrical power from a thermonuclear fusion reactor located within a confining magnetic field and within a toroidal structure. A hot liquid metal flows from a liquid metal blanket region into a pump duct of an electromagnetic pump which moves the liquid metal to a mixer where a gas of predetermined pressure is mixed with the pressurized liquid metal to form a Froth mixture. Electrical power is generated by flowing the Froth mixture between electrodes in a generator duct. When the Froth mixture exits the generator the gas is separated from the liquid metal and both are recycled.

  11. Colliding Beam Fusion Reactor Space Propulsion System

    NASA Astrophysics Data System (ADS)

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

    2004-02-01

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

  12. Enhancement of the anaerobic hydrolysis and fermentation of municipal solid waste in leachbed reactors by varying flow direction during water addition and leachate recycle.

    PubMed

    Uke, Matthew N; Stentiford, Edward

    2013-06-01

    Poor performance of leachbed reactors (LBRs) is attributed to channelling, compaction from waste loading, unidirectional water addition and leachate flow causing reduced hydraulic conductivity and leachate flow blockage. Performance enhancement was evaluated in three LBRs M, D and U at 22 ± 3°C using three water addition and leachate recycle strategies; water addition was downflow in D throughout, intermittently upflow and downflow in M and U with 77% volume downflow in M, 54% volume downflow in U while the rest were upflow. Leachate recycle was downflow in D, alternately downflow and upflow in M and upflow in U. The strategy adopted in U led to more water addition (30.3%), leachate production (33%) and chemical oxygen demand (COD) solubilisation (33%; 1609 g against 1210 g) compared to D (control). The total and volatile solids (TS and VS) reductions were similar but the highest COD yield (g-COD/g-TS and g-COD/g-VS removed) was in U (1.6 and 1.9); the values were 1.33 and 1.57 for M, and 1.18 and 1.41 for D respectively. The strategy adopted in U showed superior performance with more COD and leachate production compared to reactors M and D.

  13. RADIATION FACILITY FOR NUCLEAR REACTORS

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1961-12-12

    A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

  14. Expanding the scope of CE reactor to ssDNA-binding protein-ssDNA complexes as exemplified for a tool for direct measurement of dissociation kinetics of biomolecular complexes.

    PubMed

    Takahashi, Toru; Ohtsuka, Kei-Ichirou; Tomiya, Yoriyuki; Iki, Nobuhiko; Hoshino, Hitoshi

    2009-09-01

    CE reactor (CER), which was developed as a tool for direct measurement of the dissociation kinetics of metal complexes, was successfully applied to the complexes of Escherichia coli ssDNA-binding protein (SSB) with ssDNA. The basic concept of CER is the application of CE separation process as a dissociation kinetic reactor for the complex, and the observation of the on-capillary dissociation reaction profile of the complex as the decrease of the peak height of the complex with increase of the migration time. The peak height of [SSB-ssDNA] decreases as the migration time increases since the degree of the decrease of [SSB-ssDNA] through the on-capillary dissociation reaction is proportional to the degree of the decrease of the peak height of [SSB-ssDNA]. The dissociation degree-time profiles for the complexes are quantitatively described by analyzing a set of electropherograms with different migration times. Dissociation rate constants of [SSB-ssDNA] consisting of 20-mer, 25-mer and 31-mer ssDNA were directly determined to be 3.99x10(-4), 4.82x10(-4) and 1.50x10(-3)/s, respectively. CER is a concise and effective tool for dissociation kinetic analysis of biomolecular complexes.

  15. Compact Fusion Advanced Rankine (CFARII) power cycle

    SciTech Connect

    Logan, B.G.

    1991-08-23

    The Compact Fusion Advanced Rankine (CFARII) power cycle is a direct plasma energy conversion scheme for inertial fusion (ICF) and magnetically-insulated, inertially confined fusion (MICF) reactors utilizing: (1) conversion of plasma thermal ionization and thermal energy into kinetic energy of a supersonic plasma jet, (2) conversion of the plasma jet kinetic energy into DC electricity by slowing down in an ``impulse`` type of magnetohydrodynamic (MHD) generator, and (3) condensation and heat rejection of the exhaust plasma on droplets of recirculating condensate (``raindrop`` condensor). A preliminary evaluation of a particular reference case CFARII Balance-of-Plant (BoP) is found sufficiently attractive (52% gross cycle efficiency, 40 million 1991 $ BoP for 1 GWe gross electric) to warrant further work on several design issues.

