Science.gov

Sample records for innovative reactor systems

  1. Integrated nuclear data utilisation system for innovative reactors.

    PubMed

    Yamano, N; Hasegawa, A; Kato, K; Igashira, M

    2005-01-01

    A five-year research and development project on an integrated nuclear data utilisation system was initiated in 2002, for developing innovative nuclear energy systems such as accelerator-driven systems. The integrated nuclear data utilisation system will be constructed as a modular code system, which consists of two sub-systems: the nuclear data search and plotting sub-system, and the nuclear data processing and utilisation sub-system. The system will be operated with a graphical user interface in order to enable easy utilisation through the Internet by both nuclear design engineers and nuclear data evaluators. This paper presents an overview of the integrated nuclear data utilisation system, describes the development of a prototype system to examine the operability of the user interface and discusses specifications of the two sub-systems. PMID:16381697

  2. Integrated nuclear data utilisation system for innovative reactors.

    PubMed

    Yamano, N; Hasegawa, A; Kato, K; Igashira, M

    2005-01-01

    A five-year research and development project on an integrated nuclear data utilisation system was initiated in 2002, for developing innovative nuclear energy systems such as accelerator-driven systems. The integrated nuclear data utilisation system will be constructed as a modular code system, which consists of two sub-systems: the nuclear data search and plotting sub-system, and the nuclear data processing and utilisation sub-system. The system will be operated with a graphical user interface in order to enable easy utilisation through the Internet by both nuclear design engineers and nuclear data evaluators. This paper presents an overview of the integrated nuclear data utilisation system, describes the development of a prototype system to examine the operability of the user interface and discusses specifications of the two sub-systems.

  3. Innovative energy production in fusion reactors

    NASA Astrophysics Data System (ADS)

    Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-10-01

    Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are: (1) traveling wave direct energy conversion of 14.7 MeV protons; (2) cusp type direct energy conversion of charged particles; (3) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas; and (4) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising.

  4. Innovative Approach to Validation of Ultraviolet (UV) Reactors for Disinfection in Drinking Water Systems

    EPA Science Inventory

    Slide presentation at Conference: ASCE 7th Civil Engineering Conference in the Asian Region. USEPA in partnership with the Cadmus Group, Carollo Engineers, and other State & Industry collaborators, are evaluating new approaches for validating UV reactors to meet groundwater &...

  5. Steam Generator of the International Reactor Innovative and Secure

    SciTech Connect

    Cinotti, L.; Bruzzone, M.; Meda, N.; Corsini, G.; Lombardi, C.V.; Ricotti, M.; Conway, L.E.

    2002-07-01

    IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the main reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long-life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. The design of the steam generators, which are internally contained within the reactor vessel, is a major design effort in the development of the integral IRIS concept. The ongoing design activity about the steam generator is the subject of this paper. (authors)

  6. Reactor surface contamination stabilization. Innovative technology summary report

    SciTech Connect

    1998-11-01

    Contaminated surfaces, such as the face of a nuclear reactor, need to be stabilized (fixed) to avoid airborne contamination during decontamination and decommissioning activities, and to prepare for interim safe storage. The traditional (baseline) method of fixing the contamination has been to spray a coating on the surfaces, but ensuring complete coverage over complex shapes, such as nozzles and hoses, is difficult. The Hanford Site C Reactor Technology Demonstration Group demonstrated innovative technologies to assess stabilization properties of various coatings and to achieve complete coverage of complex surfaces on the reactor face. This demonstration was conducted in two phases: the first phase consisted of a series of laboratory assessments of various stabilization coatings on metal coupons. For the second phase, coatings that passed the laboratory tests were applied to the front face of the C Reactor and evaluated. The baseline coating (Rust-Oleum No. 769) and one of the innovative technologies did not completely cover nozzle assemblies on the reactor face, the most critical of the second-phase evaluation criteria. However, one of the innovative coating systems, consisting of a base layer of foam covered by an outer layer of a polymeric film, was successful. The baseline technology would cost approximately 33% as much as the innovative technology cost of $64,000 to stabilize an entire reactor face (196 m{sup 2} or 2116 ft{sup 2}) with 2,004 nozzle assemblies, but the baseline system failed to provide complete surface coverage.

  7. Innovations and enhancements in neutronic analysis of the Big-10 university research and training reactors based on the AGENT code system

    SciTech Connect

    Hursin, M.; Shanjie, X.; Burns, A.; Hopkins, J.; Satvat, N.; Gert, G.; Tsoukalas, L. H.; Jevremovic, T.

