Science.gov

Sample records for reactors cycles directs

  1. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    DOEpatents

    Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

    1962-08-14

    A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  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. 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

  8. 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.

  9. 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)

  10. 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

  11. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1993-03-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions.

  12. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1991-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. 10 refs.

  13. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1993-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions.

  14. 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.

  15. 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

  16. 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.

  17. Bifurcations of limit cycles in open and closed loop reverse flow reactors

    NASA Astrophysics Data System (ADS)

    Russo, Lucia; Crescitelli, Silvestro; Brasiello, Antonio

    2013-10-01

    The present work analyses the bifurcations of limit cycles in open and loop reverse flow reactors. The open loop system consists of a reactor where the flow direction is periodically forced whereas in the closed loop system, the flow inversion is dictated by a control law which activates when the temperature at the edge of catalytic bed falls below the set-point value. We performed the bifurcation analysis of the open loop system as the switch time is varied and we constructed the solution diagram through the application of continuation technique. Many Naimark-Sacker bifurcations leading to quasi-periodic regimes have been found on the limit cycles branches. Finally, we compared these limit cycles with those of the closed loop system where the flow inversion is dictated by a control system which acts if the temperature measured at the edge of reactor falls below a set-point value.

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

    DOE PAGES

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

    2017-01-12

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

  19. 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.

  20. 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.

  1. 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.

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

    SciTech Connect

    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.

  3. 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

  4. 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

  5. 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.

  6. An Idealized Direct-Contact Biomass Pyrolysis Reactor Model

    NASA Technical Reports Server (NTRS)

    Miller, R. S.; Bellan, J.

    1996-01-01

    A numerical study is performed in order to assess the performance of biomass pyrolysis reactors which utilize direct particle-wall thermal conduction heating. An idealized reactor configuration consisting of a flat-plate turbulent boundary layer flow with particle convection along the heated wall and incorporating particle re-entrainment is considered.

  7. 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.

  8. 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.

  9. 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)

  10. 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

  11. Cycle and flow trusses in directed networks

    NASA Astrophysics Data System (ADS)

    Takaguchi, Taro; Yoshida, Yuichi

    2016-11-01

    When we represent real-world systems as networks, the directions of links often convey valuable information. Finding module structures that respect link directions is one of the most important tasks for analysing directed networks. Although many notions of a directed module have been proposed, no consensus has been reached. This lack of consensus results partly because there might exist distinct types of modules in a single directed network, whereas most previous studies focused on an independent criterion for modules. To address this issue, we propose a generic notion of the so-called truss structures in directed networks. Our definition of truss is able to extract two distinct types of trusses, named the cycle truss and the flow truss, from a unified framework. By applying the method for finding trusses to empirical networks obtained from a wide range of research fields, we find that most real networks contain both cycle and flow trusses. In addition, the abundance of (and the overlap between) the two types of trusses may be useful to characterize module structures in a wide variety of empirical networks. Our findings shed light on the importance of simultaneously considering different types of modules in directed networks.

  12. Cycle and flow trusses in directed networks

    PubMed Central

    Yoshida, Yuichi

    2016-01-01

    When we represent real-world systems as networks, the directions of links often convey valuable information. Finding module structures that respect link directions is one of the most important tasks for analysing directed networks. Although many notions of a directed module have been proposed, no consensus has been reached. This lack of consensus results partly because there might exist distinct types of modules in a single directed network, whereas most previous studies focused on an independent criterion for modules. To address this issue, we propose a generic notion of the so-called truss structures in directed networks. Our definition of truss is able to extract two distinct types of trusses, named the cycle truss and the flow truss, from a unified framework. By applying the method for finding trusses to empirical networks obtained from a wide range of research fields, we find that most real networks contain both cycle and flow trusses. In addition, the abundance of (and the overlap between) the two types of trusses may be useful to characterize module structures in a wide variety of empirical networks. Our findings shed light on the importance of simultaneously considering different types of modules in directed networks. PMID:28018610

  13. 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.

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

    NASA Astrophysics Data System (ADS)

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

    1992-10-01

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

  15. 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

  16. The basic features of a closed fuel cycle without fast reactors

    NASA Astrophysics Data System (ADS)

    Bobrov, E. A.; Alekseev, P. N.; Teplov, P. S.

    2017-01-01

    In this paper the basic features of a closed fuel cycle with thermal reactors are considered. The three variants of multiple Pu and U recycling in VVER reactors was investigated. The comparison of MOX and REMIX fuel approaches for closed fuel cycle with thermal reactors is presented. All variants make possible to recycle several times the total amount of Pu and U obtained from spent fuel. The reported study was funded by RFBR according to the research project № 16-38-00021

  17. 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.

  18. 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.

  19. 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.

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2017-01-01

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

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Economic prospects of the Integral Fast Reactor (IFR) fuel cycle

    SciTech Connect

    Chang, Y.I.; Till, C.E.

    1991-01-01

    The IFR fuel cycle based on pyroprocessing involves only few operational steps and the batch-oriented process equipment systems are compact. This results in major cost reductions in all of three areas of reprocessing, fabrication, and waste treatment. This document discusses the economic aspects of this fuel cycle.

  12. Development of alternate extractant systems for fast reactor fuel cycle

    SciTech Connect

    Vasudeva Rao, P.R.; Suresh, A.; Venkatesan, K.A.; Srinivasan, T.G.; Raj, Baldev

    2007-07-01

    Due to the limitations of TBP in processing of high burn-up, Pu-rich fast reactor fuels, there is a need to develop alternate extractants for fast reactor fuel processing. In this context, our Centre has been examining the suitability of alternate tri-alkyl phosphates. Third phase formation in the extraction of Th(IV) by TBP, tri-n-amyl phosphate (TAP) and tri-2-methyl-butyl phosphate (T2MBP) from nitric acid media has been investigated under various conditions to derive conclusions on their application for extraction of Pu at macro levels. The chemical and radiolytic degradation of tri-n-amyl-phosphate (TAP) diluted in normal paraffin hydrocarbon (NPH) in the presence of nitric acid has been investigated by the measurement of plutonium retention in organic phase. The potential application of room temperature ionic liquids (RTILs) for reprocessing of spent nuclear fuel has been explored. Extraction of uranium (VI) and palladium (II) from nitric acid medium by commercially available RTIL and tri-n-butyl phosphate solution in RTIL have been studied and the feasibility of electrodeposition of uranium as uranium oxide (UO{sub 2}) and palladium (II) as metallic palladium from the loaded organic phase have been demonstrated. This paper describes results of the above studies and discusses the suitability of the systems for fast reactor fuel reprocessing. (authors)

  13. 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)

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

    SciTech Connect

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

    2004-01-22

    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 the 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 candidate membrane performance under the gasification conditions, a high temperature/high pressure hydrogen permeation unit will be constructed in this project. During this reporting period, the design of this unit was completed. The unit will be capable of operating 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. By operating at higher temperatures and higher hydrogen partial pressures, we expect to demonstrate commercially relevant hydrogen flux, 10 {approx} 50 cc/min/cm{sup 2}, from the membranes made of the perovskite type of ceramic material. The construction of the unit is planned to be completed by the end of the next reporting period.

  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. 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

  17. 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

  18. 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

  19. 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

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

    SciTech Connect

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

    1984-01-01

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

  1. 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.

  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. 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,...

  4. 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.

  5. 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

  6. 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

  7. 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.

  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. 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.

  11. 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.

  12. 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.

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. 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.

  19. 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

  20. Energy efficiency analysis of reactor for torrefaction of biomass with direct heating

    NASA Astrophysics Data System (ADS)

    Kuzmina, J. S.; Director, L. B.; Shevchenko, A. L.; Zaichenko, V. M.

    2016-11-01

    Paper presents energy analysis of reactor for torrefaction with direct heating of granulated biomass by exhaust gases. Various schemes of gas flow through the reactor zones are presented. Performed is a comparative evaluation of the specific energy consumption for the considered schemes. It has been shown that one of the most expensive processes of torrefaction technology is recycling of pyrolysis gases.

  1. 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.

  2. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    NASA Astrophysics Data System (ADS)

    Stankovskiy, Evgeny Yuryevich

    In the recently completed RACE Project of the AFCI, accelerator-driven subcritical systems (ADS) experiments were conducted to develop technology of coupling accelerators to nuclear reactors. In these experiments electron accelerators induced photon-neutron reactions in heavy-metal targets to initiate fission reactions in ADS. Although the Idaho State University (ISU) RACE ADS was constructed only to develop measurement techniques for advanced experiments, many reactor kinetics experiments were conducted there. In the research reported in this dissertation, a method was developed to calculate kinetics parameters for measurement and calculation of the reactivity of ADS, a safety parameter that is necessary for control and monitoring of power production. Reactivity is measured in units of fraction of delayed versus prompt neutron from fission, a quantity that cannot be directly measured in far-subcritical reactors such as the ISU RACE configuration. A new technique is reported herein to calculate it accurately and to predict kinetic behavior of a far-subcritical ADS. Experiments conducted at ISU are first described and experimental data are presented before development of the kinetic theory used in the new computational method. Because of the complexity of the ISU ADS, the Monte-Carlo method as applied in the MCNP code is most suitable for modeling reactor kinetics. However, the standard method of calculating the delayed neutron fraction produces inaccurate values. A new method was developed and used herein to evaluate actual experiments. An advantage of this method is that its efficiency is independent of the fission yield of delayed neutrons, which makes it suitable for fuel with a minor actinide component (e.g. transmutation fuels). The implementation of this method is based on a correlated sampling technique which allows the accurate evaluation of delayed and prompt neutrons. The validity of the obtained results is indicated by good agreement between experimental

  3. Optimization strategies for sustainable fuel cycle of the BR2 Reactor

    SciTech Connect

    Kalcheva, S.; Van Den Branden, G.; Koonen, E.

    2013-07-01

    The objective of the present study is to achieve a sustainable fuel cycle in a long term of reactor operation applying advanced in-core loading strategies. The optimization criteria concern mainly enhancement of nuclear safety by means of reactivity margins and minimization of the operational fuel cycle cost at a given (constant) power level and same or longer cycle length. An important goal is also to maintain the same or to improve the experimental performances. Current developments are focused on optimization of control rods localization; optimization of fresh and burnt fuel assemblies in-core distribution; optimization of azimuth and axial fuel burn up strategies, including fuel assembly rotating and flipping upside down. (authors)

  4. 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)

  5. Direct coal-fired gas turbines for combined cycle plants

    SciTech Connect

    Rothrock, J.; Wenglarz, R.; Hart, P.; Mongia, H.

    1993-11-01

    The combustion/emissions control island of the CFTCC plant produces cleaned coal combustion gases for expansion in the gas turbine. The gases are cleaned to protect the turbine from flow-path degeneration due to coal contaminants and to reduce environmental emissions to comparable or lower levels than alternate clean coal power plant tedmologies. An advantage of the CFTCC system over other clean coal technologies using gas turbines results from the CFTCC system having been designed as an adaptation to coal of a natural gas-fired combined cycle plant. Gas turbines are built for compactness and simplicity. The RQL combustor is designed using gas turbine combustion technology rather than process plant reactor technology used in other pressurized coal systems. The result is simpler and more compact combustion equipment than for alternate technologies. The natural effect is lower cost and improved reliability. In addition to new power generation plants, CFTCC technology will provide relatively compact and gas turbine compatible coal combustion/emissions control islands that can adapt existing natural gas-fired combined cycle plants to coal when gas prices rise to the point where conversion is economically attractive. Because of the simplicity, compactness, and compatibility of the RQL combustion/emission control island compared to other coal technologies, it could be a primary candidate for such conversions.

  6. 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.

  7. 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

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

    DOE PAGES

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

    2017-03-01

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

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

    SciTech Connect

    Not Available

    1983-01-01

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

  10. 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.

  11. 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.

  12. Proliferation resistance of the fuel cycle for the Integral Fast Reactor

    SciTech Connect

    Burris, L.

    1993-09-01

    Argonne National Laboratory has developed an electrorefining pyrochemical process for recovery and recycle of metal fuel discharged from the Integral Fast Reactor (FR). This inherently low decontamination process has an overall decontamination factor of only about 100 for the plutonium metal product. As a result, all of the fuel cycle operations must be conducted in heavily shielded cells containing a high-purity argon atmosphere. The FR fuel cycle possesses high resistance to clandestine diversion or overt, state- supported removal of plutonium for nuclear weapons production because of two main factors: the highly radioactive product, which is also contaminated with heat- and neutron-producing isotopes of plutonium and other actinide elements, and the difficulty of removing material from the FR facility through the limited number of cell transfer locks without detection.

  13. 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.

  14. 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.

  15. 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.

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

    SciTech Connect

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

    2017-01-01

    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.

  17. Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

    SciTech Connect

    Chandler, David; Betzler, Ben; Hirtz, Gregory John; Ilas, Germina; Sunny, Eva

    2016-09-01

    The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length 252Cf production targets, 75Se production capsules, 63Ni production capsules, a 188W production capsule, and various materials irradiation targets. Fully loaded 238Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.

  18. 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)

  19. 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.

  20. 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)

  1. 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.

  2. 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.

  3. 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.

  4. 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)

  5. 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.

  6. SPERTI Reactor Pit Building (PER605) from contrasting direction as photo ...

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

    SPERT-I Reactor Pit Building (PER-605) from contrasting direction as photo above (ID-33-F-32). Note Guard House door, security fencing around facility. Photographer: R.G. Larsen. Date: July 22, 1955. INEEL negative no. 55-1702. - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  7. 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.

  8. 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)

  9. Management of transuranics using the Integral Fast Reactor (IFR) fuel cycle

    SciTech Connect

    Wade, D.C.

    1994-01-01

    The 50 years of activities following the discovery of self-sustaining fission chains have given rise to a buildup of roughly 900 tons of manmade transuranics. Of the total, about 260 tons of Pu{sup 239} were generated for use in weapons while the remainder were generated as a byproduct of electrical power produced worldwide by the commercial thermal nuclear power industry. What is to be done with these actinides? The options for disposition include interminable storage, burial, or recycle for use. The pros and cons of each option are being vigorously debated regarding the impact upon the issues of human and ecological risk -- both current and future; weapons proliferation potential -- both current and future; and total life cycle benefits and costs. As to the options for utilization, commercial uses for actinides (uranium and transuranics) are of limited diversity. The actinides have in the past and will in the future find application in large scale mostly by virtue of their ability to release energy through fission, and here their utility is unmatched -- whether the application be in commercial electricity generation or in armaments. The integral Fast Reactor (IFR) fuel cycle offers a number of features for management of the current and future burden of manmade transuranic materials and for capturing the energy content of the U{sup 238}. These features are discussed here.

  10. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2017-01-01

    The potential of elastic antineutrino-electron scattering 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 and the detector response was modeled using a Geant4-based simulation package. Background was estimated via independent simulations and by scaling published measurements from similar detectors. Background contributions were estimated for solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclides, water-borne radon, and gamma rays from the photomultiplier tubes (PMTs), detector walls, and surrounding rock. We show that with the use of low background PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. Directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. The results provide a list of experimental conditions that, if satisfied in practice, would enable antineutrino directional reconstruction at 3σ significance in large Gd-doped water Cherenkov detectors with greater than 10-km standoff from a nuclear reactor.