  16. Compact Fusion Advanced Rankine (CFARII) power cycle

    SciTech Connect

    Logan, B.G.

    1991-08-23

    The Compact Fusion Advanced Rankine (CFARII) power cycle is a direct plasma energy conversion scheme for inertial fusion (ICF) and magnetically-insulated, inertially confined fusion (MICF) reactors utilizing: (1) conversion of plasma thermal ionization and thermal energy into kinetic energy of a supersonic plasma jet, (2) conversion of the plasma jet kinetic energy into DC electricity by slowing down in an impulse'' type of magnetohydrodynamic (MHD) generator, and (3) condensation and heat rejection of the exhaust plasma on droplets of recirculating condensate ( raindrop'' condensor). A preliminary evaluation of a particular reference case CFARII Balance-of-Plant (BoP) is found sufficiently attractive (52% gross cycle efficiency, 40 million 1991 $ BoP for 1 GWe gross electric) to warrant further work on several design issues.

  17. Compact Reactor

    SciTech Connect

    Williams, Pharis E.

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  18. NUCLEAR REACTOR

    DOEpatents

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  19. Enhancement of the anaerobic hydrolysis and fermentation of municipal solid waste in leachbed reactors by varying flow direction during water addition and leachate recycle

    SciTech Connect

    Uke, Matthew N.; Stentiford, Edward

    2013-06-15

    Highlights: ► Combined downflow and upflow water addition improved hydraulic conductivity. ► Upflow water addition unclogged perforated screen leading to more leachate flow. ► The volume of water added and transmitted positively correlated with hydrolysis process. ► Combined downflow and upflow water addition increased COD production and yield. ► Combined downflow and upflow leachate recycle improved leachate and COD production. - Abstract: Poor performance of leachbed reactors (LBRs) is attributed to channelling, compaction from waste loading, unidirectional water addition and leachate flow causing reduced hydraulic conductivity and leachate flow blockage. Performance enhancement was evaluated in three LBRs M, D and U at 22 ± 3 °C using three water addition and leachate recycle strategies; water addition was downflow in D throughout, intermittently upflow and downflow in M and U with 77% volume downflow in M, 54% volume downflow in U while the rest were upflow. Leachate recycle was downflow in D, alternately downflow and upflow in M and upflow in U. The strategy adopted in U led to more water addition (30.3%), leachate production (33%) and chemical oxygen demand (COD) solubilisation (33%; 1609 g against 1210 g) compared to D (control). The total and volatile solids (TS and VS) reductions were similar but the highest COD yield (g-COD/g-TS and g-COD/g-VS removed) was in U (1.6 and 1.9); the values were 1.33 and 1.57 for M, and 1.18 and 1.41 for D respectively. The strategy adopted in U showed superior performance with more COD and leachate production compared to reactors M and D.

  20. Inertial fusion energy power reactor fuel recovery system

    SciTech Connect

    Gentile, C. A.; Kozub, T.; Langish, S. W.; Ciebiera, L. P.; Nobile, A.; Wermer, J.; Sessions, K.

    2008-07-15

    A conceptual design is proposed to support the recovery of un-expended fuel, ash, and associated post-detonation products resident in plasma exhaust from a {approx}2 GWIFE direct drive power reactor. The design includes systems for the safe and efficient collection, processing, and purification of plasma exhaust fuel components. The system has been conceptually designed and sized such that tritium bred within blankets, lining the reactor target chamber, can also be collected, processed, and introduced into the fuel cycle. The system will nominally be sized to process {approx}2 kg of tritium per day and is designed to link directly to the target chamber vacuum pumping system. An effort to model the fuel recovery system (FRS) using the Aspen Plus engineering code has commenced. The system design supports processing effluent gases from the reactor directly from the exhaust of the vacuum pumping system or in batch mode, via a buffer vessel in the Receiving and Analysis System. Emphasis is on nuclear safety, reliability, and redundancy as to maximize availability. The primary goal of the fuel recovery system design is to economically recycle components of direct drive IFE fuel. The FRS design is presented as a facility sub-system in the context of supporting the larger goal of producing safe and economical IFE power. (authors)

  1. On the Relationship between Solar Wind Speed, Earthward-Directed Coronal Mass Ejections, Geomagnetic Activity, and the Sunspot Cycle Using 12-Month Moving Averages

    NASA Technical Reports Server (NTRS)

    Wilson, Robert M.; Hathaway, David H.