    2006-07-01

    Introduction. This paper summarizes salient aspects of the 'virtual' reactor system developed at Purdue Univ. emphasizing efficient neutronic modeling through AGENT (Arbitrary Geometry Neutron Transport) a deterministic neutron transport code. DOE's Big-10 Innovations in Nuclear Infrastructure and Education (INIE) Consortium was launched in 2002 to enhance scholarship activities pertaining to university research and training reactors (URTRs). Existing and next generation URTRs are powerful campus tools for nuclear engineering as well as a number of disciplines that include, but are not limited to, medicine, biology, material science, and food science. Advancing new computational environments for the analysis and configuration of URTRs is an important Big-10 INIE aim. Specifically, Big-10 INIE has pursued development of a 'virtual' reactor, an advanced computational environment to serve as a platform on which to build operations, utilization (research and education), and systemic analysis of URTRs physics. The 'virtual' reactor computational system will integrate computational tools addressing the URTR core and near core physics (transport, dynamics, fuel management and fuel configuration); thermal-hydraulics; beam line, in-core and near-core experiments; instrumentation and controls; confinement/containment and security issues. Such integrated computational environment does not currently exist. The 'virtual' reactor is designed to allow researchers and educators to configure and analyze their systems to optimize experiments, fuel locations for flux shaping, as well as detector selection and configuration. (authors)

  8. Providing the Basis for Innovative Improvements in Advanced LWR Reactor Passive Safety Systems Design: An Educational R&D Project

    SciTech Connect

    Brian G. Williams; Jim C. P. Liou; Hiral Kadakia; Bill Phoenix; Richard R. Schultz

    2007-02-27

    This project characterizes typical two-phase stratified flow conditions in advanced water reactor horizontal pipe sections, following activation of passive cooling systems. It provides (1) a means to educate nuclear engineering students regarding the importance of two-phase stratified flow in passive cooling systems to the safety of advanced reactor systems and (2) describes the experimental apparatus and process to measure key parameters essential to consider when designing passive emergency core cooling flow paths that may encounter this flow regime. Based on data collected, the state of analysis capabilities can be determined regarding stratified flow in advanced reactor systems and the best paths forward can be identified to ensure that the nuclear industry can properly characterize two-phase stratified flow in passive emergency core cooling systems.

  9. Reactor System Transient Code.

    1999-07-14

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

  10. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

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

    1959-11-01

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

  11. Improved vortex reactor system

    DOEpatents

    Diebold, James P.; Scahill, John W.

    1995-01-01

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

  12. Innovative approaches to inertial confinement fusion reactors: Final report

    SciTech Connect

    Bourque, R.F.; Schultz, K.R.

    1986-11-01

    Three areas of innovative approaches to inertial confinement fusion (ICF) reactor design are given. First, issues pertaining to the Cascade reactor concept are discussed. Then, several innovative concepts are presented which attempt to directly recover the blast energy from a fusion target. Finally, the Turbostar concept for direct recovery of that energy is evaluated. The Cascade issues discussed are combustion of the carbon granules in the event of air ingress, the use of alternate granule materials, and the effect of changes in carbon flow on details of the heat exchanger. Carbon combustion turns out to be a minor problem. Four ICF innovative concepts were considered: a turbine with ablating surfaces, a liquid piston system, a wave generator, and a resonating pump. In the final analysis, none show any real promise. The Turbostar concept of direct recovery is a very interesting idea and appeared technically viable. However, it shows no efficiency gain or any decrease in capital cost compared to reactors with conventional thermal conversion systems. Attempts to improve it by placing a close-in lithium sphere around the target to increase gas generation increased efficiency only slightly. It is concluded that these direct conversion techniques require thermalization of the x-ray and debris energy, and are Carnot limited. They therefore offer no advantage over existing and proposed methods of thermal energy conversion or direct electrical conversion.

  13. Reactor vessel support system

    DOEpatents

    Golden, Martin P.; Holley, John C.

    1982-01-01

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

  14. Reactor water cleanup system

    DOEpatents

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

    1994-12-20

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

  15. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

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

  16. HORIZONTAL BOILING REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1958-11-18

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

  17. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Goett, J.J.

    1961-01-24

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

  18. Improved vortex reactor system

    DOEpatents

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

    1995-05-09

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

  19. 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.