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

    SciTech Connect

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

    2016-10-17

    The potential of elastic antineutrino-electron scattering 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 and the detector response was modeled using a Geant4-based simulation package. Background was estimated via independent simulations and by scaling published measurements from similar detectors. Background contributions were estimated for solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclides, water-borne radon, and gamma rays from the photomultiplier tubes, detector walls, and surrounding rock. We show that with the use of low background PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. Directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. Lastly, the results provide a list of experimental conditions that, if satisfied in practice, would enable antineutrino directional reconstruction at 3 sigma significance in large Gd-doped water Cherenkov detectors with greater than 10 km standoff from a nuclear reactor.

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

    DOE PAGES

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

    2016-10-17

    The potential of elastic antineutrino-electron scattering 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 and the detector response was modeled using a Geant4-based simulation package. Background was estimated via independent simulations and by scaling published measurements from similar detectors. Background contributions were estimated for solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclides, water-borne radon, and gamma rays frommore » the photomultiplier tubes, detector walls, and surrounding rock. We show that with the use of low background PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. Directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. Lastly, the results provide a list of experimental conditions that, if satisfied in practice, would enable antineutrino directional reconstruction at 3 sigma significance in large Gd-doped water Cherenkov detectors with greater than 10 km standoff from a nuclear reactor.« less

  14. 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

  15. 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.

  16. 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.

  17. 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.

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

    SciTech Connect

    Scarangella, M. J.

    2012-07-01

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

  19. Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors

    SciTech Connect

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

    1983-11-01

    Several distinct operating modes (conventional ohmic, noninductive steady state, internal transformer, etc.) 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 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 (current drive efficiency) and engineering (superior materials) which will help achieve these goals for different burn cycles.

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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.

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

    NASA Technical Reports Server (NTRS)

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

    2003-01-01

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

  6. 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.

  7. 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.

  8. Nitrification-denitrification of UASB effluents highly loaded with nitrogen in an activated sludge reactor operated with short cycled aeration.

    PubMed

    Villaverde, S; Lacalle, M L; García-Encina, P A; Fdz-Polanco, F

    2001-01-01

    A conventional activated sludge reactor operated with short cycled aeration was used for total nitrogen removal of UASB anaerobic reactor effluent containing nitrogen (up to 1,200 mg NKT/L) and organic matter (up to 2,000 mg COD/L). Initially the reactor was fed with synthetic water to progressively introduce the UASB effluent. This favored the acclimation of the microorganisms to the real environment. The results obtained throughout this study showed that initially the tested technology is feasible and can report significant cuts on operation and maintenance when compared to conventional activated sludge processes. Total nitrogen removal up to 66% was attained treating the effluent of an UASB process designed for treating the wastewater of a potato starch factory. Total nitrogen removal capacities ranging between 0.1 and 0.58 kg of nitrogen per cubic metre per day are reported. Short-cycled aeration allowed for a more efficient use of the oxygen supply for nitrification and the organic carbon content present in the wastewater for denitrification. This operating protocol has demonstrated serious advantages in terms of operation costs and simplicity when total nitrogen removal is wanted. Most of the existing activated sludge processes, i.e. single continuous flow reactors, can be updated for total nitrogen removal essentially at no cost, the inversion (aeration control system) is rapidly returned as reduction in energy expenditure.

  9. 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).

  10. Comparison of the technical and economic parameters of different variants of the nuclear fuel cycle reactors of the nuclear power plants

    NASA Astrophysics Data System (ADS)

    Adamov, E. O.; Rachkov, V. I.; Tolstoukhov, D. A.; Panov, S. A.

    2016-12-01

    The article presents the comparative economic analysis of the variants of implementation of the fuel cycles for fast and thermal reactors. Calculations are carried out for the formed reference conditions by the cost of processing stages of the fuel cycles for the foreign and expert Russian information sources. The comparative data on the resource supply, absolute and specific costs of the fuel cycle variants for the whole life cycle of the thermal power plant projects with the fast and thermal reactors are considered. The conclusions of the efficiency of the fuel cycle variants for the assumed reference conditions of the calculation are made.

  11. Development of a thermal scheme for a cogeneration combined-cycle unit with an SVBR-100 reactor

    NASA Astrophysics Data System (ADS)

    Kasilov, V. F.; Dudolin, A. A.; Krasheninnikov, S. M.

    2017-02-01

    At present, the prospects for development of district heating that can increase the effectiveness of nuclear power stations (NPS), cut down their payback period, and improve protection of the environment against harmful emissions are being examined in the nuclear power industry of Russia. It is noted that the efficiency of nuclear cogeneration power stations (NCPS) is drastically affected by the expenses for heat networks and heat losses during transportation of a heat carrier through them, since NPSs are usually located far away from urban area boundaries as required for radiation safety of the population. The prospects for using cogeneration power units with small or medium power reactors at NPSs, including combined-cycle units and their performance indices, are described. The developed thermal scheme of a cogeneration combined-cycle unit (CCU) with an SBVR-100 nuclear reactor (NCCU) is presented. This NCCU should use a GE 6FA gasturbine unit (GTU) and a steam-turbine unit (STU) with a two-stage district heating plant. Saturated steam from the nuclear reactor is superheated in a heat-recovery steam generator (HRSG) to 560-580°C so that a separator-superheater can be excluded from the thermal cycle of the turbine unit. In addition, supplemental fuel firing in HRSG is examined. NCCU effectiveness indices are given as a function of the ambient air temperature. Results of calculations of the thermal cycle performance under condensing operating conditions indicate that the gross electric efficiency η el NCCU gr of = 48% and N el NCCU gr = 345 MW can be achieved. This efficiency is at maximum for NCCU with an SVBR-100 reactor. The conclusion is made that the cost of NCCU installed kW should be estimated, and the issue associated with NCCUs siting with reference to urban area boundaries must be solved.

  12. Microcapillary reactors using solid-phase DNA sequencing for direct sample introduction into slab gels.

    PubMed

    Xu, Y; Bruch, R C; Soper, S A

    2000-05-01

    Solid-phase micro-reactors have been prepared in glass capillaries for DNA sequencing applications using slab gel electrophoresis, which consisted of a fused silica capillary (i.d. = 100 microns; o.d. = 365 microns; length = 15 cm; volume = 1.2 microL) that contained a covalently bound biotin molecule. With the addition of streptavidin to the capillary, an anchoring site was produced for the tethering of biotinylated DNA sequencing templates to the wall of the capillary. Using a four-lane, single dye primer chemistry sequencing strategy, the individual tracts were prepared in the capillaries using cycle sequencing (20 thermal cycles) on a PCR-generated lambda-bacteriophage template (about 1000 bp). The dye label in this case was a fluorescent tag that displayed emission properties in the near-IR and could be processed on an automated sequencer. The read length was found to be 589 bases, which was determined primarily by the fractionating power of the gel. It was also found that the tethering system was very stable to typical cycle sequencing conditions, with the amount of tethered DNA lost amounting to 40% after 120 thermal cycles. The ability to use dye terminator chemistry was also investigated by using a near-IR dye-labeled terminator (ddGTP). It was found that the quality of the ladder that was generated was comparable to that obtained in a conventional sample preparation format. However, ethanol precipitation was required before gel loading to remove excess terminator.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. 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

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

    PubMed

    Vitor-Costa, Marcelo; Okuno, Nilo Massaru; Bortolotti, Henrique; Bertollo, Maurizio; Boggio, Paulo Sergio; Fregni, Felipe; Altimari, Leandro Ricardo

    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

  1. DIRECT ENERGY CONVERSION (DEC) FISSION REACTORS - A U.S. NERI PROJECT

    SciTech Connect

    D. BELLER; G. POLANSKY; ET AL

    2000-11-01

    The direct conversion of the electrical energy of charged fission fragments was examined early in the nuclear reactor era, and the first theoretical treatment appeared in the literature in 1957. Most of the experiments conducted during the next ten years to investigate fission fragment direct energy conversion (DEC) were for understanding the nature and control of the charged particles. These experiments verified fundamental physics and identified a number of specific problem areas, but also demonstrated a number of technical challenges that limited DEC performance. Because DEC was insufficient for practical applications, by the late 1960s most R&D ceased in the US. Sporadic interest in the concept appears in the literature until this day, but there have been no recent programs to develop the technology. This has changed with the Nuclear Energy Research Initiative that was funded by the U.S. Congress in 1999. Most of the previous concepts were based on a fission electric cell known as a triode, where a central cathode is coated with a thin layer of nuclear fuel. A fission fragment that leaves the cathode with high kinetic energy and a large positive charge is decelerated as it approaches the anode by a charge differential of several million volts, it then deposits its charge in the anode after its kinetic energy is exhausted. Large numbers of low energy electrons leave the cathode with each fission fragment; they are suppressed by negatively biased on grid wires or by magnetic fields. Other concepts include magnetic collimators and quasi-direct magnetohydrodynamic generation (steady flow or pulsed). We present the basic principles of DEC fission reactors, review the previous research, discuss problem areas in detail and identify technological developments of the last 30 years relevant to overcoming these obstacles. A prognosis for future development of direct energy conversion fission reactors will be presented.

  2. 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)

  3. 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)

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Solar reforming of methane in a direct absorption catalytic reactor on a parabolic dish

    SciTech Connect

    Muir, J.F.; Hogan, R.E. Jr.; Skocypec, R.D. ); Buck, R. , Stuttgart . Inst. fuer Technische Thermodynamik)

    1990-01-01

    The concept of solar driven chemical reactions in a commercial-scale volumetric receiver/reactor on a parabolic concentrator was successfully demonstrated in the CAtalytically Enhanced Solar Absorption Receiver (CAESAR) test. Solar reforming of methane (CH{sub 4}) with carbon dioxide (CO{sub 2}) was achieved in a 64-cm diameter direct absorption reactor on a parabolic dish capable of 150 kW solar power. The reactor was a catalytic volumetric absorber consisting of a multi-layered, porous alumina foam disk coated with rhodium (Rh) catalyst. The system was operated during both steady-state and solar transient (cloud passage) conditions. The total solar power absorbed reached values up to 97 kW and the maximum methane conversion was 70%. Receiver thermal efficiencies ranged up to 85% and chemical efficiencies peaked at 54%. The absorber performed satisfactorily in promoting the reforming reaction during the tests without carbon formation. However, problems of cracking and degradation of the porous matrix, nonuniform dispersion of the Rh through the absorber, and catalyst deactivation due to sintering and possible encapsulation, must be resolved to achieve long-term operation and eventual commercialization. 17 refs., 11 figs., 1 tab.

  10. 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.

  11. Turbostar: an ICF reactor using both direct and thermal power conversion. Revision 1

    SciTech Connect

    Pitts, J.H.

    1986-07-31

    Combining direct and thermal power conversion results in a 52% gross plant efficiency with DT fuel and 68% with advanced DD fuel. We maximize the fraction of fusion-yield energy converted to kinetic energy in a liquid-lithium blanket, and use this energy directly with turbine generators to produce electricity. We use the remainder of the energy to produce electricity in a standard Rankine thermal power conversion cycle.

  12. Plans for the development of the IFR (Integral Fast Reactor) fuel cycle

    SciTech Connect

    Johnson, T.R.

    1986-01-01

    The Integral Fast Reactor (IFR) is a concept for a self-contained facility in which several sodium-cooled fast reactors of moderate size are located at the same site along with complete fuel-recycle and waste-treatment facilities. After the initial core loading with enriched uranium or plutonium, only natural or depleted uranium is shipped to the plant, and only wastes in final disposal forms are shipped out. The reactors have driver and blanket fuels of uranium-plutonium-zirconium alloys in stainless steel cladding. The use of metal alloy fuels is central to the IFR concept, contributing to the inherent safety of the reactor, the ease of reprocessing, and the relatively low capital and operating costs. Discharged fuels are recovered in a pyrochemical process that consists of two basic steps: an electrolytic process to separate fission products from actinides, and halide slagging to separate plutonium from uranium.

  13. 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

  14. Direct energy conversion in fission reactors: A U.S. NERI project

    SciTech Connect

    SLUTZ,STEPHEN A.; SEIDEL,DAVID B.; POLANSKY,GARY F.; ROCHAU,GARY E.; LIPINSKI,RONALD J.; BESENBRUCH,G.; BROWN,L.C.; PARISH,T.A.; ANGHAIE,S.; BELLER,D.E.

    2000-05-30

    In principle, the energy released by a fission can be converted directly into electricity by using the charged fission fragments. The first theoretical treatment of direct energy conversion (DEC) appeared in the literature in 1957. Experiments were conducted over the next ten years, which identified a number of problem areas. Research declined by the late 1960's due to technical challenges that limited performance. Under the Nuclear Energy Research Initiative the authors are determining if these technical challenges can be overcome with todays technology. The authors present the basic principles of DEC reactors, review previous research, discuss problem areas in detail, and identify technological developments of the last 30 years that can overcome these obstacles. As an example, the fission electric cell must be insulated to avoid electrons crossing the cell. This insulation could be provided by a magnetic field as attempted in the early experiments. However, from work on magnetically insulated ion diodes they know how to significantly improve the field geometry. Finally, a prognosis for future development of DEC reactors will be presented .

  15. 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

  16. 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

  17. 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).

  18. Integral fast reactor concept. [Pool type; metal fuel; integral fuel cycle

    SciTech Connect

    Chang, Y.I.; Marchaterre, J.F.; Sevy, R.H.

    1984-01-01

    Key features of the IFR consist of a pool-type plant arrangement, a metal fuel-based core design, and an integral fuel cycle with colocated fuel cycle facility. Both the basic concept and the technology base have been demonstrated through actual integral cycle operation in EBR-II. This paper discusses the inherent safety characteristics of the IFR concept. (DLC)

  19. 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).

  20. Utilization of the catalyzed-DD fuel cycle in reversed-field pinch reactors (RFPRs)

    SciTech Connect

    Hagenson, R.L.; Krakowski, R.A.

    1981-01-01

    The utilization of deuterium-based fuels offers the potential advantages of greater flexibility in blanket design, significantly reduced tritium inventory, potential reduction in radioactivity level, and utilization of an inexhaustible fuel supply. The extension of the conventional DT-fueled Reversed-Field Pinch Reactor (RFPR) designs to advanced-fuel (catalyzed-DD) operation has recently been reported. Attractive and economically competitive DD/RFPR systems are identified having power densities and plasma parameters comparable to the DT systems. These designs are compared to other fusion reactor designs.

  1. Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Sarkar, Apu; Kumawat, Bhupendra K.; Chakravartty, J. K.

    2015-07-01

    The cyclic stress-strain response and the low cycle fatigue (LCF) behavior of 20MnMoNi55 pressure vessel steel were studied. Tensile strength and LCF properties were examined at room temperature (RT) using specimens cut from rolling direction of a rolled block. The fully reversed strain-controlled LCF tests were conducted at a constant total strain rate with different axial strain amplitude levels. The cyclic strain-stress relationships and the strain-life relationships were obtained through the test results, and related LCF parameters of the steel were calculated. The studied steel exhibits cyclic softening behavior. Furthermore, analysis of stabilized hysteresis loops showed that the steel exhibits non-Masing behavior. Complementary scanning electron microscopy examinations were also carried out on fracture surfaces to reveal dominant damage mechanisms during crack initiation, propagation and fracture. Multiple crack initiation sites were observed on the fracture surface. The investigated LCF behavior can provide reference for pressure vessel life assessment and fracture mechanisms analysis.

  2. 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.

  3. In-reactor testing of the closed cycle gas core reactor—the nuclear light bulb concept

    NASA Astrophysics Data System (ADS)

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

    1993-01-01

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

  4. 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

  5. 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.