    2008-01-01

    For 1996 .2006 (cycle 23), 12-month moving averages of the aa geomagnetic index strongly correlate (r = 0.92) with 12-month moving averages of solar wind speed, and 12-month moving averages of the number of coronal mass ejections (CMEs) (halo and partial halo events) strongly correlate (r = 0.87) with 12-month moving averages of sunspot number. In particular, the minimum (15.8, September/October 1997) and maximum (38.0, August 2003) values of the aa geomagnetic index occur simultaneously with the minimum (376 km/s) and maximum (547 km/s) solar wind speeds, both being strongly correlated with the following recurrent component (due to high-speed streams). The large peak of aa geomagnetic activity in cycle 23, the largest on record, spans the interval late 2002 to mid 2004 and is associated with a decreased number of halo and partial halo CMEs, whereas the smaller secondary peak of early 2005 seems to be associated with a slight rebound in the number of halo and partial halo CMEs. Based on the observed aaM during the declining portion of cycle 23, RM for cycle 24 is predicted to be larger than average, being about 168+/-60 (the 90% prediction interval), whereas based on the expected aam for cycle 24 (greater than or equal to 14.6), RM for cycle 24 should measure greater than or equal to 118+/-30, yielding an overlap of about 128+/-20.

  2. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production Progress Report for Year 1, Quarter 2 (January - March 2002)

    SciTech Connect

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-03-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  3. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  4. NEUTRONIC REACTOR

    DOEpatents

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  5. Chemical Reactors.

    ERIC Educational Resources Information Center

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  6. NEUTRONIC REACTOR

    DOEpatents

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  7. NUCLEAR REACTOR

    DOEpatents

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  8. REACTOR SHIELD

    DOEpatents

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  9. NUCLEAR REACTOR

    DOEpatents

    Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

    1962-10-23

    A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

  10. 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.

  11. A novel sorbent for transport reactors and fluidized bed reactors

    SciTech Connect

    Copeland, R.; Cesario, M.; Gershanovich, Y.; Sibold, J.; Windecker, B.

    1998-12-31

    Coal Fired Gasifier Combined Cycles (GCC) have both high efficiency and very low emissions. GCCs critically need a method of removing the H{sub 2}S produced from the sulfur in the coal from the hot gases. There has been extensive research on hot gas cleanup systems, focused on the use of a zinc oxide based sorbent (e.g., zinc titanate). TDA Research, Inc. (TDA) is developing a novel sorbent with improved attrition resistance for transport reactors and fluidized bed reactors. The authors are testing sorbents at conditions simulating the operating conditions of the Pinon Pine clean coal technology plant. TDA sulfided several different formulations at 538 C and found several that have high sulfur capacity when tested in a fluidized bed reactor. TDA initiated sorbent regeneration at 538 C. The sorbents retained chemical activity with multiple cycles. Additional tests will be conducted to evaluate the best sorbent formulation.

  12. NEUTRONIC REACTORS

    DOEpatents

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  13. NUCLEAR REACTOR

    DOEpatents

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  14. NUCLEAR REACTOR

    DOEpatents

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  15. Heterogeneous Transmutation Sodium Fast Reactor

    SciTech Connect

    S. E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even neutron number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both non-flattened and flattened core geometries. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of comparable size. A mass balance analysis revealed that the heterogeneous design may reduce the number of fast reactors needed to close the current once-through light water reactor fuel cycle.

  16. Principles and rationale of the Fusion-Fission Hybrid burner reactor

    NASA Astrophysics Data System (ADS)

    Stacey, Weston M.

    2012-06-01

    The potential advantages of Fusion-Fission Hybrid (FFH) reactors (relative to critical fast reactors) for closing the back end of the nuclear fuel cycle are discussed. The choices of fission and fusion technologies for FFH burner reactors that would fission the transuranics remaining in spent fuel discharged from nuclear power reactors are summarized. The conceptual design and fuel cycle performance of the SABR FFH burner reactor are presented, and a fusion power development schedule with a symbiotic dual FFH path is outlined.