  20. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1959-02-10

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

  1. 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.

  2. [Stability control of aerobic granules using an innovative reactor].

    PubMed

    Li, Zhi-Hua; Yang, Fan; Li, Sheng; Xie, Lei; Wang, Xiao-Chang

    2012-06-01

    Uncontrolled variation of diameter and density of aerobic granules frequently resulted in instability and thus brought about operation failure. An innovative reactor was therefore developed for the control of diameter and density of aerobic granules. There were two ways to select the sludge, one was the short settling time select the big and dense granules in the reactor, and the other was the hydro cyclone that washed out the big and compact granules preventing big and compact fourthly growth in the reactor. By these means, the diameter of granules could maintained in the range of 300-1 000 microm for a long time, consequently, the long term stability could be obtained. According to the kinetic analysis, it was found that the energy maintenance coefficient was 0.08-0.10, which was much higher than the conventional granular system (0.06), and the ratio of the COD used for maintenance to the influent was higher than the conventional one. Additionally, the removal efficiencies of COD and ammonia were 92% and 60%, respectively.

  3. A brief history of design studies on innovative nuclear reactors

    SciTech Connect

    Sekimoto, Hiroshi

    2014-09-30

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  4. A brief history of design studies on innovative nuclear reactors

    NASA Astrophysics Data System (ADS)

    Sekimoto, Hiroshi

    2014-09-01

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  5. Nuclear reactor sealing system

    DOEpatents

    McEdwards, James A.

    1983-01-01

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

  6. Nuclear reactor shutdown system

    DOEpatents

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

    1981-01-01

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

  7. 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.

  8. 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.

  9. Reactor vessel support system. [LMFBR

    DOEpatents

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

    1980-05-09

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

  10. FLOW SYSTEM FOR REACTOR

    DOEpatents

    Zinn, W.H.

    1963-06-11

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

  11. NUCLEAR REACTOR FUEL SYSTEMS

    DOEpatents

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

    1959-09-15

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

  12. An Innovative Reactor Technology to Improve Indoor Air Quality

    SciTech Connect

    Rempel, Jane

    2013-03-30

    As residential buildings achieve tighter envelopes in order to minimize energy used for space heating and cooling, accumulation of indoor air pollutants such as volatile organic compounds (VOCs), becomes a major concern causing poor air quality and increased health risks. Current VOC removal methods include sorbents, ultraviolet photocatalytic oxidation (UVPCO), and increased ventilation, but these methods do not capture or destroy all VOCs or are prohibitively expensive to implement. TIAX's objective in this program was to develop a new VOC removal technology for residential buildings. This novel air purification technology is based on an innovative reactor and light source design along with UVPCO properties of the chosen catalyst to purify indoor air and enhance indoor air quality (IAQ). During the program we designed, fabricated and tested a prototype air purifier to demonstrate its feasibility and effectiveness. We also measured kinetics of VOC destruction on photocatalysts, providing deep insight into reactor design.

  13. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    SciTech Connect

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

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

    SciTech Connect

    Not Available

    1987-06-01

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

  15. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Daniels, F.

    1957-10-15

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

  16. SYSTEM FOR UNLOADING REACTORS

    DOEpatents

    Rand, A.C. Jr.

    1961-05-01

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

  17. Reactor refueling containment system

    DOEpatents

    Gillett, James E.; Meuschke, Robert E.

    1995-01-01

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

  18. Reactor refueling containment system

    DOEpatents

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

    1995-05-02

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

  19. Reactor core isolation cooling system

    DOEpatents

    Cooke, F.E.

    1992-12-08

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

  20. Reactor core isolation cooling system

    DOEpatents

    Cooke, Franklin E.

    1992-01-01

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

  1. Advanced Reactor Innovation Evaluation Study (ARIES) Properties Archive

    DOE Data Explorer

    ARIES stands for Advanced Reactor Innovation Evaluation Study. It is a program and a team that explores the commercial potential of fusion as an energy resource. Though it is a multi-institutional program, ARIES is led by the University of California at San Diego. ARIES studies both magnetic fusion energy (MFE) and inertial fusion energy (IFE), using an approach that integrates theory, experiments, and technology. The ARIES team proposes fusion reactor designs and works to understand how technology, materials and plasma physics processes interact and influence each other. A 2005 report to the Fusion Energy Sciences Advisory Committee ("Scientific Challenges, Opportunities, and Priorities for the U.S. Fusion Energy Sciences Program") noted on page 98 an example of the importance of this materials properties aspect: "For instance, effects on plasma edge by various plasma facing materials and effects on various plasma stabilization and control techniques by highly conducting liquid metal blankets are being considered by physicists." This web page is an archive of material properties collected here for the use of the ARIES Fusion Power Plant Studies Team.