  6. A COMPARISON OF PEBBLE MIXING AND DEPLETION ALGORITHMS USED IN PEBBLE-BED REACTOR EQUILIBRIUM CYCLE SIMULATION

    SciTech Connect

    Hans D. Gougar; Frederik Reitsma; Wessel Joubert

    2009-05-01

    Recirculating pebble-bed reactors are distinguished from all other reactor types by the downward movement through and reinsertion of fuel into the core during operation. Core simulators must account for this movement and mixing in order to capture the physics of the equilibrium cycle core. VSOP and PEBBED are two codes used to perform such simulations, but they do so using different methods. In this study, a simplified pebble-bed core with a specified flux profile and cross sections is used as the model for conducting analyses of two types of burnup schemes. The differences between the codes are described and related to the differences observed in the nuclide densities in pebbles discharged from the core. Differences in the methods for computing fission product buildup and average number densities lead to significant differences in the computed core power and eigenvalue. These test models provide a key component of an overall equilibrium cycle benchmark involving neutron transport, cross section generation, and fuel circulation.

  7. 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.

  8. 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.

  9. Implications of the DD fuel cycle on tokamak reactor technology considerations

    SciTech Connect

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Evans, K. Jr.; Jung, J.; Mattas, R.; Misra, B.; Smith, D.L.; Stevens, H.C.; Turner, L.R.

    1980-01-01

    The initial results of a study to develop a conceptual design of a DD tokamak reactor are reported in this paper. A comprehensive steady-state power and particle balance code has been used to survey a variety of design points. A reference set of parameters (R/sub 0/ = 8.6 m, B/sub max/ = 14 T, ..beta.. = 11%, P/sub th/ = 2500 to 2700 MW) has been selected for further study. The concepts of rf driven currents and a limiter-vacuum system have been examined and appear to offer promise. However, the penalty of rf current drive, in terms of recirculating power, is relatively larger for DD than for DT reactors. Blanket engineering studies have emphasized the optimization of energy multiplication and inboard blanket thickness.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    SciTech Connect

    Brower, Jeffrey O.; Glazoff, Michael V.; Eiden, Thomas J.; Rezvoi, Aleksey V.

    2016-08-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed scalloping

  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. 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.

  19. 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

  20. 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.

  1. 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.-

  2. Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

    DOE PAGES

    Brown, Nicholas R.; Worrall, Andrew; Todosow, Michael

    2016-11-18

    Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example 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. The hypothetical light water SMR example case considered in these preliminary scoping studies ismore » a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs 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 burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. 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 burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium

  3. Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

    SciTech Connect

    Brown, Nicholas R.; Worrall, Andrew; Todosow, Michael

    2016-11-18

    Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example 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. The hypothetical light water SMR example case considered in these preliminary scoping studies is a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs 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 burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. 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 burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium disposed

  4. 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.

  5. 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.

  6. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for FHRs

    NASA Astrophysics Data System (ADS)

    Lu, Qiuping

    Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines coated particle fuel and a 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. These loops are coupled through two heat exchangers, namely, the DRACS Heat Exchanger (DHX) and the Natural Draft Heat Exchanger (NDHX). In addition, a fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during normal operation of the reactor, but to keep the DRACS ready for activation, if needed, during accidents. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. The primary goal of the present research is to design, test, and model the DRACS for FHR applications. Previously, a detailed modular design of the DRACS for a 20-MWth FHR was developed. As a starting point, the DRACS was designed to remove 1% of the reactor nominal power, i.e., 200 kW decay power. In addition, a detailed scaling analysis has been performed to develop the key non-dimensional numbers that characterize the DRACS system. Based on the previous work on the prototypic DRACS design and scaling analysis, two scaled-down test facilities have been designed and constructed, namely, Low-temperature DRACS Test Facility (LTDF) and High-temperature DRACS Test Facility (HTDF). The LTDF has a nominal power capacity of 6 kW. It uses 1.0-MPa water as the primary coolant, 0.1-MPa water as the secondary coolant, and ambient air as the ultimate heat sink. The main purpose of the LTDF is to examine the couplings among the three natural circulation/convection loops in the DRACS, as well as to provide design and operation experience for the HTDF. An extensive test matrix has

  7. 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.

  8. Detailed heat load calculations at the beginning, middle, and end of cycle for the conceptual design of the Advanced Neutron Source Reactor

    SciTech Connect

    Wemple, C. A.; Schnitzler, B. G.

    1995-04-01

    The Advanced Neutron Source (ANS) is a world-class research reactor and experimental center for neutron research, presently being designed at the Oak Ridge National Laboratory (ORNL). The reactor consists of a 330-MW(f) highly enriched uranium core, which is cooled, moderated, and reflected with heavy water. When completed, it will be the preeminent ultrahigh neutron flux reactor in the world, with facilities for research programs in biology, materials science, chemistry, fundamental and nuclear physics, and analytical chemistry. Irradiation facilities are provided for a variety of isotope production capabilities, as well as materials irradiation. The ANS reactor design, at the time of this report, has completed the conceptual design phase and entered the advanced conceptual design phase. This report is part of an effort to fully document the analysis methods and results for the conceptual design. It details the methods used to perform heat load calculations on the ANS reactor design, describes the model used, and gives the resulting heat loads in all components of the reactor, in both a differential (by segment) and integral (by component) fashion. These heat load data are provided at three times within the ANS fuel cycle - at beginning (0 days), middle (8.5 days), and end (17 days) of cycle. The remainder of the report is dedicated to this description. In Chapter 2, some necessary background on the reactor design is provided. Chapters 3 and 4 give details of the depletion methods used and revisions to previous MCNP models. Chapter 5 analyzes the results of these calculations, and Chapter 6 provides a summary and conclusions.

  9. 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

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

    NASA Technical Reports Server (NTRS)

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

    2007-01-01

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

  11. 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)

  12. 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

  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. 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.

  15. 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)

  16. 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.

  17. 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.

  18. 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).

  19. 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.

  20. Progress and interim results of the INPRO joint study on assessment of INS based on closed nuclear fuel cycle with fast reactors

    SciTech Connect

    Usanov, Vladimir; Raj, Baldev; Vasile, Alfredo

    2007-07-01

    The purpose of the work is to review interim results of the Joint Study on assessment of an Innovative Nuclear System based on a Closed Nuclear Fuel Cycle with Fast Reactors (INS CNFC-FR). This study is a part of the IAEA international project for innovative reactors and fuel cycle technologies (INPRO). Now it is being implemented by Canada, China, France, India, Japan, Republic of Korea, Russia, and Ukraine. A report on results of implementation of the first phase of the Joint Study was presented to the INPRO Steering Committee meeting in December 2006. It was also agreed by the Joint Study participants to reveal these results to broader discussion at scientific conferences and meetings. The authors' interpretation of the Joint Study findings and issues is presented in the paper. (authors)

  1. 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.

  2. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    SciTech Connect

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  3. 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.

  4. 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

  5. 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.

  6. Once-through CANDU reactor models for the ORIGEN2 computer code

    SciTech Connect

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given.

  7. Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK{sup TM}

    SciTech Connect

    Wright, Steven A.; Sanchez, Travis

    2005-02-06

    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{sup TM} (Simulink, 2004). SIMULINK{sup TM} is a development environment packaged with MatLab{sup TM} (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{sup TM} 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{sup TM} 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)

  8. 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).

  9. Effect of condensate cycles in driving atmospheric circulation on brown dwarfs and directly imaged giant planets

    NASA Astrophysics Data System (ADS)

    Tan, Xianyu; Showman, Adam

    2014-11-01

    Growing observations of brown dwarfs and directly imaged giant planets, including properties of the L/T transition, chemical disequilibrium, brightness variability, and surface maps have provided evidence for strong atmospheric circulation on these worlds. Previous studies that serve to understand the atmospheric circulation of brown dwarfs include modeling of convection from the interior both in a two-dimensional and global fashion, a two-layer shallow water model and a global circulation model with dry thermal perturbation at the bottom of atmosphere. These models show that interactions between the stably stratified layer and the convective interior can drive an atmospheric circulation, including zonal jets and/or vortices. However, these models are dry models, not including the condensation cycles such as silicate and iron in hot dwarfs. Condensation of water has previously been shown to play an important role on driving the zonal jets on four giant planets in our solar system. As such, condensation cycles of various species is believed to be an important source in driving the atmospheric circulation of brown dwarfs and directly imaged planets as well. Here we present results from three-dimensional simulations for the stably stratified atmospheres of brown dwarfs based on a general circulation model that includes the effect of a condensate cycle. Large-scale latent heating and molecular weight effect due to condensation of a single species are treated explicitly in our model. We examine the atmospheric circulation patterns of brown dwarfs caused by large-scale latent heating that results from condensation of silicates in hot dwarfs and water in the cold dwarfs. By varying the parameters such as abundances of condensates, effective temperature and rotational period, we explore possible configurations of the circulation, and determine implications for the observed cloud patchiness and brightness variability for brown dwarfs.

  10. Oligodendrocyte Progenitor Cells Directly Utilize Lactate for Promoting Cell Cycling and Differentiation

    PubMed Central

    Ichihara, Yoshinori; Doi, Toru; Ryu, Youngjae; Nagao, Motoshi; Sawada, Yasuhiro

    2016-01-01

    Oligodendrocyte progenitor cells (OPCs) undergo marked morphological changes to become mature oligodendrocytes, but the metabolic resources for this process have not been fully elucidated. Although lactate, a metabolic derivative of glycogen, has been reported to be consumed in oligodendrocytes as a metabolite, and to ameliorate hypomyelination induced by low glucose conditions, it is not clear about the direct contribution of lactate to cell cycling and differentiation of OPCs, and the source of lactate for remyelination. Therefore, we evaluated the effect of 1,4‐dideoxy‐1,4‐imino‐d‐arabinitol (DAB), an inhibitor of the glycogen catabolic enzyme glycogen phosphorylase, in a mouse cuprizone model. Cuprizone induced demyelination in the corpus callosum and remyelination occurred after cuprizone treatment ceased. This remyelination was inhibited by the administration of DAB. To further examine whether lactate affects proliferation or differentiation of OPCs, we cultured mouse primary OPC‐rich cells and analyzed the effect of lactate. Lactate rescued the slowed cell cycling induced by 0.4 mM glucose, as assessed by the BrdU‐positive cell ratio. Lactate also promoted OPC differentiation detected by monitoring the mature oligodendrocyte marker myelin basic protein, in the presence of both 36.6 mM and 0.4 mM glucose. Furthermore, these lactate‐mediated effects were suppressed by the reported monocarboxylate transporter inhibitor, α‐cyano‐4‐hydroxy‐cinnamate. These results suggest that lactate directly promotes the cell cycling rate and differentiation of OPCs, and that glycogen, one of the sources of lactate, contributes to remyelination in vivo. J. Cell. Physiol. 232: 986–995, 2017. © 2016 The Authors. Journal of Cellular Physiology Published by Wiley Periodicals, Inc. PMID:27861886

  11. Oligodendrocyte Progenitor Cells Directly Utilize Lactate for Promoting Cell Cycling and Differentiation.

    PubMed

    Ichihara, Yoshinori; Doi, Toru; Ryu, Youngjae; Nagao, Motoshi; Sawada, Yasuhiro; Ogata, Toru

    2017-05-01

    Oligodendrocyte progenitor cells (OPCs) undergo marked morphological changes to become mature oligodendrocytes, but the metabolic resources for this process have not been fully elucidated. Although lactate, a metabolic derivative of glycogen, has been reported to be consumed in oligodendrocytes as a metabolite, and to ameliorate hypomyelination induced by low glucose conditions, it is not clear about the direct contribution of lactate to cell cycling and differentiation of OPCs, and the source of lactate for remyelination. Therefore, we evaluated the effect of 1,4-dideoxy-1,4-imino-d-arabinitol (DAB), an inhibitor of the glycogen catabolic enzyme glycogen phosphorylase, in a mouse cuprizone model. Cuprizone induced demyelination in the corpus callosum and remyelination occurred after cuprizone treatment ceased. This remyelination was inhibited by the administration of DAB. To further examine whether lactate affects proliferation or differentiation of OPCs, we cultured mouse primary OPC-rich cells and analyzed the effect of lactate. Lactate rescued the slowed cell cycling induced by 0.4 mM glucose, as assessed by the BrdU-positive cell ratio. Lactate also promoted OPC differentiation detected by monitoring the mature oligodendrocyte marker myelin basic protein, in the presence of both 36.6 mM and 0.4 mM glucose. Furthermore, these lactate-mediated effects were suppressed by the reported monocarboxylate transporter inhibitor, α-cyano-4-hydroxy-cinnamate. These results suggest that lactate directly promotes the cell cycling rate and differentiation of OPCs, and that glycogen, one of the sources of lactate, contributes to remyelination in vivo. J. Cell. Physiol. 232: 986-995, 2017. © 2016 The Authors. Journal of Cellular Physiology Published by Wiley Periodicals, Inc.

  12. 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)

  13. An introduction to thermodynamics of renewable cycles for direct solar energy conversion

    SciTech Connect

    Sukhodolsky, A.T.

    1998-07-01

    Mechanical energy is known to be converted into thermal form (heat) without any restriction. Any opposite conversion of heat into mechanical energy owing to work produced in the heat engines has been restricted by principle of Carnot. This communication is to introduce into the elements of thermodynamics for direct energy conversion of light into mechanical energy. The aim is to explain: why without any machines, the efficiency of conversion is able to be by many orders of magnitude greater than the efficiency of direct mechanical action of light given in framework of equilibrium radiation thermodynamics. The development of such a concept is to find out an actual fundamental restriction for the maximum conversion available in several new solar renewable technologies for both direct generation of mechanical vibrations and for extraction of pure water from mixtures owing to self-organization of heat cycles. In order to involve solar light under non-equilibrium with respect to matter as a motive power within thermodynamics, the principle of Carnot for heat engines is assumed to be also valid for renewable processes (cycles). The absorbing part of condensed matter by optical pumping is considered as a heat source for production of entropy by heat-transfer into dark surrounding that plays role of heat sink. Principle of Carnot is used together with common accepted definition of non-equilibrium entropy in order to describe the excitement of heat source and its next relaxation. The new formulation of Carnot theorem and fundamental maximum of renewable conversion is derived. The simplest system of two equal bodies with difference temperatures is considered to show how to find the maximum energy available for renewable conversion. The major difference between equilibrium (reversible) thermodynamics and proposed renewable (non-equilibrium) approach is discussed on this example together with a typical mathematical paradox.

  14. Gimme shelter--the relative sensitivity of parasitic nematodes with direct and indirect life cycles to climate change.

    PubMed

    Molnár, Péter K; Dobson, Andrew P; Kutz, Susan J

    2013-11-01

    Climate change is expected to alter the dynamics of host-parasite systems globally. One key element in developing predictive models for these impacts is the life cycle of the parasite. It is, for example, commonly assumed that parasites with an indirect life cycle would be more sensitive to changing environmental conditions than parasites with a direct life cycle due to the greater chance that at least one of their obligate host species will go extinct. Here, we challenge this notion by contrasting parasitic nematodes with a direct life cycle against those with an indirect life cycle. Specifically, we suggest that behavioral thermoregulation by the intermediate host may buffer the larvae of indirectly transmitted parasites against temperature extremes, and hence climate warming. We term this the 'shelter effect'. Formalizing each life cycle in a comprehensive model reveals a fitness advantage for the direct life cycle over the indirect life cycle at low temperatures, but the shelter effect reverses this advantage at high temperatures. When examined for seasonal environments, the models suggest that climate warming may in some regions create a temporal niche in mid-summer that excludes parasites with a direct life cycle, but allows parasites with an indirect life cycle to persist. These patterns are amplified if parasite larvae are able to manipulate their intermediate host to increase ingestion probability by definite hosts. Furthermore, our results suggest that exploiting the benefits of host sheltering may have aided the evolution of indirect life cycles. Our modeling framework utilizes the Metabolic Theory of Ecology to synthesize the complexities of host behavioral thermoregulation and its impacts on various temperature-dependent parasite life history components in a single measure of fitness, R0 . It allows quantitative predictions of climate change impacts, and is easily generalized to many host-parasite systems.