  17. Irradiation Facilities at the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-12-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC – formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world’s data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities1. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens.

  18. Evaluation of Alternate Materials for Coated Particle Fuels for the Gas-Cooled Fast Reactor. Laboratory Directed Research and Development Program FY 2006 Final Report

    SciTech Connect

    Paul A. Demkowicz; Karen Wright; Jian Gan; David Petti; Todd Allen; Jake Blanchard

    2006-09-01

    Candidate ceramic materials were studied to determine their suitability as Gas-Cooled Fast Reactor particle fuel coatings. The ceramics examined in this work were: TiC, TiN, ZrC, ZrN, AlN, and SiC. The studies focused on (i) chemical reactivity of the ceramics with fission products palladium and rhodium, (ii) the thermomechanical stresses that develop in the fuel coatings from a variety of causes during burnup, and (iii) the radiation resiliency of the materials. The chemical reactivity of TiC, TiN, ZrC, and ZrN with Pd and Rh were all found to be much lower than that of SiC. A number of important chemical behaviors were observed at the ceramic-metal interfaces, including the formation of specific intermetallic phases and a variation in reaction rates for the different ceramics investigated. Based on the data collected in this work, the nitride ceramics (TiN and ZrN) exhibit chemical behavior that is characterized by lower reaction rates with Pd and Rh than the carbides TiC and ZrC. The thermomechanical stresses in spherical fuel particle ceramic coatings were modeled using finite element analysis, and included contributions from differential thermal expansion, fission gas pressure, fuel kernel swelling, and thermal creep. In general the tangential stresses in the coatings during full reactor operation are tensile, with ZrC showing the lowest values among TiC, ZrC, and SiC (TiN and ZrN were excluded from the comprehensive calculations due to a lack of available materials data). The work has highlighted the fact that thermal creep plays a critical role in the development of the stress state of the coatings by relaxing many of the stresses at high temperatures. To perform ion irradiations of sample materials, an irradiation beamline and high-temperature sample irradiation stage was constructed at the University of Wisconsin’s 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature

  19. Prometheus Reactor I&C Software Development Methodology, for Action

    SciTech Connect

    T. Hamilton

    2005-07-30

    The purpose of this letter is to submit the Reactor Instrumentation and Control (I&C) software life cycle, development methodology, and programming language selections and rationale for project Prometheus to NR for approval. This letter also provides the draft Reactor I&C Software Development Process Manual and Reactor Module Software Development Plan to NR for information.

  20. Quantification of Back-End Nuclear Fuel Cycle Metrics Uncertainties Due to Cross Sections

    SciTech Connect

    Tracy E. Stover, Jr.

    2007-11-01

    This work examines uncertainties in the back end fuel cycle metrics of isotopic composition, decay heat, radioactivity, and radiotoxicity. Most advanced fuel cycle scenarios, including the ones represented in this work, are limited by one or more of these metrics, so that quantification of them becomes of great importance in order to optimize or select one of these scenarios. Uncertainty quantification, in this work, is performed by propagating cross-section covariance data, and later number density covariance data, through a reactor physics and depletion code sequence. Propagation of uncertainty is performed primarily via the Efficient Subspace Method (ESM). ESM decomposes the covariance data into singular pairs and perturbs input data along independent directions of the uncertainty and only for the most significant values of that uncertainty. Results of these perturbations being collected, ESM directly calculates the covariance of the observed output posteriori. By exploiting the rank deficient nature of the uncertainty data, ESM works more efficiently than traditional stochastic sampling, but is shown to produce equivalent results. ESM is beneficial for very detailed models with large amounts of input data that make stochastic sampling impractical. In this study various fuel cycle scenarios are examined. Simplified, representative models of pressurized water reactor (PWR) and boiling water reactor (BWR) fuels composed of both uranium oxide and mixed oxides are examined. These simple models are intended to give a representation of the uncertainty that can be associated with open uranium oxide fuel cycles and closed mixed oxide fuel cycles. The simplified models also serve as a demonstration to show that ESM and stochastic sampling produce equivalent results, because these models require minimum computer resources and have amounts of input data small enough such that either method can be quickly implemented and a numerical experiment performed. The simplified

  1. Research reactors - an overview

    SciTech Connect

    West, C.D.