  2. 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.

  3. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

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

    1961-06-27

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

  4. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, R.M.

    1983-11-08

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

  5. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, Robert M.

    1983-01-01

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

  6. Rapid starting methanol reactor system

    DOEpatents

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

    1984-01-01

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

  7. Reactor control rod timing system

    DOEpatents

    Wu, Peter T. K.

    1982-01-01

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

  8. Reactor control rod timing system

    SciTech Connect

    Wu, P.T.

    1982-02-09

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

  9. Innovating blackberry production system

    Technology Transfer Automated Retrieval System (TEKTRAN)

    This article provides an overview of progress made on the blackberry production system at the Appalachian Fruit Research Station in Kearneysville, WV. Several lines of research have made a significant positive impact to the blackberry industry as the new technology has have been successfully transf...

  10. Reactor system for upgrading light olefins in staged reactors

    SciTech Connect

    Harandi, M.N.; Owen, H.

    1991-05-28

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

  11. Reactor vessel annealing system

    DOEpatents

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

    1991-01-01

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

  12. Design and analysis of a nuclear reactor core for innovative small light water reactors

    NASA Astrophysics Data System (ADS)

    Soldatov, Alexey I.

    In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

  13. POWER GENERATING NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Vernon, H.C.

    1958-03-01

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

  14. Fast breeder reactor protection system

    DOEpatents

    van Erp, J.B.

    1973-10-01

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

  15. Liquid metal cooled nuclear reactor plant system

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

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

  16. Design requirements for innovative homogeneous reactor, lesson learned from Fukushima accident

    NASA Astrophysics Data System (ADS)

    Arbie, Bakri; Pinem, Suryan; Sembiring, Tagor; Subki, Iyos

    2012-06-01

    The Fukushima disaster is the largest nuclear accident since the 1986 Chernobyl disaster, but it is more complex as multiple reactors and spent fuel pools are involved. The severity of the nuclear accident is rated 7 in the International Nuclear Events Scale. Expert said that "Fukushima is the biggest industrial catastrophe in the history of mankind". According to Mitsuru Obe, in The Wall Street Journal, May 16th of 2011, TEPCO estimates the nuclear fuel was exposed to the air less than five hours after the earthquake struck. Fuel rods melted away rapidly as the temperatures inside the core reached 2800 C within six hours. In less than 16 hours, the reactor core melted and dropped to the bottom of the pressure vessel. The information should be evaluated in detail. In Germany several nuclear power plant were shutdown, Italy postponed it's nuclear power program and China reviewed their nuclear power program. Different news come from Britain, in October 11, 2011, the Safety Committee said all clear for nuclear power in Britain, because there are no risk of strong earthquake and tsunami in the region. Due to this severe fact, many nuclear scientists and engineer from all over the world are looking for a new approach, such as homogeneous reactor which was developed in Oak Ridge National Laboratory in 1960-ies, during Dr. Alvin Weinberg tenure as the Director of ORNL. The paper will describe the design requirement that will be used as the basis for innovative homogeneous reactor. Innovative Homogeneous Reactor is expected to reduce core melt by two decades (4), since the fuel is intermix homogeneously with coolant and secondly we eliminate the used fuel rod which need to be cooled for a long period of time. In order to be successful for its implementation of the innovative system, testing and validation, three phases of development will be introduced. The first phase is Low Level Goals is really the proof of concept;the Medium Level Goal is Technical Goalsand the High

  17. CALMOS: Innovative device for the measurement of nuclear heating in material testing reactors

    SciTech Connect

    Carcreff, H.

    2011-07-01

    An R and D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed. (authors)

  18. CONTROL SYSTEM FOR NEUTRONIC REACTORS

    DOEpatents

    Crever, F.E.