  15. 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.

  16. 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.

  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. 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

  19. 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.

  20. 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.

  1. Coral spawn timing is a direct response to solar light cycles and is not an entrained circadian response

    NASA Astrophysics Data System (ADS)

    Brady, A. K.; Hilton, J. D.; Vize, P. D.

    2009-09-01

    Broadcast spawning corals release gametes into the oceans with extraordinarily accurate timing. While the date of spawning is set by the lunar cycle, the hour/minute of spawning is set by the solar cycle. In this report, we describe experiments that test whether the time of spawning is regulated by an entrained biological clock or whether it is directly controlled by the solar cycle. Montastraea franksi samples were collected on the morning of the predicted spawning. Fragments from colonies were kept under three different lighting conditions and spawning monitored. The three conditions were sunset times of 0, 1 or 2 h earlier than normal. Fragments from the same colony spawned differently under these three conditions, with an early sunset causing a corresponding early shift in spawning. These results indicate that spawn timing is not controlled by a circadian rhythm and that it is directly controlled by local solar light cycle.

  2. Comparison between direct and predicted maximal oxygen uptake measurement during cycling.

    PubMed

    Santtila, Matti; Häkkinen, Keijo; Pihlainen, Kai; Kyröläinen, Heikki

    2013-02-01

    Predicted maximal oxygen uptake (VO2max) measurements are based on the assumption of linear relationship between heart rate or power output and oxygen consumption during various intensities. To develop more reliable predicted test for soldiers, the purpose of the present study was to compare the results of direct measurements of VO2max to respective predicted values in cycling (military fitness test). The predicted mean (+/- SD) peak oxygen uptake (VO2peak) value was 45.2 +/- 7.7 mL kg(-1) min(-1) during first week, whereas the respective direct value was 44.8 +/- 8.5 mL kg(-1) min(-1). During the ninth week, the predicted and measured mean (+/-SD) VO2max values were 47.4 +/- 6.7 mL kg(-1) min(-1) and 48.7 +/- 7.3 mL kg(-1) min(-1), respectively. The absolute differences between the methods were -0.42 mL kg(-1) min(-1) (p = 0.46) and 1.28 mL kg(-1) min(-1) (p < 0.05), which correspond to relative values of 0.9% and 2.7%, respectively. A Bland-Altman plot of measured VO2max and predicted VO2max showed no significant trend between the mean and the difference of the 2 methods either before (r = 0.14, p = 0.24) or after the basic military training period (r = 0.11, p = 0.36). Intraclass correlation coefficient varied between r = 0.82 to 0.94. In conclusion, the predicted protocol is fairly accurate (+/-3%) and reliable to predict VO2max values in male soldiers but the use for clinical purposes should be considered individually.

  3. 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.

  4. 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.

  5. Response of the first wetted wall of an IFE reactor chamber to the energy release from a direct-drive DT capsule

    NASA Astrophysics Data System (ADS)

    Medin, Stanislav A.; Basko, Mikhail M.; Orlov, Yurii N.; Suslin, Victor M.

    2012-07-01

    Radiation hydrodynamics 1D simulations were performed with two concurrent codes, DEIRA and RAMPHY. The DEIRA code was used for DT capsule implosion and burn, and the RAMPHY code was used for computation of X-ray and fast ions deposition in the first wall liquid film of the reactor chamber. The simulations were run for 740 MJ direct drive DT capsule and Pb thin liquid wall reactor chamber of 10 m diameter. Temporal profiles for DT capsule leaking power of X-rays, neutrons and fast 4He ions were obtained and spatial profiles of the liquid film flow parameter were computed and analyzed.

  6. 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.

  7. 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.

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

    SciTech Connect

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

    1984-01-01

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

  9. 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

  10. 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.

  11. The SoxRS response of Escherichia coli is directly activated by redox-cycling drugs rather than by superoxide

    PubMed Central

    Gu, Mianzhi; Imlay, James A.

    2011-01-01

    When Escherichia coli is exposed to redox-cycling drugs, its SoxR transcription factor is activated by oxidation of its [2Fe–2S] cluster. In aerobic cells these drugs generate superoxide, and because superoxide dismutase (SOD) is a member of the SoxRS regulon, superoxide was initially thought to be the activator of SoxR. Its many-gene regulon was therefore believed to comprise a defense against superoxide stress. However, we found that abundant superoxide did not effectively activate SoxR in an SOD− mutant, that overproduced SOD could not suppress activation by redox-cycling drugs, and that redox-cycling drugs were able to activate SoxR in anaerobic cells as long as alternative respiratory acceptors were provided. Thus superoxide is not the signal that SoxR senses. Indeed, redox-cycling drugs directly oxidized the cluster of purified SoxR in vitro, while superoxide did not. Redox-cycling drugs are excreted by both bacteria and plants. Their toxicity does not require superoxide, as they poisoned E. coli under anaerobic conditions, in part by oxidizing dehydratase iron-sulfur clusters. Under these conditions SoxRS induction was protective. Thus it is physiologically appropriate that the SoxR protein directly senses redox-cycling drugs rather than superoxide. PMID:21226770

  12. 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.

  13. Denitrification and neutralization treatment by direct feeding of an acidic wastewater containing copper ion and high-strength nitrate to a bio-electrochemical reactor process.

    PubMed

    Watanabe, T; Motoyama, H; Kuroda, M

    2001-12-01

    The feasibility of the direct denitrification treatment of copper metal pickling wastewater by using a bio-electrochemical reactor process was investigated experimentally. Carbon electrodes were installed in the reactor as the anode and cathode and denitrifying microorganisms were fixed on the surface of the cathode. The reactor was continuously operated by applying an electric current and feeding acetate. In this reactor, copper ion removal and denitrification proceeded simultaneously and the pH value of the treated water was increased almost to neutral. The electric current that passed through the cathode contributed to the removal of the copper ion and the generation of hydrogen gas. The generated hydrogen gas as well as the added acetate was effectively utilized for denitrification. A theoretical evaluation of pH in the effluent suggested that the pH increase was mainly caused by the generation of hydroxyl ion during denitrification. In addition, the inorganic carbon species generated during denitrification with acetate and by the electrochemical oxidation of anodic carbon acted as a buffer to minimize a further increase of pH at higher nitrate removal efficiencies. These results demonstrated that copper ion removal, denitrification and neutralization could be achieved simultaneously by using a single bioelectrochemical reactor.

  14. 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.

  15. 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.

  16. 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.

  17. 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)

  18. 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)

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

    NASA Astrophysics Data System (ADS)

    Yehia, Ashraf; Mizuno, Akira

    2013-05-01

    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.

  20. 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.

  1. 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.

  2. 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)

  3. 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)

  4. 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

  5. Effect of thermal cycling on the microstructure of a directionally solidified Fe, Cr, Al-TaC eutectic alloy

    NASA Technical Reports Server (NTRS)

    Harf, F. H.; Tewari, S. N.

    1977-01-01

    Cylindrical bars (1.2 cm diameter) of Fe-13.6Cr-3.7Al-9TaC (wt %) eutectic alloy were directionally solidified in a modified Bridgman type furnace at 1 cm/h. The alloy microstructure consisted of aligned TaC fibers imbedded in a bcc Fe-Cr-Al matrix. Specimens of the alloy were thermally cycled from 1100 to 425 C in a burner rig. The effects of 1800 thermal cycles on the microstructure was examined by scanning electron microscopy, revealing a zig-zag shape of TaC fibers aligned parallel to the growth direction. The mechanism of carbide solution and reprecipitation on the (111) easy growth planes, suggested previously to account for the development of irregular serrations in Co-Cr-Ni matrix alloys, is believed to be responsible for these zig-zag surfaces.

  6. Characterization of the existence of an N0-completion of a partial N0-matrix with an associated directed cycle.

    PubMed

    Jordán, Cristina; Torregrosa, Juan R

    2014-01-01

    An n × n matrix is called an N 0-matrix if all its specified principal minors are nonpositive. In the context of partial matrices, a partial matrix is called a partial N 0-matrix if all its specified principal minors are nonpositive. In this paper we characterize the existence of an N 0-matrix completion of a partial N 0-matrix whose associated graph is a directed cycle.

  7. Characterization of the Existence of an N0-Completion of a Partial N0-Matrix with an Associated Directed Cycle

    PubMed Central

    Jordán, Cristina

    2014-01-01

    An n × n matrix is called an N0-matrix if all its specified principal minors are nonpositive. In the context of partial matrices, a partial matrix is called a partial N0-matrix if all its specified principal minors are nonpositive. In this paper we characterize the existence of an N0-matrix completion of a partial N0-matrix whose associated graph is a directed cycle. PMID:24688437

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Effect of water on sulfur dioxide (SO{sub 2}) and nitrogen oxides (NO{sub x}) removal from flue gas in a direct current corona discharge reactor

    SciTech Connect

    Yang, Jiaxiang; Chi, Xiaochun; Dong, Limin

    2007-05-15

    A direct current (dc) corona discharge reactor composed of needle-plate electrodes in a glass container filled with flue gas was designed. To clarify the influence of water on discharge characteristics, water was introduced in the plasma reactor as electrode where plate electrode is immersed, under the application of dc voltage. Experiment results show that (1) corona wind forming between high-voltage needle electrode and water by corona discharge enhances the cleaning efficiency of flue gas due to the existence of water and the cleaning efficiency will increase with the increase of applied dc voltage within definite range and (2) both removal efficiencies of NO{sub x} and SO{sub 2} increased in the presence of water, which reach up to 98% for SO{sub 2}, and about 85% for NO{sub x} under suitable conditions. These results play an important role in flue gas cleanup research.

  13. 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.

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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/.

  4. 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.

  5. 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.

  6. Direct elicitation of template concentration from quantification cycle (Cq) distributions in digital PCR

    PubMed Central

    Mojtahedi, Mitra; Fouquier d'Hérouël, Aymeric; Huang, Sui

    2014-01-01

    Digital PCR (dPCR) exploits limiting dilution of a template into an array of PCR reactions. From this array the number of reactions that contain at least one (as opposed to zero) initial template is determined, allowing inferring the original template concentration. Here we present a novel protocol to efficiently infer the concentration of a sample and its optimal dilution for dPCR from few targeted qPCR assays. By taking advantage of the real-time amplification feature of qPCR as opposed to relying on endpoint PCR assessment as in standard dPCR prior knowledge of template concentration is not necessary. This eliminates the need for serial dilutions in a separate titration and reduces the number of necessary reactions. We describe the theory underlying our approach and discuss experimental moments that contribute to uncertainty. We present data from a controlled experiment where the initial template concentration is known as proof of principle and apply our method on directly monitoring transcript level change during cell differentiation as well as gauging amplicon numbers in cDNA samples after pre-amplification. PMID:25104023

  7. 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.

  8. Fast Reactors

    NASA Astrophysics Data System (ADS)

    Esposito, S.; Pisanti, O.

    The following sections are included: * Elementary Considerations * The Integral Equation to the Neutron Distribution * The Critical Size for a Fast Reactor * Supercritical Reactors * Problems and Exercises

  9. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-09-01

    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.

  11. Site-Directed Mutagenesis of the Nidovirus Replicative Endoribonuclease NendoU Exerts Pleiotropic Effects on the Arterivirus Life Cycle

    PubMed Central

    Posthuma, Clara C.; Nedialkova, Danny D.; Zevenhoven-Dobbe, Jessika C.; Blokhuis, Jeroen H.; Gorbalenya, Alexander E.; Snijder, Eric J.

    2006-01-01

    The highly conserved NendoU replicative domain of nidoviruses (arteriviruses, coronaviruses, and roniviruses) belongs to a small protein family whose cellular branch is prototyped by XendoU, a Xenopus laevis endoribonuclease involved in nucleolar RNA processing. Recently, sequence-specific in vitro endoribonuclease activity was demonstrated for the NendoU-containing nonstructural protein (nsp) 15 of several coronaviruses. To investigate the biological role of this novel enzymatic activity, we have characterized a comprehensive set of arterivirus NendoU mutants. Deleting parts of the NendoU domain from nsp11 of equine arteritis virus was lethal. Site-directed mutagenesis of conserved residues exerted pleiotropic effects. In a first-cycle analysis, replacement of two conserved Asp residues in the C-terminal part of NendoU rendered viral RNA synthesis and virus production undetectable. In contrast, mutagenesis of other conserved residues, including two putative catalytic His residues that are absolutely conserved in NendoU and cellular homologs, produced viable mutants displaying reduced plaque sizes (20 to 80% reduction) and reduced yields of infectious progeny of up to 5 log units. A more detailed analysis of these mutants revealed a moderate reduction in RNA synthesis, with subgenomic RNA synthesis consistently being more strongly affected than genome replication. Our data suggest that the arterivirus nsp11 is a multifunctional protein with a key role in viral RNA synthesis and additional functions in the viral life cycle that are as yet poorly defined. PMID:16439522

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

    PubMed

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

    2015-09-23

    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.

  13. Ovol2 suppresses cell cycling and terminal differentiation of keratinocytes by directly repressing c-Myc and Notch1.

    PubMed

    Wells, Julie; Lee, Briana; Cai, Anna Qianyao; Karapetyan, Adrine; Lee, Wan-Ju; Rugg, Elizabeth; Sinha, Satrajit; Nie, Qing; Dai, Xing

    2009-10-16

    Ovol2 belongs to the Ovo family of evolutionarily conserved zinc finger transcription factors that act downstream of key developmental signaling pathways including Wg/Wnt and BMP/TGF-beta. We previously reported Ovol2 expression in the basal layer of epidermis, where epidermal stem/progenitor cells reside. In this work, we use HaCaT human keratinocytes to investigate the cellular and molecular functions of Ovol2. We show that depletion of Ovol2 leads to transient cell expansion but a loss of cells with long term proliferation potential. Mathematical modeling and experimental findings suggest that both faster cycling and precocious withdrawal from the cell cycle underlie this phenotype. Ovol2 depletion also accelerates extracellular signal-induced terminal differentiation in two- and three-dimensional culture models. By chromatin immunoprecipitation, luciferase reporter, and functional rescue assays, we demonstrate that Ovol2 directly represses two critical downstream targets, c-Myc and Notch1, thereby suppressing keratinocyte transient proliferation and terminal differentiation, respectively. These findings shed light on how an epidermal cell maintains a proliferation-competent and differentiation-resistant state.

  14. The First Ground Level Enhancement Event of Solar Cycle 24: Direct Observation of Shock Formation and Particle Release Heights

    NASA Technical Reports Server (NTRS)

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

    2013-01-01

    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.

  15. 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.

  16. 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

  17. 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.

  18. 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.

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

    PubMed

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

    2016-02-01

    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.

  20. Moving bed reactor for solar thermochemical fuel production

    DOEpatents

    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.

  1. 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.

  2. 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.

  3. 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.

  4. 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)

  5. Evidence for a 17-year cycle in the IMF directions at 1 AU, in coronal hole variations, and in planetary magnetospheric modulations

    NASA Astrophysics Data System (ADS)

    Juckett, David A.