    1997-03-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  2. Accident analysis of heavy water cooled thorium breeder reactor

    NASA Astrophysics Data System (ADS)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  3. Accident analysis of heavy water cooled thorium breeder reactor

    SciTech Connect

    Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  4. NEUTRONIC REACTOR

    DOEpatents

    Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

    1958-09-01

    This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

  5. Heterogeneous Recycling in Fast Reactors

    SciTech Connect

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  6. Coupled modeling of a directly heated tubular solar receiver for supercritical carbon dioxide Brayton cycle: Structural and creep-fatigue evaluation

    DOE PAGES

    Ortega, Jesus; Khivsara, Sagar; Christian, Joshua; Ho, Clifford; Dutta, Pradip

    2016-06-06

    A supercritical carbon dioxide (sCO2) Brayton cycle is an emerging high energy-density cycle undergoing extensive research due to the appealing thermo-physical properties of sCO2 and single phase operation. Development of a solar receiver capable of delivering sCO2 at 20 MPa and 700 °C is required for implementation of the high efficiency (~50%) solar powered sCO2 Brayton cycle. In this work, extensive candidate materials are review along with tube size optimization using the ASME Boiler and Pressure Vessel Code. Moreover, temperature and pressure distribution obtained from the thermal-fluid modeling (presented in a complementary publication) are used to evaluate the thermal andmore » mechanical stresses along with detailed creep-fatigue analysis of the tubes. For resulting body stresses were used to approximate the lifetime performance of the receiver tubes. A cyclic loading analysis is performed by coupling the Strain-Life approach and the Larson-Miller creep model. The structural integrity of the receiver was examined and it was found that the stresses can be withstood by specific tubes, determined by a parametric geometric analysis. The creep-fatigue analysis display the damage accumulation due to cycling and the permanent deformation on the tubes showed that the tubes can operate for the full lifetime of the receiver.« less

  7. Advanced Space Nuclear Reactors from Fiction to Reality

    NASA Astrophysics Data System (ADS)

    Popa-Simil, L.

    The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

  8. Performance improvement options for the supercritical carbon dioxide brayton cycle.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.; Nuclear Engineering Division

    2008-07-17

    The supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle is under development at Argonne National Laboratory as an advanced power conversion technology for Sodium-Cooled Fast Reactors (SFRs) as well as other Generation IV advanced reactors as an alternative to the traditional Rankine steam cycle. For SFRs, the S-CO{sub 2} Brayton cycle eliminates the need to consider sodium-water reactions in the licensing and safety evaluation, reduces the capital cost of the SFR plant, and increases the SFR plant efficiency. Even though the S-CO{sub 2} cycle has been under development for some time and optimal sets of operating parameters have been determined, those earlier development and optimization studies have largely been directed at applications to other systems such as gas-cooled reactors which have higher operating temperatures than SFRs. In addition, little analysis has been carried out to investigate cycle configurations deviating from the selected 'recompression' S-CO{sub 2} cycle configuration. In this work, several possible ways to improve S-CO{sub 2} cycle performance for SFR applications have been identified and analyzed. One set of options incorporates optimization approaches investigated previously, such as variations in the maximum and minimum cycle pressure and minimum cycle temperature, as well as a tradeoff between the component sizes and the cycle performance. In addition, the present investigation also covers options which have received little or no attention in the previous studies. Specific options include a 'multiple-recompression' cycle configuration, intercooling and reheating, as well as liquid-phase CO{sub 2} compression (pumping) either by CO{sub 2} condensation or by a direct transition from the supercritical to the liquid phase. Some of the options considered did not improve the cycle efficiency as could be anticipated beforehand. Those options include: a double recompression cycle, intercooling between the compressor stages, and reheating

  9. POWER REACTOR

    DOEpatents

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

  10. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.

    1957-10-01

    A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

  11. NUCLEAR REACTOR

    DOEpatents

    Christy, R.F.

    1958-07-15

    A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

  12. Bioconversion reactor

    DOEpatents

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  13. Catalytic reactor

    SciTech Connect

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  14. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  15. 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.

  16. 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.

  17. NUCLEAR REACTOR

    DOEpatents

    Young, G.