    1962-05-01

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

  19. Shutdown system for a nuclear reactor

    DOEpatents

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

    1984-06-05

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

  20. Shutdown system for a nuclear reactor

    DOEpatents

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

    1984-01-01

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

  1. Inertial confinement fusion reactor systems

    SciTech Connect

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

    1980-01-01

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

  2. Options Study Documenting the Fast Reactor Fuels Innovative Design Activity

    SciTech Connect

    Jon Carmack; Kemal Pasamehmetoglu

    2010-07-01

    This document provides presentation and general analysis of innovative design concepts submitted to the FCRD Advanced Fuels Campaign by nine national laboratory teams as part of the Innovative Transmutation Fuels Concepts Call for Proposals issued on October 15, 2009 (Appendix A). Twenty one whitepapers were received and evaluated by an independent technical review committee.

  3. Tandem Mirror Reactor Systems Code (Version I)

    SciTech Connect

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

    1985-09-01

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

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

    2003-04-21

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

  5. Nuclear electric propulsion reactor control systems status

    NASA Technical Reports Server (NTRS)

    Ferg, D. A.

    1973-01-01

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

  6. A Profile of Innovative School Systems.

    ERIC Educational Resources Information Center

    Hilfiker, Leo R.

    An earlier report examined the relationship of school system innovativeness to selected dimensions of interpersonal behavior in eight school systems as revealed through data collected in 1967. The major implication of this study suggested that the climate of the schools might be changed to make them more receptive to innovation from within or from…

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

    SciTech Connect

    Not Available

    1987-06-01

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

  8. Fission control system for nuclear reactor

    DOEpatents

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

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

  9. The 5-kwe reactor thermoelectric system summary

    NASA Technical Reports Server (NTRS)

    Vanosdol, J. H. (Editor)

    1973-01-01

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

  10. A comparison of the technological effectiveness of dairy wastewater treatment in anaerobic UASB reactor and anaerobic reactor with an innovative design.

    PubMed

    Jedrzejewska-Cicinska, M; Kozak, K; Krzemieniewski, M

    2007-10-01

    The present research was an investigation of the influence of an innovative design of reactor filled with polyethylene (PE) granulate on model dairy wastewater treatment efficiency under anaerobic conditions compared to that obtained in a typical UASB reactor. The experiment was conducted at laboratory scale. An innovative reactor was designed with the reaction chamber inclined 30 degrees in relation to the ground with upward waste flow and was filled with PE granular material. Raw model dairy wastewater was fed to two anaerobic reactors of different design at the organic loading rate of 4 kg COD m(-3)d(-1). Throughout the experiment, a higher removal efficiency of organic compounds was observed in the reactor with an innovative design and it was higher by 7.1% on average than in the UASB reactor. The total suspended solids was lower in the wastewater treated in the anaerobic reactor with the innovative design. Applying a PE granulated filling in the chamber of the innovative reactor contributed to an even distribution of sludge biomass in the reactor, reducing washout of anaerobic sludge biomass from the reaction chamber and giving a higher organic compounds removal efficiency.

  11. TREAT Reactor Control and Protection System

    SciTech Connect

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

    1985-01-01

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

  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. An innovative reactor-type biosensor for BOD rapid measurement.

    PubMed

    Wang, Jianlong; Zhang, Yixin; Wang, Yeyao; Xu, Runhua; Sun, Zhonghua; Jie, Zhou

    2010-03-15

    Biochemical oxygen demand (BOD) is one of the most important and widely used parameters for characterizing the organic pollution of water and wastewater. In this paper, a novel reactor-type biosensor for rapid measurement of BOD was developed, based on using immobilized microbial cell (IMC) beads as recognition bio-element in a completely mixed reactor which was used as determining chamber, replacing the traditionally used membrane as recognition bio-element. The IMC beads were freely suspended in the aqueous solution, so the mass transfer resistance for dissolved oxygen and organic compounds significantly reduced, and the quantity of the microbial cells used as recognition element can be easily adjusted, in comparison with the traditional membrane-type BOD biosensor, in which exists a unadjustable contradiction between the quantity of biomass and the thickness of the bio-membrane, thus limiting the stability and the detection limit. This novel kind of BOD biosensor significantly increased the sensitivity of the response, the detecting precision and prolonged the life time of the recognition element. The experimental data showed that the most appropriate temperature for biochemical reaction in the reactor was 30 degrees C, and the IMC beads could keep the bioactivity for about 70d at the detecting frequency of 8 times every day. The standard deviation of repeatability and the reproducibility of responses were within +/-6.4% and +/-5.0%, respectively, which are within acceptable bias limits, and meet the requirement of BOD rapid measurement.