    1998-11-01

    A dominant 16-17yr cycle was observed in the net exposure times of the Earth to Toward and Away field directions of the interplanetary magnetic field (IMF). A cycle of the same frequency and phase was observed in the polarity of the long-term hemispheric differences in coronal hole distributions. This was determined from north/south differences in average Fexiv green line `quiet' regions at high- and mid-latitudes. It is argued that the 17-yr cycle is a fundamental oscillation of coronal hole topology, which is transferred to Earth via variations in the neutral sheet. A comparison of the 17-yr cycle to the 22-yr Hale cycle indicated that they are not identical, but rather, can mix to form a 75-yr cycle plus a 9-yr cycle. Evidence for the 75-yr cycle existed in the Earth's net exposure times to fields from the solar North and South, and in the long-term imbalance of solar quiet regions between the northern and southern hemispheres. The 9-yr cycle was manifested in the mid- to low- latitude Fexiv modulations and in solar wind velocity variations in the ecliptic. At Earth, evidence for a similar 17-yr cycle was observed in the horizontal magnetic field observations in a multitude of surface magnetic recording stations. In addition, the detection of a 17-yr cycle in the Huancayo neutron monitor cosmic ray series suggests that the effects of this cycle extend to the heliospheric boundaries. It is concluded that sufficient preliminary evidence exists to consider the hypothesis that the Sun contains a magnetic moment with an oscillatory cycle of 17years.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Chronic fluoxetine treatment directs energy metabolism towards the citric acid cycle and oxidative phosphorylation in rat hippocampal nonsynaptic mitochondria.

    PubMed

    Filipović, Dragana; Costina, Victor; Perić, Ivana; Stanisavljević, Andrijana; Findeisen, Peter

    2017-03-15

    Fluoxetine (Flx) is the principal treatment for depression; however, the precise mechanisms of its actions remain elusive. Our aim was to identify protein expression changes within rat hippocampus regulated by chronic Flx treatment versus vehicle-controls using proteomics. Fluoxetine-hydrohloride (15mg/kg) was administered daily to adult male Wistar rats for 3weeks, and cytosolic and nonsynaptic mitochondrial hippocampal proteomes were analyzed. All differentially expressed proteins were functionally annotated according to biological process and molecular function using Uniprot and Blast2GO. Our comparative study revealed that in cytosolic and nonsynaptic mitochondrial fractions, 60 and 3 proteins respectively, were down-regulated, and 23 and 60 proteins, respectively, were up-regulated. Proteins differentially regulated in cytosolic and nonsynaptic mitochondrial fractions were primarily related to cellular and metabolic processes. Of the identified proteins, the expressions of calretinin and parvalbumine were confirmed. The predominant molecular functions of differentially expressed proteins in both cell hippocampal fractions were binding and catalytic activity. Most differentially expressed proteins in nonsynaptic mitochondria were catalytic enzymes involved in the pyruvate metabolism, citric acid cycle, oxidative phosphorylation, ATP synthesis, ATP transduction and glutamate metabolism. Results indicate that chronic Flx treatment may influence proteins involved in calcium signaling, cytoskeletal structure, chaperone system and stimulates energy metabolism via the upregulation of GAPDH expression in cytoplasm, as well as directing energy metabolism toward the citric acid cycle and oxidative phosphorylation in nonsynaptic mitochondria. This approach provides new insight into the chronic effects of Flx treatment on protein expression in a key brain region associated with stress response and memory.

  11. 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)

  12. 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

  13. Advances by the Integral Fast Reactor Program

    SciTech Connect

    Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

    1991-01-01

    The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs.

  14. 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

  15. Integral Fast Reactor concept

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  16. 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.

  17. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  18. 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.

  19. 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.

  20. 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)

  1. 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.

  2. Nuclear fuel cycle information workshop

    SciTech Connect

    Not Available

    1983-01-01

    This overview of the nuclear fuel cycle is divided into three parts. First, is a brief discussion of the basic principles of how nuclear reactors work; second, is a look at the major types of nuclear reactors being used and world-wide nuclear capacity; and third, is an overview of the nuclear fuel cycle and the present industrial capability in the US.

  3. 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.

  4. 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.

  5. Sludge reduction by direct addition of chlorine dioxide into a sequencing batch reactor under operational mode of repeatedly alternating aeration/non-aeration.

    PubMed

    Peng, Hong; Liu, Weiyi; Li, Yuanmei; Xiao, Hong

    2015-01-01

    The effect of direct addition of chlorine dioxide (ClO2) into a repeatedly alternating aeration/non-aeration sequencing batch reactor (SBR) on its sludge reduction and process performance was investigated. The experimental results showed that the sludge reduction efficiency was 32.9% and the observed growth yield (Yobs) of SBR was 0.11 kg VSS (volatile suspended solids) /kg COD (chemical oxygen demand) for 80 days' operation at the optimum ClO2 dosage of 2.0 mg/g TSS (total suspended solids). It was speculated that cell lysis and cryptic growth, uncoupled metabolism and endogenous metabolism were jointly responsible for the sludge reduction in this study. COD, NH3-N, total nitrogen (TN) and total phosphorus (TP) in the effluent increased on average 29.47, 4.44, 1.97 and 0.05 mg/L, respectively. However, the effluent quality still satisfied the first-class B discharge standards for municipal wastewater treatment plants in China. In that case, the sludge maintained fine viability with the specific oxygen uptake rate (SOUR) being 14.47 mg O2/(g VSS·h) and demonstrated good settleability with the sludge volume index (SVI) being 116 mL/g. The extra cost of sludge reduction at the optimum ClO2 dosage was estimated to be 2.24 CNY (or 0.36 dollar)/kg dry sludge.

  6. Development of alloy 718 tubular product for nuclear reactor internals

    SciTech Connect

    1981-01-01

    The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides mixing and flow direction of the core outlet flow. Alloy 718 tubes are the major components used in the UIS to provide this flow direction. The UIS is located directly above the reactor core and is exposed to a severe environment. This environment consists of high temperature sodium, alternating temperatures induced by mixing high temperature core assembly outlet flow with cooler core assembly outlet flow and rapid changes in temperature of the core outlet flow. The paper presents the UIS configuration, functions and environmental conditions that led to the selection of Alloy 718 as the material used to protect the basic UIS structure and to provide flow direction. The paper describes the tube fabrication process, the development of a finish sanding procedure and the results of high temperature thermal cycle testing.

  7. 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.

  8. 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

  9. 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.

  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. 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

  12. 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.

  13. Experimental demonstration of half cycle 64-QAM Nyquist-SCM direct-detection optical communication system with data-aided estimation and overlap frequency-domain equalization

    NASA Astrophysics Data System (ADS)

    Li, Danyu; He, Jing; Tang, Jin; Chen, Ming; Chen, Lin

    2014-12-01

    A half cycle 64-quadrature amplitude modulation (QAM) Nyquist subcarrier modulation (SCM) polarization division multiplexing (PDM) intensity modulation direct detection optical communication system is experimentally demonstrated. At the receiver, training sequences-based channel estimation and an overlap frequency domain equalization method are proposed to enhance the system performance. The experimental results show that the half cycle 64-QAM Nyquist-SCM PDM signal can be transmitted over 43-km standard single-mode fibers with a bit error rate below the forward error coding threshold of 2.4×10-2.

  14. 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.

  15. 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)

  16. 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.

  17. 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)

  18. First Cycle of PISA (2000-2006)--International Perspectives on Successes and Challenges: Research and Policy Directions

    ERIC Educational Resources Information Center

    Anderson, John O.; Chiu, Mei-Hung; Yore, Larry D.

    2010-01-01

    This special edition of IJMSE focuses on the Programme of International Student Assessment (PISA) project now that it has completed a full cycle of administration--reading, mathematics, and science--to look at ways in which PISA has been used in participating countries and with what consequences, and to identify potential research and policy…

  19. 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.

  20. 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.

  1. 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.

  2. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

    SciTech Connect

    Mac Donald, Philip Elsworth

    2002-06-01

    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. Metallic fuels for advanced reactors

    NASA Astrophysics Data System (ADS)

    Carmack, W. J.; Porter, D. L.; Chang, Y. I.; Hayes, S. L.; Meyer, M. K.; Burkes, D. E.; Lee, C. B.; Mizuno, T.; Delage, F.; Somers, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.

  4. 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.

  5. Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

    SciTech Connect

    Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N.

    2012-07-01

    The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existing facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)

  6. 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

  7. 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

  8. Safeguards Considerations for Thorium Fuel Cycles

    DOE PAGES

    Worrall, Louise G.; Worrall, Andrew; Flanagan, George F.; ...

    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

  9. Safeguards Considerations for Thorium Fuel Cycles

    SciTech Connect

    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 protocols 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.

  10. 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.

  11. Magnetic susceptibility as a direct measure of oxidation state in LiFePO4 batteries and cyclic water gas shift reactors.

    PubMed

    Kadyk, Thomas; Eikerling, Michael

    2015-08-14

    The possibility of correlating the magnetic susceptibility to the oxidation state of the porous active mass in a chemical or electrochemical reactor was analyzed. The magnetic permeability was calculated using a hierarchical model of the reactor. This model was applied to two practical examples: LiFePO4 batteries, in which the oxidation state corresponds with the state-of-charge, and cyclic water gas shift reactors, in which the oxidation state corresponds to the depletion of the catalyst. In LiFePO4 batteries phase separation of the lithiated and delithiated phases in the LiFePO4 particles in the positive electrode gives rise to a hysteresis effect, i.e. the magnetic permeability depends on the history of the electrode. During fast charge or discharge, non-uniform lithium distributionin the electrode decreases the hysteresis effect. However, the overall sensitivity of the magnetic response to the state-of-charge lies in the range of 0.03%, which makes practical measurement challenging. In cyclic water gas shift reactors, the sensitivity is 4 orders of magnitude higher and without phase separation, no hysteresis occurs. This shows that the method is suitable for such reactors, in which large changes of the magnetic permeability of the active material occurs.

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

    SciTech Connect

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

    2004-01-01

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

  13. 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

  14. 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.

  15. 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.

  16. 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.

  17. 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)

  18. Retrospective Dosimetry of Vver 440 Reactor Pressure Vessel at the 3RD Unit of Dukovany Npp

    NASA Astrophysics Data System (ADS)

    Marek, M.; Viererbl, L.; Sus, F.; Klupak, V.; Rataj, J.; Hogel, J.

    2009-08-01

    Reactor pressure vessel (RPV) residual lifetime of the Czech VVER-440 is currently monitored under Surveillance Specimens Programs (SSP) focused on reactor pressure vessel materials. Neutron fluence in the samples and its distribution in the RPV are determined by a combination of calculation results and the experimental data coming from the reactor dosimetry measurements both in the specimen containers and in the reactor cavity. The direct experimental assessment of the neutron flux density incident onto RPV and neutron fluence for the entire period of nuclear power plant unit operation can be based on the evaluation of the samples taken from the inner RPV cladding. The Retrospective Dosimetry was also used at Dukovany NPP at its 3rd unit after the 18th cycle. The paper describes methodology, experimental setup for sample extraction, measurement of activities, and the determination of the neutron flux and fluence averaged over the samples.

  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.

    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.

  20. Investigating the Dendritic Growth during Full Cell Cycling of Garnet Electrolyte in Direct Contact with Li Metal.

    PubMed

    Aguesse, Frederic; Manalastas, William; Buannic, Lucienne; Lopez Del Amo, Juan Miguel; Singh, Gurpreet; Llordés, Anna; Kilner, John

    2017-02-01

    All-solid-state batteries including a garnet ceramic as electrolyte are potential candidates to replace the currently used Li-ion technology, as they offer safer operation and higher energy storage performances. However, the development of ceramic electrolyte batteries faces several challenges at the electrode/electrolyte interfaces, which need to withstand high current densities to enable competing C-rates. In this work, we investigate the limits of the anode/electrolyte interface in a full cell that includes a Li-metal anode, LiFePO4 cathode, and garnet ceramic electrolyte. The addition of a liquid interfacial layer between the cathode and the ceramic electrolyte is found to be a prerequisite to achieve low interfacial resistance and to enable full use of the active material contained in the porous electrode. Reproducible and constant discharge capacities are extracted from the cathode active material during the first 20 cycles, revealing high efficiency of the garnet as electrolyte and the interfaces, but prolonged cycling leads to abrupt cell failure. By using a combination of structural and chemical characterization techniques, such as SEM and solid-state NMR, as well as electrochemical and impedance spectroscopy, it is demonstrated that a sudden impedance drop occurs in the cell due to the formation of metallic Li and its propagation within the ceramic electrolyte. This degradation process is originated at the interface between the Li-metal anode and the ceramic electrolyte layer and leads to electromechanical failure and cell short-circuit. Improvement of the performances is observed when cycling the full cell at 55 °C, as the Li-metal softening favors the interfacial contact. Various degradation mechanisms are proposed to explain this behavior.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.

    1960-11-22

    A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

  6. 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.

  7. 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)

  8. 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.

  9. 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.

  10. 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)

  11. 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.

  12. 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).

  13. 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.

  14. 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

  15. 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

  16. Interplanetary Coronal Mass Ejections Resulting from Earth-Directed CMEs Using SOHO and ACE Combined Data During Solar Cycle 23

    NASA Astrophysics Data System (ADS)

    Paouris, Evangelos; Mavromichalaki, Helen

    2017-02-01

    In this work a total of 266 interplanetary coronal mass ejections observed by the Solar and Heliospheric Observatory/ Large Angle and Spectrometric Coronagraph (SOHO/LASCO) and then studied by in situ observations from Advanced Composition Explorer (ACE) spacecraft, are presented in a new catalog for the time interval 1996 - 2009 covering Solar Cycle 23. Specifically, we determine the characteristics of the CME which is responsible for the upcoming ICME and the associated solar flare, the initial/background solar wind plasma and magnetic field conditions before the arrival of the CME, the conditions in the sheath of the ICME, the main part of the ICME, the geomagnetic conditions of the ICME's impact at Earth and finally we remark on the visual examination for each event. Interesting results revealed from this study include the high correlation coefficient values of the magnetic field Bz component against the Ap index (r = 0.84), as well as against the Dst index (r = 0.80) and of the effective acceleration against the CME linear speed (r = 0.98). We also identify a north-south asymmetry for X-class solar flares and an east-west asymmetry for CMEs associated with strong solar flares (magnitude ≥ M1.0) which finally triggered intense geomagnetic storms (with Ap ≥179). The majority of the geomagnetic storms are determined to be due to the ICME main part and not to the extreme conditions which dominate inside the sheath. For the intense geomagnetic storms the maximum value of the Ap index is observed almost 4 hours before the minimum Dst index. The amount of information makes this new catalog the most comprehensive ICME catalog for Solar Cycle 23.

  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. 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

    2015-10-06

    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

  19. Nuclear data requirements for fusion reactor nucleonics

    SciTech Connect

    Bhat, M.R.; Abdou, M.A.

    1980-01-01

    Nuclear data requirements for fusion reactor nucleonics are reviewed and the present status of data are assessed. The discussion is divided into broad categories dealing with data for Fusion Materials Irradiation Test Facility (FMIT), D-T Fusion Reactors, Alternate Fuel Cycles and the Evaluated Data Files that are available or would be available in the near future.

  20. 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.

  1. 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.

  2. 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.