    1963-01-01

    This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

  18. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  19. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  20. Neutronic reactor construction

    DOEpatents

    Huston, Norman E.

    1976-07-06

    1. A neutronic reactor comprising a moderator including horizontal layers formed of horizontal rows of graphite blocks, alternate layers of blocks having the rows extending in one direction, the remaining alternate layers having the rows extending transversely to the said one direction, alternate rows of blocks in one set of alternate layers having longitudinal ducts, the moderator further including slotted graphite tubes positioned in the ducts, the reactor further comprising an aluminum coolant tube positioned within the slotted tube in spaced relation thereto, bodies of thermal-neutron-fissionable material, and jackets enclosing the bodies and being formed of a corrosion-resistant material having a low neutron-capture cross section, the bodies and jackets being positioned within the coolant tube so that the jackets are spaced from the coolant tube.

  1. Current Comparison of Advanced Nuclear Fuel Cycles

    SciTech Connect

    Steven Piet; Trond Bjornard; Brent Dixon; Robert Hill; Gretchen Matthern; David Shropshire

    2007-04-01

    This paper compares potential nuclear fuel cycle strategies – once-through, recycling in thermal reactors, sustained recycle with a mix of thermal and fast reactors, and sustained recycle with fast reactors. Initiation of recycle starts the draw-down of weapons-usable material and starts accruing improvements for geologic repositories and energy sustainability. It reduces the motivation to search for potential second geologic repository sites. Recycle in thermal-spectru

  2. Design Study of Small Pb-Bi Cooled Modified Candle Reactors

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Sekimoto, H.

    2010-06-01

    In this study application of modified CANDLE burnup scheme based long life Pb-Bi Cooled Fast Reactors for small long life reactors with natural Uranium as Fuel Cycle Input has been performed. The reactor cores are subdivided into several parts with the same volume in the axial directions. The natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2, and 10 years after that it is shifted to region 3. This concept is applied to all regions, i.e. shifted the core of I'th region into I+1 region after the end of 10 years burn-up cycle. The first region 1 is filled by fresh natural uranium fuel. Compared to the previous works, in a smaller reactor core the criticality need to be considered more carefully especially at the beginning of life. As an optimized design, a core of 85 cm radius and 150 cm height with 300 MWt power are selected. This core can be operated 10 years without refueling or fuel shuffling. The average discharge burn-up is 350 GWd/ton HM.

  3. Non-Nuclear Validation Test Results of a Closed Brayton Cycle Test-Loop

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.

    2007-01-01

    Both NASA and DOE have programs that are investigating advanced power conversion cycles for planetary surface power on the moon or Mars, or for next generation nuclear power plants on earth. Although open Brayton cycles are in use for many applications (combined cycle power plants, aircraft engines), only a few closed Brayton cycles have been tested. Experience with closed Brayton cycles coupled to nuclear reactors is even more limited and current projections of Brayton cycle performance are based on analytic models. This report describes and compares experimental results with model predictions from a series of non-nuclear tests using a small scale closed loop Brayton cycle available at Sandia National Laboratories. A substantial amount of testing has been performed, and the information is being used to help validate models. In this report we summarize the results from three kinds of tests. These tests include: 1) test results that are useful for validating the characteristic flow curves of the turbomachinery for various gases ranging from ideal gases (Ar or Ar/He) to non-ideal gases such as CO2, 2) test results that represent shut down transients and decay heat removal capability of Brayton loops after reactor shut down, and 3) tests that map a range of operating power versus shaft speed curve and turbine inlet temperature that are useful for predicting stable operating conditions during both normal and off-normal operating behavior. These tests reveal significant interactions between the reactor and balance of plant. Specifically these results predict limited speed up behavior of the turbomachinery caused by loss of load, the conditions for stable operation, and for direct cooled reactors, the tests reveal that the coast down behavior during loss of power events can extend for hours provided the ultimate heat sink remains available.

  4. The effect of thermal cycling to 1100 C on the alpha /Mo/ phase in directionally solidified gamma/gamma-prime-alpha alloys

    NASA Technical Reports Server (NTRS)

    Harf, F. H.