  14. SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

    DOEpatents

    Dickinson, R.W.

    1963-03-01

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

  15. Reactor system for olefin conversion and etherification

    SciTech Connect

    Harandi, M.N.

    1990-09-18

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

  16. Scanning tunneling microscope assembly, reactor, and system

    SciTech Connect

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

    2014-11-18

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

  17. 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.

  18. REACTOR CONTROL ROD OPERATING SYSTEM

    DOEpatents

    Miller, G.

    1961-12-12

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

  19. Innovative power conversion system for the French SFR prototype, ASTRID

    SciTech Connect

    Cachon, L.; Biscarrat, C.; Morin, F.; Haubensack, D.; Rigal, E.; Moro, I.; Baque, F.; Madeleine, S.; Rodriguez, G.; Laffont, G.

    2012-07-01

    In the framework of the French Act of 28 June 2006 about nuclear materials and waste management, the prototype ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), foreseen in operation by the 20's, will have to demonstrate not only the minor actinide transmutation capability, but also the progress made in Sodium Fast Reactor (SFR) technology on an industrial scale, by qualifying innovative options. Some of these options still require improvements, especially in the field of operability and safety. In fact, one of the main issues with the standard steam/water Power Conversion System (PCS) of SFR is the fast and energetic chemical reaction between water and sodium, which could occur in steam generators in case of tube failure. To manage the sodium/water reaction, one way consists in minimizing the impact of such event: hence studies are carried out on steam generator design, improvement of the physical knowledge of this phenomenon, development of numerical simulation to predict the reaction onset and consequences, and associated detection improvement. On the other hand, the other way consists in eliminating sodium/water reaction. In this frame, the CEA contribution to the feasibility evaluation of an alternative innovative PCS (replacing steam/water by 180 bar pressurised nitrogen) is focused on the following main topics: - The parametric study leading to nitrogen selection: the thermodynamic cycle efficiency optimisation on Brayton cycles is performed with several gases at different pressures. - The design of innovative compact heat exchangers for the gas loop: here the key points are the nuclear codification associated with inspection capability, the innovative welding process and the thermal-hydraulic and thermal-mechanic optimisations. After a general introduction of the ASTRID project, this paper presents in detail these different feasibility studies being led on the innovative gas PCS for an SFR. (authors)

  20. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  1. Control system for a small fission reactor

    DOEpatents

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

    1986-01-01

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

  2. Control system for a small fission reactor

    DOEpatents

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

    1985-02-08

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

  3. NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

    DOEpatents

    Moore, W.T.

    1958-09-01

    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  4. Organizational innovation and the laboratory information system.

    PubMed

    Friedman, B A; Mitchell, W

    1990-01-01

    There is an urgent need for more innovation in health care and in clinical laboratories. Innovation can be divided into five separate categories: core products, components of the core products, production processes, uses of products, and the organization of production. Organizational innovation is an important as technological advances in increasing the quality and efficiency of clinical laboratories. The use of a laboratory information system (LIS) can stimulate organizational innovation such as the assignment of computer-oriented tasks to personnel within individual clinical laboratories. The authors refer to such LIS support personnel as "hidden personnel" and suggest that such a shift of responsibility empowers laboratory professionals and makes them stakeholders in automated information processing.

  5. Pressure Vessel and Internals of the International Reactor Innovative and Secure

    SciTech Connect

    Lombardi, C.V.; Padovani, E.; Cammi, A.; Collado, J.M.; Santoro, R.T.; Barnes, J.M.

    2002-07-01

    IRIS (International Reactor Innovative and Secure) is a modular, integral light water cooled, low-to-medium power reactor, which addresses the requirements defined by the US DOE for Generation IV reactors. Its integrated layout features a pressure vessel containing all the main primary circuit components: the internals and the biological shield, here described together with the pressure vessel, plus the steam generators, the pressurizer, and the main coolant pumps described in companion papers. For this reason the pressure vessel is a crucial component of the plant, which deserves the most demanding design effort. The wide inner annulus around the core is exploited to insert steel plates, in order to improve the inner shielding capability up to the elimination of the external biological shielding and to simplify decommissioning activities by having all the irradiated components inside the vessel. (authors)

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

    1999-10-19

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

  7. An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor

    SciTech Connect

    Haihua Zhao; Hongbin Zhang

    2007-11-01

    The existing sodium cooled fast reactors (SFR) have two types of designs – loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphénix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL’s Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed.