  3. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    SciTech Connect

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  4. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    SciTech Connect

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  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. Demonstration of 48-Gb/s 16-QAM signal transmission using half cycle sub-carrier modulation in intensity modulation/direct detection system

    NASA Astrophysics Data System (ADS)

    Tang, Jin; He, Jing; Chen, Ming; Li, Danyu; Chen, Lin

    2015-01-01

    A simple spectral-efficiency intensity modulation/direct detection (IM/DD) system based on half cycle sub-carrier modulation (SCM) signal is proposed for short reach fiber communications in this paper. The signal impairment of frequency selective fading due to fiber chromatics dispersion (CD) is mathematically analyzed. To reduce the performance deterioration caused by the non-flat transfer function, digital pre- and post-equalization is applied in the system. The peak to average power ratio (PAPR) of the signal is also discussed in comparison with that of orthogonal frequency division multiplexing (OFDM). The transmission of 16-QAM half cycle SCM signal with a sub-carrier frequency of half the symbol rate and Nyquist pulse shaping is experimentally demonstrated. The bit-error rate (BER) of 48 Gb/s polarization multiplexing division (PDM) 16 QAM half cycle SCM signal is less than 7% forward-error-correction (FEC) threshold of 3.8 ×10-3 after transmission over 83 km standard single-mode fiber (SSMF).

  12. Direct Coulomb explosion imaging of coherent rotational dynamics induced by few-cycle laser pulses in light and heavy hydrogen

    NASA Astrophysics Data System (ADS)

    Bocharova, Irina; Magrakvelidze, Maia; Ranitovic, Predrag; Ray, Dipanwita; Cocke, Lewis; Litvinyuk, Igor

    2008-05-01

    We followed fast evolution of angular distributions for H2 and D2 molecules after their interaction with 8 fs 800 nm laser pulses. The rotating molecules were exploded by another few-cycle probe pulse time-delayed for up to 10 ps in respect to the pump. For neutral molecules we observed coherent rotational dynamics characterized by periodic revivals without noticeable decoherence within the 10 ps time-scale. For D2 up to 4 rotational states were involved in the wavepackets for each of the two spin isomers. In light hydrogen the resulting dynamics was dominated by beating of just two rotational states. The experimental data are in excellent agreement with our numerical simulations obtained by solving time-dependent Schr"odinger equation. For molecules that were ionized by the pump pulse we observed both vibrational and rotational dynamics. Time-dependent angular distributions for the molecular ions exhibit transient alignment only soon after the pulse (20 fs for H2^+ and 40 fs for D2^+) with no consequent revivals within the next 10 ps.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. 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.

  1. Calculation of reactor life in delayed coking units

    SciTech Connect

    Kuzeev, I.R.; Filimonov, E.A.; Gribanov, A.V.; Polyakov, I.V.

    1988-01-01

    The operational reliability of delayed coking reactors may vary considerably from one unit to another. This paper presents a method for calculating the reliability and service life of such reactors based on cycle frequency, materials defects, stresses, fatigue, and temperature and service loads. The method was developed for service life prediction in the design stage of delayed coking reactors.

  2. 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.

  3. A Life Cycle Assessment Case Study of Coal-Fired Electricity Generation with Humidity Swing Direct Air Capture of CO2 versus MEA-Based Postcombustion Capture.

    PubMed

    van der Giesen, Coen; Meinrenken, Christoph J; Kleijn, René; Sprecher, Benjamin; Lackner, Klaus S; Kramer, Gert Jan

    2017-01-17

    Most carbon capture and storage (CCS) envisions capturing CO2 from flue gas. Direct air capture (DAC) of CO2 has hitherto been deemed unviable because of the higher energy associated with capture at low atmospheric concentrations. We present a Life Cycle Assessment of coal-fired electricity generation that compares monoethanolamine (MEA)-based postcombustion capture (PCC) of CO2 with distributed, humidity-swing-based direct air capture (HS-DAC). Given suitable temperature, humidity, wind, and water availability, HS-DAC can be largely passive. Comparing energy requirements of HS-DAC and MEA-PCC, we find that the parasitic load of HS-DAC is less than twice that of MEA-PCC (60-72 kJ/mol versus 33-46 kJ/mol, respectively). We also compare other environmental impacts as a function of net greenhouse gas (GHG) mitigation: To achieve the same 73% mitigation as MEA-PCC, HS-DAC would increase nine other environmental impacts by on average 38%, whereas MEA-PCC would increase them by 31%. Powering distributed HS-DAC with photovoltaics (instead of coal) while including recapture of all background GHG, reduces this increase to 18%, hypothetically enabling coal-based electricity with net-zero life-cycle GHG. We conclude that, in suitable geographies, HS-DAC can complement MEA-PCC to enable CO2 capture independent of time and location of emissions and recapture background GHG from fossil-based electricity beyond flue stack emissions.

  4. Schizophrenia susceptibility genes directly implicated in the life cycles of pathogens: cytomegalovirus, influenza, herpes simplex, rubella, and Toxoplasma gondii.

    PubMed

    Carter, C J

    2009-11-01

    Many genes implicated in schizophrenia can be related to glutamatergic transmission and neuroplasticity, oligodendrocyte function, and other families clearly related to neurobiology and schizophrenia phenotypes. Others appear rather to be involved in the life cycles of the pathogens implicated in the disease. For example, aspartylglucosaminidase (AGA), PLA2, SIAT8B, GALNT7, or B3GAT1 metabolize chemical ligands to which the influenza virus, herpes simplex, cytomegalovirus (CMV), rubella, or Toxoplasma gondii bind. The epidermal growth factor receptor (EGR/EGFR) is used by the CMV to gain entry to cells, and a CMV gene codes for an interleukin (IL-10) mimic that binds the host cognate receptor, IL10R. The fibroblast growth factor receptor (FGFR1) is used by herpes simplex. KPNA3 and RANBP5 control the nuclear import of the influenza virus. Disrupted in schizophrenia 1 (DISC1) controls the microtubule network that is used by viruses as a route to the nucleus, while DTNBP1, MUTED, and BLOC1S3 regulate endosomal to lysosomal routing that is also important in viral traffic. Neuregulin 1 activates ERBB receptors releasing a factor, EBP1, known to inhibit the influenza virus transcriptase. Other viral or bacterial components bind to genes or proteins encoded by CALR, FEZ1, FYN, HSPA1B, IL2, HTR2A, KPNA3, MED12, MED15, MICB, NQO2, PAX6, PIK3C3, RANBP5, or TP53, while the cerebral infectivity of the herpes simplex virus is modified by Apolipoprotein E (APOE). Genes encoding for proteins related to the innate immune response, including cytokine related (CCR5, CSF2RA, CSF2RB, IL1B, IL1RN, IL2, IL3, IL3RA, IL4, IL10, IL10RA, IL18RAP, lymphotoxin-alpha, tumor necrosis factor alpha [TNF]), human leukocyte antigen (HLA) antigens (HLA-A10, HLA-B, HLA-DRB1), and genes involved in antigen processing (angiotensin-converting enzyme and tripeptidyl peptidase 2) are all concerned with defense against invading pathogens. Human microRNAs (Hsa-mir-198 and Hsa-mir-206) are predicted to bind

  5. 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.

  6. 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.

  7. 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

  8. 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)

  9. 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.

  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

    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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  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

    2011-10-04

    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. 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.

  1. Programming for a nuclear reactor instrument simulator

    SciTech Connect

    Cohn, C.E.

    1989-01-01

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

  2. Plasma core reactor applications

    NASA Technical Reports Server (NTRS)

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

    1976-01-01

    Analytical and experimental investigations were conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration. Axial working fluid channels are located along a fraction of each cavity peripheral wall. Results of calculations for outward-directed radiant energy fluxes corresponding to radiating temperatures of 2000 to 5000 K indicate total operating pressures from 80 to 650 atm, centerline temperatures from 6900 to 30,000 K, and total radiated powers from 25 to 2500 MW, respectively. Applications are described for this type of reactor such as (1) high-thrust, high specific impulse space propulsion, (2) highly efficient systems for generation of electricity, and (3) hydrogen or synthetic fuel production systems using the intense radiant energy fluxes.

  3. Experimental demonstration of 608Gbit/s short reach transmission employing half-cycle 16QAM Nyquist-SCM signal and direct detection with 25Gbps EML.

    PubMed

    Zhong, Kangping; Zhou, Xian; Wang, Yiguang; Wang, Liang; Yuan, Jinhui; Yu, Changyuan; Lau, Alan Pak Tao; Lu, Chao

    2016-10-31

    In this paper, we experimentally demonstrated an IM/DD short reach transmission system with a total capacity of 608Gbit/s (net capacity of 565.4Gbit/s exclude 7% FEC overhead) employing half-cycle 16QAM Nyquist-SCM signal and 25Gbps EML at O band. Direct detection-faster than Nyquist (DD-FTN) technique was employed to compensate channel impairments. Number of taps of DD-LMS and tap coefficient of post filter in DD-FTN were experimentally studied for different baud rates. Single-lane 152Gbit/s transmission over 10km of SSMF was experimentally demonstrated. Employing a 4-lanes LAN-WDM architecture, a total capacity of 608Gbit/s transmission over 2km was successfully achieved with a receiver sensitivity lower than -4dBm. To the best of authors' knowledge, this is the highest reported baud rate of half-cycle 16QAM Nyquist-SCM signal and the highest bit rate employing IM/DD and 25Gbps EML in a four lanes LAN-WDM architecture for short reach systems in the O band.

  4. Cycling hypoxia affects cell invasion and proliferation through direct regulation of claudin1 / claudin7 expression, and indirect regulation of P18 through claudin7.

    PubMed

    Liu, Hong; Jiang, Feifei; Jia, Xinshan; Lan, Jing; Guo, Hao; Li, Erran; Yan, Aihui; Wang, Yan

    2017-02-07

    Claudins (CLDNs), the major integral membrane proteins at tight junction, play critical roles in apical cell-to-cell adhesion, maintenance of epithelial polarity, and formation of impermeable barriers between epithelial cells.We investigated in this study the expression of CLDNs- Claudin1 (CLDN1) and Claudin7 (CLDN7), and their relation to tumor progression in nasopharyngeal cancer (NPC). CLDN7, rather than CLDN1, showed higher expression in both undifferentiated tumor tissue and the poorly differentiated CNE2 cells, compared with differentiated tissue and the highly differentiated CNE1 cells. Furthermore, knockdown of CLDN7 dramatically inhibited the metastasis and invasion of CNE2 cells suggesting that CLDN7 could act as a biomarker for NPC metastasis.Cycling hypoxia could induce significant changes in CLDN1 and CLDN7 expression in NPC cells. Genetics analysis demonstrated that CLDN1/CLDN7 were not only regulated directly by HIF1a but also affected each other through a feedback mechanism. CLDN7 acted as a bridge to promote HIF1a-induced P18 expression and cell differentiation. Taken together, our results provide evidence that adjusting the oxygenation time and cycles in NPC might be an effective method to prevent / delay the metastasis of poorly differentiated NPC cells.

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

    SciTech Connect

    Not Available

    1980-06-25

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

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

    SciTech Connect

    Not Available

    1980-03-07

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

  7. Advanced 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. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. NUCLEAR REACTORS

    DOEpatents

    Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

    1961-12-01

    An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

  14. Neutronic reactor

    DOEpatents

    Lewis, Warren R.

    1978-05-30

    A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

  15. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1961-01-24

    A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

  16. 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.

  17. Nuclear reactor vessel fuel thermal insulating barrier

    DOEpatents

    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.

  18. REACTOR CONTROL

    DOEpatents

    Ruano, W.J.

    1957-12-10

    This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

  19. Cermet-fueled reactors for advanced space applications

    SciTech Connect

    Cowan, C.L.; Palmer, R.S.; Taylor, I.N.; Vaidyanathan, S.; Bhattacharyya, S.K.; Barner, J.O.

    1987-12-01

    Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel were carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper.

  20. 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; ...

    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

  1. 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.

  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. Neutron shielding panels for reactor pressure vessels

    SciTech Connect

    Singleton, Norman R

    2011-11-22

    In a nuclear reactor neutron panels varying in thickness in the circumferential direction are disposed at spaced circumferential locations around the reactor core so that the greatest radial thickness is at the point of highest fluence with lesser thicknesses at adjacent locations where the fluence level is lower. The neutron panels are disposed between the core barrel and the interior of the reactor vessel to maintain radiation exposure to the vessel within acceptable limits.

  4. 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

  5. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

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

    NASA Astrophysics Data System (ADS)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

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

  7. Integral fast reactor safety features

    SciTech Connect

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents.

  8. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  9. REACTOR COMPONETN

    DOEpatents

    Creutz, E.C.

    1959-10-27

    A reactor fuel element comprised of a slug of fissionable material disposed in a sheath of corrosion resistantmaterial is described. The sheath is in the form of a tubular container closed at one end and is in tight-fitting engagement with the peripheral sunface of the slug. An inner cap is insented into the open end of the sheath against the slug, which end is then bent around the inner cap and welded thereto. An outer cap is then welded around its peripheny to the bent portion of the container.

  10. Photocatalytic reactor

    DOEpatents

    Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

    1999-01-19

    A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

  11. 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

    2008-10-09

    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.

  12. Proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the integral fast reactor. [Metal fuel

    SciTech Connect

    Burris, L.; Steindler, M.; Miller, W.

    1984-01-01

    The pool-type Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps: a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented.

  13. Rapid starting methanol reactor system

    DOEpatents

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

    1984-01-01

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

  14. Simplifying microbial electrosynthesis reactor design

    PubMed Central

    Giddings, Cloelle G. S.; Nevin, Kelly P.; Woodward, Trevor; Lovley, Derek R.; Butler, Caitlyn S.

    2015-01-01

    Microbial electrosynthesis, an artificial form of photosynthesis, can efficiently convert carbon dioxide into organic commodities; however, this process has only previously been demonstrated in reactors that have features likely to be a barrier to scale-up. Therefore, the possibility of simplifying reactor design by both eliminating potentiostatic control of the cathode and removing the membrane separating the anode and cathode was investigated with biofilms of Sporomusa ovata. S. ovata reduces carbon dioxide to acetate and acts as the microbial catalyst for plain graphite stick cathodes as the electron donor. In traditional ‘H-cell’ reactors, where the anode and cathode chambers were separated with a proton-selective membrane, the rates and columbic efficiencies of microbial electrosynthesis remained high when electron delivery at the cathode was powered with a direct current power source rather than with a potentiostat-poised cathode utilized in previous studies. A membrane-less reactor with a direct-current power source with the cathode and anode positioned to avoid oxygen exposure at the cathode, retained high rates of acetate production as well as high columbic and energetic efficiencies. The finding that microbial electrosynthesis is feasible without a membrane separating the anode from the cathode, coupled with a direct current power source supplying the energy for electron delivery, is expected to greatly simplify future reactor design and lower construction costs. PMID:26029199

  15. Sandia National Laboratories Medical Isotope Reactor concept.

    SciTech Connect

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

    2010-04-01

    This report describes the Sandia National Laboratories Medical Isotope Reactor and hot cell facility concepts. The reactor proposed is designed to be capable of producing 100% of the U.S. demand for the medical isotope {sup 99}Mo. The concept is novel in that the fuel for the reactor and the targets for the {sup 99}Mo production are the same. There is no driver core required. The fuel pins that are in the reactor core are processed on a 7 to 21 day irradiation cycle. The fuel is low enriched uranium oxide enriched to less than 20% {sup 235}U. The fuel pins are approximately 1 cm in diameter and 30 to 40 cm in height, clad with Zircaloy (zirconium alloy). Approximately 90 to 150 fuel pins are arranged in the core in a water pool {approx}30 ft deep. The reactor power level is 1 to 2 MW. The reactor concept is a simple design that is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days. The fuel fabrication, reactor design and operation, and {sup 99}Mo production processing use well-developed technologies that minimize the technological and licensing risks. There are no impediments that prevent this type of reactor, along with its collocated hot cell facility, from being designed, fabricated, and licensed today.