    1981-01-01

    Specimens of gamma/gamma-prime-alpha (Mo) eutectic alloy were thermally cycled or isothermally exposed at temperatures of 1075 to 1100 C. Transmission electron microscopy examination of cycled specimens indicated that even an exposure of 10 minutes effected noticeable changes in the shape of the alpha phase, and that the changes were cumulative as more cycles were added. The cross sections of fine, smooth fibers changed from rectangles to octagons, while lamellae and irregular shapes spheroidized. These effects are attributed to the differences in thermal expansion coefficients between the alpha phase and the gamma/gamma-prime matrix, and to the higher diffusion rates prevailing at elevated temperatures. Where the configuration of the alpha phase is a simple shape, such as a fiber, increasing the temperature eventually brings about a stress free interface between the alpha phase and the matrix by differential thermal expansion. Where the shape of the alpha phase is more complex, a stressed interface persists to higher temperatures where diffusion produces the more drastic morphological changes.

  5. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  6. Neutronic reactor

    DOEpatents

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  7. NEUTRONIC REACTOR

    DOEpatents

    Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

    1959-10-27

    BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

  8. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  9. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1962-12-25

    A reactor is described comprising a plurality of horizontal trays containing a solution of a fissionable material, the trays being sleeved on a vertical tube which contains a vertically-reciprocable control rod, a gas-tight chamber enclosing the trays, and means for conducting vaporized moderator from the chamber and for replacing vaporized moderator in the trays. (AEC)

  10. NEUTRONIC REACTORS

    DOEpatents

    Anderson, H.L.

    1958-10-01

    The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

  11. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation. PMID:27573503

  12. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1960-09-27

    A unit assembly is described for a neutronic reactor comprising a tube and plurality of spaced parallel sandwiches in the tube extending lengthwise thereof, each sandwich including a middle plate having a central opening for plutonium and other openings for fertile material at opposite ends of the plate.

  13. Solar Thermal Reactor Materials Characterization

    SciTech Connect

    Lichty, P. R.; Scott, A. M.; Perkins, C. M.; Bingham, C.; Weimer, A. W.

    2008-03-01

    Current research into hydrogen production through high temperature metal oxide water splitting cycles has created a need for robust high temperature materials. Such cycles are further enhanced by the use of concentrated solar energy as a power source. However, samples subjected to concentrated solar radiation exhibited lifetimes much shorter than expected. Characterization of the power and flux distributions representative of the High Flux Solar Furnace(HFSF) at the National Renewable Energy Laboratory(NREL) were compared to ray trace modeling of the facility. In addition, samples of candidate reactor materials were thermally cycled at the HFSF and tensile failure testing was performed to quantify material degradation. Thermal cycling tests have been completed on super alloy Haynes 214 samples and results indicate that maximum temperature plays a significant role in reduction of strength. The number of cycles was too small to establish long term failure trends for this material due to the high ductility of the material.

  14. Direct Observation of Active Material Concentration Gradients and Crystallinity Breakdown in LiFePO4 Electrodes During Charge/Discharge Cycling of Lithium Batteries

    PubMed Central

    2014-01-01

    The phase changes that occur during discharge of an electrode comprised of LiFePO4, carbon, and PTFE binder have been studied in lithium half cells by using X-ray diffraction measurements in reflection geometry. Differences in the state of charge between the front and the back of LiFePO4 electrodes have been visualized. By modifying the X-ray incident angle the depth of penetration of the X-ray beam into the electrode was altered, allowing for the examination of any concentration gradients that were present within the electrode. At high rates of discharge the electrode side facing the current collector underwent limited lithium insertion while the electrode as a whole underwent greater than 50% of discharge. This behavior is consistent with depletion at high rate of the lithium content of the electrolyte contained in the electrode pores. Increases in the diffraction peak widths indicated a breakdown of crystallinity within the active material during cycling even during the relatively short duration of these experiments, which can also be linked to cycling at high rate. PMID:24790684

  15. Nuclear reactor vessel fuel thermal insulating barrier

    SciTech Connect

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  16. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  17. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  18. Nuclear reactor

    DOEpatents

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  19. REACTOR UNLOADING

    DOEpatents

    Leverett, M.C.

    1958-02-18

    This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

  20. Thermionic reactors for space nuclear power

    NASA Technical Reports Server (NTRS)

    Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

    1985-01-01

    Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.