  16. Hybrid adsorptive membrane reactor

    NASA Technical Reports Server (NTRS)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  17. Hybrid adsorptive membrane reactor

    SciTech Connect

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  18. Control Means for Reactor

    DOEpatents

    Manley, J. H.

    1961-06-27

    An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

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

    SciTech Connect

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

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

  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. BOILER-SUPERHEATED REACTOR

    DOEpatents

    Heckman, T.P.

    1961-05-01

    A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

  2. MLTAP. Modular Helium Reactor Plant Transient Thermal-Hydraulic Analysis

    SciTech Connect

    Chan, T.W.; Openshaw, F.L.

    1992-11-06

    MLTAP is an integrated system transient analysis code for modular helium reactor (MHR) plants with superheated steam for Rankine power cycle and/or process heat applications. It is used for normal operational transient analyses as well as design basis/accident condition analyses with forced convection reactor cooling. MLTAP calculates the time-dependent temperatures, pressures, and flow rates for helium primary coolant and steam/water secondary coolant; reactor system and steam system structural temperatures; reactor neutronic behavior; pump, compressor, and steam turbine performance; reactivity control and other plant control systems responses; reactor and plant protection systems responses.

  3. Coupled modeling of a directly heated tubular solar receiver for supercritical carbon dioxide Brayton cycle: Optical and thermal-fluid evaluation

    SciTech Connect

    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 receiver 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.

  4. Nutrient-cycling mechanisms other than the direct absorption from soil may control forest structure and dynamics in poor Amazonian soils.

    PubMed

    Grau, Oriol; Peñuelas, Josep; Ferry, Bruno; Freycon, Vincent; Blanc, Lilian; Desprez, Mathilde; Baraloto, Christopher; Chave, Jérôme; Descroix, Laurent; Dourdain, Aurélie; Guitet, Stéphane; Janssens, Ivan A; Sardans, Jordi; Hérault, Bruno

    2017-03-23

    Tropical forests store large amounts of biomass despite they generally grow in nutrient-poor soils, suggesting that the role of soil characteristics in the structure and dynamics of tropical forests is complex. We used data for >34 000 trees from several permanent plots in French Guiana to investigate if soil characteristics could predict the structure (tree diameter, density and aboveground biomass), and dynamics (growth, mortality, aboveground wood productivity) of nutrient-poor tropical forests. Most variables did not covary with site-level changes in soil nutrient content, indicating that nutrient-cycling mechanisms other than the direct absorption from soil (e.g. the nutrient uptake from litter, the resorption, or the storage of nutrients in the biomass), may strongly control forest structure and dynamics. Ecosystem-level adaptations to low soil nutrient availability and long-term low levels of disturbance may help to account for the lower productivity and higher accumulation of biomass in nutrient-poor forests compared to nutrient-richer forests.

  5. 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; ...

    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

  6. Nutrient-cycling mechanisms other than the direct absorption from soil may control forest structure and dynamics in poor Amazonian soils

    PubMed Central

    Grau, Oriol; Peñuelas, Josep; Ferry, Bruno; Freycon, Vincent; Blanc, Lilian; Desprez, Mathilde; Baraloto, Christopher; Chave, Jérôme; Descroix, Laurent; Dourdain, Aurélie; Guitet, Stéphane; Janssens, Ivan A.; Sardans, Jordi; Hérault, Bruno

    2017-01-01

    Tropical forests store large amounts of biomass despite they generally grow in nutrient-poor soils, suggesting that the role of soil characteristics in the structure and dynamics of tropical forests is complex. We used data for >34 000 trees from several permanent plots in French Guiana to investigate if soil characteristics could predict the structure (tree diameter, density and aboveground biomass), and dynamics (growth, mortality, aboveground wood productivity) of nutrient-poor tropical forests. Most variables did not covary with site-level changes in soil nutrient content, indicating that nutrient-cycling mechanisms other than the direct absorption from soil (e.g. the nutrient uptake from litter, the resorption, or the storage of nutrients in the biomass), may strongly control forest structure and dynamics. Ecosystem-level adaptations to low soil nutrient availability and long-term low levels of disturbance may help to account for the lower productivity and higher accumulation of biomass in nutrient-poor forests compared to nutrient-richer forests. PMID:28332608

  7. Individual bottle brush molecules in dense 2D layers restoring high degree of extension after collapse-decollapse cycle: directly measured scaling exponent.

    PubMed

    Gallyamov, M O; Tartsch, B; Potemkin, I I; Börner, H G; Matyjaszewski, K; Khokhlov, A R; Möller, M

    2009-05-01

    We prepared dense films of adsorbed brush-like macromolecules on mica substrate by transfer of compressed Langmuir monolayers from water subphase. The main macromolecular contours in the dense films were clearly resolved by SFM. The films were subjected to successive treatments by ethanol and water vapours. In accordance with previous results for isolated macromolecules, the films underwent collapse and subsequent decollapse morphological transformations in the changing vapour environment. Statistical analysis of the macromolecular dimensions in the films allowed us to measure the values of the scaling exponent nu determining the correlation between mean lateral and linear dimensions of the macromolecules. The analysis showed that the macromolecular conformations in the film as transferred were similar to the previously described conformations of the same macromolecules deposited directly on mica as isolated chains at much lower surface densities. The determined nu was close to the 0.75 value corresponding to the 2D SAW statistics. We assumed that the molecules retained the high degree of extension during the compression step due to suppressed reorganisation of the side chains. Differently from previous observations for isolated macromolecules, the restored conformations in the dense films after collapse-decollapse cycle were more extended with the nu of about 0.73 value. A theoretical explanation of the high degree of re-extension is proposed.

  8. Fast Spectrum Molten Salt Reactor Options

    SciTech Connect

    Gehin, Jess C; Holcomb, David Eugene; Flanagan, George F; Patton, Bruce W; Howard, Rob L; Harrison, Thomas J

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

  9. NEUTRONIC REACTOR

    DOEpatents

    McGarry, R.J.

    1958-04-22

    Fluid-cooled nuclear reactors of the type that utilize finned uranium fuel elements disposed in coolant channels in a moderater are described. The coolant channels are provided with removable bushings composed of a non- fissionable material. The interior walls of the bushings have a plurality of spaced, longtudinal ribs separated by grooves which receive the fins on the fuel elements. The lands between the grooves are spaced from the fuel elements to form flow passages, and the size of the now passages progressively decreases as the dlstance from the center of the core increases for the purpose of producing a greater cooling effect at the center to maintain a uniform temperature throughout the core.

  10. NEUTRONIC REACTOR

    DOEpatents

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  11. Integral fast reactor concept inherent safety features

    SciTech Connect

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS).

  12. Integral Fast Reactor concept inherent safety features

    SciTech Connect

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS).

  13. FBR and RBR particle bed space reactors

    SciTech Connect

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

    1983-01-01

    Compact, high-performance nuclear reactor designs based on High-Temperature Gas Reactors (HTGRs) particulate fuel are investigated. The large surface area available with the small-diameter (approx. 500 microns) particulate fuel allows very high power densities (MW's/liter), small temperature differences between fuel and coolant (approx. 10/sup 0/K), high coolant-outlet temperatures (1500 to 3000/sup 0/K, depending on design), and fast reactor startup (approx. 2 to 3 seconds). Two reactor concepts are developed - the Fixed Bed Reactor (FBR), where the fuel particles are packed into a thin annular bed between two porous cylindrical drums, and the Rotating Bed Reactor (RBR), where the fuel particles are held inside a cold rotating (typically approx. 500 rpm) porous cylindrical drum. The FBR can operate steady-state in the closed-cycle He-cooled mode or in the open-cycle H/sub 2/-cooled mode. The RBR will operate only in the open-cycle H/sub 2/-cooled mode.

  14. Supply of enriched uranium for research reactors

    SciTech Connect

    Mueller, H.

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  15. Neutronics Modeling of the High Flux Isotope Reactor using COMSOL

    SciTech Connect

    Chandler, David; Primm, Trent; Freels, James D; Maldonado, G Ivan

    2011-01-01

    The High Flux Isotope Reactor located at the Oak Ridge National Laboratory is a versatile 85 MWth research reactor with cold and thermal neutron scattering, materials irradiation, isotope production, and neutron activation analysis capabilities. HFIR staff members are currently in the process of updating the thermal hydraulic and reactor transient modeling methodologies. COMSOL Multiphysics has been adopted for the thermal hydraulic analyses and has proven to be a powerful finite-element-based simulation tool for solving multiple physics-based systems of partial and ordinary differential equations. Modeling reactor transients is a challenging task because of the coupling of neutronics, heat transfer, and hydrodynamics. This paper presents a preliminary COMSOL-based neutronics study performed by creating a two-dimensional, two-group, diffusion neutronics model of HFIR to study the spatially-dependent, beginning-of-cycle fast and thermal neutron fluxes. The 238-group ENDF/B-VII neutron cross section library and NEWT, a two-dimensional, discrete-ordinates neutron transport code within the SCALE 6 code package, were used to calculate the two-group neutron cross sections required to solve the diffusion equations. The two-group diffusion equations were implemented in the COMSOL coefficient form PDE application mode and were solved via eigenvalue analysis using a direct (PARDISO) linear system solver. A COMSOL-provided adaptive mesh refinement algorithm was used to increase the number of elements in areas of largest numerical error to increase the accuracy of the solution. The flux distributions calculated by means of COMSOL/SCALE compare well with those calculated with benchmarked three-dimensional MCNP and KENO models, a necessary first step along the path to implementing two- and three-dimensional models of HFIR in COMSOL for the purpose of studying the spatial dependence of transient-induced behavior in the reactor core.

  16. High flux reactor

    DOEpatents

    Lake, James A.; Heath, Russell L.; Liebenthal, John L.; DeBoisblanc, Deslonde R.; Leyse, Carl F.; Parsons, Kent; Ryskamp, John M.; Wadkins, Robert P.; Harker, Yale D.; Fillmore, Gary N.; Oh, Chang H.

    1988-01-01

    A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

  17. PINCHED PLASMA REACTOR

    DOEpatents

    Phillips, J.A.; Suydam, R.; Tuck, J.L.

    1961-07-01

    BS>A plasma confining and heating reactor is described which has the form of a torus with a B/sub 2/ producing winding on the outside of the torus and a helical winding of insulated overlapping tunns on the inside of the torus. The inner helical winding performs the double function of shielding the plasma from the vitreous container and generating a second B/sub z/ field in the opposite direction to the first B/sub z/ field after the pinch is established.

  18. Reactor and method of operation

    DOEpatents

    Wheeler, John A.

    1976-08-10

    A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

  19. 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

  20. Sequences contained within the promoter of the human thymidine kinase gene can direct cell-cycle regulation of heterologous fusion genes

    SciTech Connect

    Kim, Yongkyu; Wells, S.; Lau, Yunfai Chris; Lee, A.S. )

    1988-08-01

    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{prime} 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 {minus}441 to {minus}63 nucleotides from the transcriptional initiation site. This region ({minus}441 to {minus}63) also confers cell-cycle regulation to the herpes simplex virus thymidine kinase (HSVtk) promoter, which is not expressed in a cell-cycle manner. The authors conclude that the {minus}441 to {minus}63 sequence within the human TK promoter is important for cell-cycle-dependent expression.

  1. Cascade ICF power reactor

    SciTech Connect

    Hogan, W.J.; Pitts, J.H.

    1986-05-20

    The double-cone-shaped Cascade reaction chamber rotates at 50 rpm to keep a blanket of ceramic granules in place against the wall as they slide from the poles to the exit slots at the equator. The 1 m-thick blanket consists of layers of carbon, beryllium oxide, and lithium aluminate granules about 1 mm in diameter. The x rays and debris are stopped in the carbon granules; the neutrons are multiplied and moderated in the BeO and breed tritium in the LiAlO/sub 2/. The chamber wall is made up of SiO tiles held in compression by a network of composite SiC/Al tendons. Cascade operates at a 5 Hz pulse rate with 300 MJ in each pulse. The temperature in the blanket reaches 1600 K on the inner surface and 1350 K at the outer edge. The granules are automatically thrown into three separate vacuum heat exchangers where they give up their energy to high pressure helium. The helium is used in a Brayton cycle to obtain a thermal-to-electric conversion efficiency of 55%. Studies have been done on neutron activation, debris recovery, vaporization and recondensation of blanket material, tritium control and recovery, fire safety, and cost. These studies indicate that Cascade appears to be a promising ICF reactor candidate from all standpoints. At the 1000 MWe size, electricity could be made for about the same cost as in a future fission reactor.

  2. Final Report on Isotope Ratio Techniques for Light Water Reactors

    SciTech Connect

    Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Mitchell, Mark R.; Meriwether, George H.; Reid, Bruce D.

    2009-07-01

    The Isotope Ratio Method (IRM) is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods.

  3. 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.

  4. Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

    NASA Astrophysics Data System (ADS)

    Lubina, A. S.; Subbotin, A. S.; Sedov, A. A.; Frolov, A. A.

    2016-12-01

    The fast sodium reactor fuel assembly (FA) with U-Pu-Zr metallic fuel is described. In comparison with a "classical" fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.

  5. Advanced High Temperature Reactor Systems and Economic Analysis

    SciTech Connect

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

    2011-09-01

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

  6. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Progress Report for Work Through September 2003, 2nd Annual/8th Quarterly Report

    SciTech Connect

    Philip E. MacDonald

    2003-09-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation-IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus the need for recirculation and jet pumps, a pressurizer, steam generators, steam separators and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which is also in use around the world.

  7. Reactor safety method

    DOEpatents

    Vachon, Lawrence J.

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  8. Modeling of Reactor Kinetics and Dynamics

    SciTech Connect

    Matthew Johnson; Scott Lucas; Pavel Tsvetkov

    2010-09-01

    In order to model a full fuel cycle in a nuclear reactor, it is necessary to simulate the short time-scale kinetic behavior of the reactor as well as the long time-scale dynamics that occur with fuel burnup. The former is modeled using the point kinetics equations, while the latter is modeled by coupling fuel burnup equations with the kinetics equations. When the equations are solved simultaneously with a nonlinear equation solver, the end result is a code with the unique capability of modeling transients at any time during a fuel cycle.

  9. 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.

  10. MicroRNA 543 suppresses breast cancer cell proliferation, blocks cell cycle and induces cell apoptosis via direct targeting of ERK/MAPK

    PubMed Central

    Chen, Po; Xu, Wentao; Luo, Yi; Zhang, Yi; He, Yi; Yang, Shuo; Yuan, Zhijun

    2017-01-01

    Background Breast cancer affects millions of people with a high mortality rate throughout the world; microRNA 543 (miR-543) has been reported to suppress progression in some kinds of cancers, but has not been reported in breast cancer. Thus, the purpose of this study is to investigate the function of miR-543 in breast cancer cells. Methods Two cell lines, MCF-7 and MDA-MB-231, were selected to be the research objects; the miR-543 overexpression and knockdown models were established in the study by transforming miR-543 mimics and miR-543 inhibitor. Real-time polymerase chain reaction, 3-(4,5-dimethylthiazol-2-yl)-2,5-diphenyltetrazolium bromide, Western blot, clone formation and cell flow cytometer assay were used to test the miR-543’s function. Dual-luciferase assay was used for the detection of miR-543 and ERK2 targeting relationship. Results The results showed that the cell proliferation and cell cycle were inhibited, and the capability of cell apoptosis was upregulated when miR-543 was overexpressed; we found that there was a target relationship between ERK2 and miR-543. Furthermore, downstream factors of mitogen-activated protein kinase (MAPK)/extracellular signal-regulated kinase-2 (ERK2) pathway, including RSK2 and MSK1, were decreased in miR-543 overexpression model. Conclusion This study provides series evidences to support that breast cancer progression was inhibited by miR-543 via direct targeting of ERK2 in MAPK/ERK signal pathway, which may provide a molecular basis for better treatment for patients. PMID:28331335

  11. Nuclear reactor

    DOEpatents

    Thomson, Wallace B.

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  12. Considerations of Alloy N for Fluoride Salt-Cooled High-Temperature Reactor Applications

    SciTech Connect

    Ren, Weiju; Muralidharan, Govindarajan; Wilson, Dane F; Holcomb, David Eugene

    2011-01-01

    Fluoride Salt-Cooled High-Temperature Reactors (FHRs) are a promising new class of thermal-spectrum nuclear reactors. The reactor structural materials must possess high-temperature strength and chemical compatibility with the liquid fluoride salt as well as with a power cycle fluid such as supercritical water while remaining resistant to residual air within the containment. Alloy N was developed for use with liquid fluoride salts and it possesses adequate strength and chemical compatibility up to about 700 C. A distinctive property of FHRs is that their maximum allowable coolant temperature is restricted by their structural alloy maximum service temperature. As the reactor thermal efficiency directly increases with the maximum coolant temperature, higher temperature resistant alloys are strongly desired. This paper reviews the current status of Alloy N and its relevance to FHRs including its design principles, development history, high temperature strength, environmental resistance, metallurgical stability, component manufacturability, ASME codification status, and reactor service requirements. The review will identify issues and provide guidance for improving the alloy properties or implementing engineering solutions.

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

    SciTech Connect

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

    1981-01-01

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

  14. Plasma reactor waste management systems

    NASA Technical Reports Server (NTRS)

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

    1992-01-01

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

  15. Optimizing a neutron-beam focusing device for the direct geometry time-of-flight spectrometer TOFTOF at the FRM II reactor source

    NASA Astrophysics Data System (ADS)

    Rasmussen, N. G.; Simeoni, G. G.; Lefmann, K.

    2016-04-01

    A dedicated beam-focusing device has been designed for the direct geometry thermal-cold neutron time-of-flight spectrometer TOFTOF at the neutron facility FRM II (Garching, Germany). The prototype, based on the compressed Archimedes' mirror concept, benefits from the adaptive-optics technology (adjustable supermirror curvature) and the compact size (only 0.5 m long). We have simulated the neutron transport across the entire guide system. We present a detailed computer characterization of the existing device, along with the study of the factors mostly influencing the future improvement. We have optimized the simulated prototype as a function of the neutron wavelength, accounting also for all relevant features of a real instrument like the non-reflecting side edges. The results confirm the "chromatic" displacement of the focal point (flux density maximum) at fixed supermirror curvature, and the ability of a variable curvature to keep the focal point at the sample position. Our simulations are in excellent agreement with theoretical predictions and the experimentally measured beam profile. With respect to the possibility of a further upgrade, we find that supermirror coatings with m-values higher than 3.5 would have only marginal influence on the optimal behaviour, whereas comparable spectrometers could take advantage of longer focusing segments, with particular impact for the thermal region of the neutron spectrum.

  16. ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

    SciTech Connect

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter; Joy L. Rempe

    2015-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Current Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in

  17. Antineutrino monitoring of thorium reactors

    NASA Astrophysics Data System (ADS)

    Akindele, Oluwatomi A.; Bernstein, Adam; Norman, Eric B.

    2016-09-01

    Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuel types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring. Thorium molten salt reactors breed 233U, that if diverted constitute a direct use material as defined by the International Atomic Energy Agency (IAEA). The antineutrino spectrum from the fission of 233U has been estimated for the first time, and the feasibility of detecting the diversion of 8 kg of 233U, within a 30 day timeliness goal has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos at a 25 m standoff. It was found that the diversion of a significant quantity of 233U could not be detected within the current IAEA timeliness detection goal using either tests. A rate-time based analysis exceeded the timeliness goal by 23 days, while a spectral based analysis exceeds this goal by 31 days.

  18. Antineutrino monitoring of thorium reactors

    DOE PAGES

    Akindele, Oluwatomi A.; Bernstein, Adam; Norman, Eric B.

    2016-09-30

    Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuel types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring. Thorium molten salt reactors breed 233U, that if diverted constitute a direct use material as defined by the International Atomic Energy Agency (IAEA). The antineutrino spectrum from the fission of 233U has been estimated for the first time, and the feasibility of detecting the diversion of 8 kg ofmore » 233U, within a 30 day timeliness goal has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos at a 25 m standoff. It was found that the diversion of a significant quantity of 233U could not be detected within the current IAEA timeliness detection goal using either tests. In conclusion, a rate-time based analysis exceeded the timeliness goal by 23 days, while a spectral based analysis exceeds this goal by 31 days.« less

  19. Antineutrino monitoring of thorium reactors

    SciTech Connect

    Akindele, Oluwatomi A.; Bernstein, Adam; Norman, Eric B.

    2016-09-30

    Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuel types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring. Thorium molten salt reactors breed 233U, that if diverted constitute a direct use material as defined by the International Atomic Energy Agency (IAEA). The antineutrino spectrum from the fission of 233U has been estimated for the first time, and the feasibility of detecting the diversion of 8 kg of 233U, within a 30 day timeliness goal has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos at a 25 m standoff. It was found that the diversion of a significant quantity of 233U could not be detected within the current IAEA timeliness detection goal using either tests. In conclusion, a rate-time based analysis exceeded the timeliness goal by 23 days, while a spectral based analysis exceeds this goal by 31 days.

  20. An Integrated Fuel Depletion Calculator for Fuel Cycle Options Analysis

    SciTech Connect

    Schneider, Erich; Scopatz, Anthony

    2016-04-25

    Bright-lite is a reactor modeling software developed at the University of Texas Austin to expand upon the work done with the Bright [1] reactor modeling software. Originally, bright-lite was designed to function as a standalone reactor modeling software. However, this aim was refocused t couple bright-lite with the Cyclus fuel cycle simulator [2] to make it a module for the fuel cycle simulator.

  1. Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

    SciTech Connect

    Pope, Michael A.; Gougar, Hans D.; Ryskamp, J. M.

    2015-03-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

  2. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    NASA Astrophysics Data System (ADS)

    Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-01

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  3. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    SciTech Connect

    Bahri, Che Nor Aniza Che Zainul Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  4. 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.

  5. Modeling and analysis of tritium dynamics in a DT fusion fuel cycle

    NASA Astrophysics Data System (ADS)

    Kuan, William

    1998-11-01

    A number of crucial design issues have a profound effect on the dynamics of the tritium fuel cycle in a DT fusion reactor, where the development of appropriate solutions to these issues is of particular importance to the introduction of fusion as a commercial system. Such tritium-related issues can be classified according to their operational, safety, and economic impact to the operation of the reactor during its lifetime. Given such key design issues inherent in next generation fusion devices using the DT fuel cycle development of appropriate models can then lead to optimized designs of the fusion fuel cycle for different types of DT fusion reactors. In this work, two different types of modeling approaches are developed and their application to solving key tritium issues presented. For the first approach, time-dependent inventories, concentrations, and flow rates characterizing the main subsystems of the fuel cycle are simulated with a new dynamic modular model of a fusion reactor's fuel cycle, named X-TRUFFLES (X-Windows TRitiUm Fusion Fuel cycLE dynamic Simulation). The complex dynamic behavior of the recycled fuel within each of the modeled subsystems is investigated using this new integrated model for different reactor scenarios and design approaches. Results for a proposed fuel cycle design taking into account current technologies are presented, including sensitivity studies. Ways to minimize the tritium inventory are also assessed by examining various design options that could be used to minimize local and global tritium inventories. The second modeling approach involves an analytical model to be used for the calculation of the required tritium breeding ratio, i.e., a primary design issue which relates directly to the feasibility and economics of DT fusion systems. A time-integrated global tritium balance scheme is developed and appropriate analytical expressions are derived for tritium self-sufficiency relevant parameters. The easy exploration of the large

  6. Hot Gas Desulfurization Using Transport Reactors

    SciTech Connect

    Moorehead, E.L.

    1996-12-31

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

  7. Tokamak reactor studies

    SciTech Connect

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features.

  8. Design-development and operation of the Experimental Boiling-Water Reactor (EBWR) facility, 1955--1967

    SciTech Connect

    Boing, L.E.; Wimunc, E.A.; Whittington, G.A.

    1990-11-01

    The Experimental Boiling-Water Reactor (EBWR) was designed, built, and operated to provide experience and engineering data that would demonstrate the feasibility of the direct-cycle, boiling-water reactor and be applicable to improved, larger nuclear power stations; and was based on information obtained in the first test boiling-water reactors, the BORAX series. EBWR initially produced 20 MW(t), 5 MW(e); later modified and upgraded, as described and illustrated, it was operated at up to 100 MW(t). The facility fulfilled its primary mission -- demonstrating the practicality of the direct-boiling concept -- and, in fact, was the prototype of some of the first commercial plants and of reactor programs in some other countries. After successful completion of the Water-Cooled Reactor Program, EBWR was utilized in the joint Argonne-Hanford Plutonium Recycle Program to develop data for the utilization of plutonium as a fuel in light- water thermal systems. Final shutdown of the EBWR facility followed the termination of the latter program. 13 refs., 12 figs.

  9. First direct observation of secondary organic aerosol formation during cloud condensation-evaporation cycles in isoprene photo-oxidation reacting mixtures (CUMULUS project)

    NASA Astrophysics Data System (ADS)

    Brégonzio-Rozier, Lola; Siekmann, Frank; Giorio, Chiara; Temime-Roussel, Brice; Pangui, Edouard; Morales, Sébastien; Ravier, Sylvain; Monod, Anne; Doussin, Jean-François

    2014-05-01

    Several field observations, laboratory and model studies suggest a potentially important role of cloud droplets in forming additional secondary organic aerosol (SOA) (Sorooshian et al., 2007; Altieri et al., 2008; Couvidat et al., 2013). While this SOAaq hypothesis seems to be robust and is considered quite established, so far, no direct observations of such a process have been provided. Recently a consortium of five laboratories has joined theirs efforts in a series of experimental simulation experiments to try to bring a direct confirmation of this hypothesis: the CUMULUS project (CloUd MULtiphase chemistry of organic compoUndS in the troposphere). The aim of the present work is to study SOA formation from isoprene photo-oxidation during cloud condensation-evaporation cycles. The chemistry occurring in the gaseous, particulate and aqueous phases, and the exchange between these phases were investigated through an original multiphase approach in a simulation chamber. Experiments were performed in the CESAM chamber (Wang et al., 2011) which was designed to investigate multiphase processes under realistic actinic flux, and accurate control of both temperature and relative humidity. A protocol was designed to generate cloud events in the simulation chamber, it has allowed us to generate clouds lasting for ca. 10 minutes in the presence of light and many clouds could be generated in a single experiment. Connected to the chamber, a large panel of instruments was used to monitor the gas-phase and the particulate phase during experiments. Gas-phase composition was analyzed in-situ via a Fourier Transform Infrared Spectrometer (FTIR) and a Proton Transfer Reaction Mass Spectrometer (PTR-TOF-MS) as well as NOx and O3 analyzers. A Scanning Mobility Particle Sizer (SMPS) measured dried SOA size distributions and total concentrations inside the chamber. An Aerodyne High Resolution Time-Of-Flight Aerosol Mass Spectrometer (HR-TOF-AMS) was also used to investigate aerosol

  10. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect

    Neil Todreas; Pavel Hejzlar

    2008-06-30

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

  11. Nuclear propulsion apparatus with alternate reactor segments

    DOEpatents

    Szekely, Thomas

    1979-04-03

    1. Nuclear propulsion apparatus comprising: A. means for compressing incoming air; B. nuclear fission reactor means for heating said air; C. means for expanding a portion of the heated air to drive said compressing means; D. said nuclear fission reactor means being divided into a plurality of radially extending segments; E. means for directing a portion of the compressed air for heating through alternate segments of said reactor means and another portion of the compressed air for heating through the remaining segments of said reactor means; and F. means for further expanding the heated air from said drive means and the remaining heated air from said reactor means through nozzle means to effect reactive thrust on said apparatus.

  12. Self-actuating reactor shutdown system

    DOEpatents

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

    1988-01-01

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

  13. On-line monitoring for control of a pilot-scale sequencing batch reactor using a submersible UV/VIS spectrometer.

    PubMed

    Langergraber, G; Gupta, J K; Pressl, A; Hofstaedter, F; Lettl, W; Weingartner, A; Fleischmann, N

    2004-01-01

    A submersible UV/VIS spectrometer was used to monitor a pilot-scale sequencing batch reactor (SBR). The instrument utilises the whole UV/VIS range between 200 and 750 nm. With just one single instrument nitrate, organic matter and suspended solids can be measured simultaneously. The spectrometer is installed directly in the reactor, measures in real-time, and is equipped with an auto-cleaning system using pressured air. The paper shows the calibration results for measurements in the SBR tank, time series for typical SBR cycles, and proposes possible ways for optimisation of the operation by using these measurements.

  14. 10 CFR 171.15 - Annual fees: Reactor licenses and independent spent fuel storage licenses.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Annual fees: Reactor licenses and independent spent fuel... REACTOR LICENSES AND FUEL CYCLE LICENSES AND MATERIALS LICENSES, INCLUDING HOLDERS OF CERTIFICATES OF... NRC § 171.15 Annual fees: Reactor licenses and independent spent fuel storage licenses. (a)...

  15. 10 CFR 171.15 - Annual fees: Reactor licenses and independent spent fuel storage licenses.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Annual fees: Reactor licenses and independent spent fuel... REACTOR LICENSES AND FUEL CYCLE LICENSES AND MATERIALS LICENSES, INCLUDING HOLDERS OF CERTIFICATES OF... NRC § 171.15 Annual fees: Reactor licenses and independent spent fuel storage licenses. (a)...

  16. UCLA program in reactor studies: The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

  17. Temperature evolution on Rh/Al{sub 2}O{sub 3} catalyst during partial oxidation of methane in a reverse flow reactor

    SciTech Connect

    Simeone, M.; Menna, L.; Salemme, L.; Allouis, C.

    2010-04-15

    Catalytic partial oxidation of methane was investigated in a reverse flow reactor with commercial Rh/Al{sub 2}O{sub 3} catalyst in pellets. The process is carried out in a catalytic fixed bed reactor and switching of feed flow direction is obtained through four electrovalves synchronized in pairs. Temperature profile along the catalyst bed was measured by fast IR thermography and product composition was measured with a continuous gas analyzer. Feed direction switching time, water to methane ratio and inert section length were investigated as process parameters. Data of catalyst bed temperature evolution during the flow cycle are presented, discussed and related to reactor performance as a function of reverse flow switching period. The effect of water addition to the reacting mixture on the dynamics of catalyst bed temperature evolution is also presented. (author)

  18. Hybrid plasmachemical reactor

    SciTech Connect

    Lelevkin, V. M. Smirnova, Yu. G.; Tokarev, A. V.

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  19. Attrition reactor system

    SciTech Connect

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

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

  20. Attrition reactor system

    SciTech Connect

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

    1993-09-28

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