  8. Innovative Design of Complex Engineering Systems

    NASA Technical Reports Server (NTRS)

    Noor, Ahmed K. (Compiler)

    2004-01-01

    The document contains the proceedings of the training workshop on Innovative Design of Complex Engineering Systems. The workshop was held at the Peninsula Higher Education Center, Hampton, Virginia, March 23 and 24, 2004. The workshop was jointly sponsored by Old Dominion University and NASA. Workshop attendees came from NASA, other government agencies, industry and universities. The objectives of the workshop were to a) provide broad overviews of the diverse activities related to innovative design of high-tech engineering systems; and b) identify training needs for future aerospace work force development in the design area. The format of the workshop included fifteen, half-hour overview-type presentations, a panel discussion on how to teach and train engineers in innovative design, and three exhibits by commercial vendors.

  9. Comparing Three Innovative Instructional Systems.

    ERIC Educational Resources Information Center

    Dunn, Rita; Waggoner, Barbara

    1995-01-01

    Examines the differences between three instructional systems: learning styles, neuro-linguistic programming, and "Suggestopoedia." Topics include the philosophical basis of each system; teaching methodologies; the use of multimedia; the use of time; environmental settings; and approaches to human emotion, individual sociological differences, and…

  10. Systems Innovation and Education Management Systems (EMS)

    ERIC Educational Resources Information Center

    Rao, Nageswararao A. V.

    2006-01-01

    Many researchers and practitioners contend that all institutions respond to changing market need and can create competitive advantage through innovation and creativity. Each year, institutions expend significant resources developing new products and processes and yet research shows that more than half these initiatives fail. Successful…

  11. The Secure, Transportable, Autonomous Reactor System

    SciTech Connect

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

    1999-05-27

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

  12. What Is a Complex Innovation System?

    PubMed Central

    Katz, J. Sylvan

    2016-01-01

    Innovation systems are sometimes referred to as complex systems, something that is intuitively understood but poorly defined. A complex system dynamically evolves in non-linear ways giving it unique properties that distinguish it from other systems. In particular, a common signature of complex systems is scale-invariant emergent properties. A scale-invariant property can be identified because it is solely described by a power law function, f(x) = kxα, where the exponent, α, is a measure of scale-invariance. The focus of this paper is to describe and illustrate that innovation systems have properties of a complex adaptive system. In particular scale-invariant emergent properties indicative of their complex nature that can be quantified and used to inform public policy. The global research system is an example of an innovation system. Peer-reviewed publications containing knowledge are a characteristic output. Citations or references to these articles are an indirect measure of the impact the knowledge has on the research community. Peer-reviewed papers indexed in Scopus and in the Web of Science were used as data sources to produce measures of sizes and impact. These measures are used to illustrate how scale-invariant properties can be identified and quantified. It is demonstrated that the distribution of impact has a reasonable likelihood of being scale-invariant with scaling exponents that tended toward a value of less than 3.0 with the passage of time and decreasing group sizes. Scale-invariant correlations are shown between the evolution of impact and size with time and between field impact and sizes at points in time. The recursive or self-similar nature of scale-invariance suggests that any smaller innovation system within the global research system is likely to be complex with scale-invariant properties too. PMID:27258040

  13. What Is a Complex Innovation System?

    PubMed

    Katz, J Sylvan

    2016-01-01

    Innovation systems are sometimes referred to as complex systems, something that is intuitively understood but poorly defined. A complex system dynamically evolves in non-linear ways giving it unique properties that distinguish it from other systems. In particular, a common signature of complex systems is scale-invariant emergent properties. A scale-invariant property can be identified because it is solely described by a power law function, f(x) = kxα, where the exponent, α, is a measure of scale-invariance. The focus of this paper is to describe and illustrate that innovation systems have properties of a complex adaptive system. In particular scale-invariant emergent properties indicative of their complex nature that can be quantified and used to inform public policy. The global research system is an example of an innovation system. Peer-reviewed publications containing knowledge are a characteristic output. Citations or references to these articles are an indirect measure of the impact the knowledge has on the research community. Peer-reviewed papers indexed in Scopus and in the Web of Science were used as data sources to produce measures of sizes and impact. These measures are used to illustrate how scale-invariant properties can be identified and quantified. It is demonstrated that the distribution of impact has a reasonable likelihood of being scale-invariant with scaling exponents that tended toward a value of less than 3.0 with the passage of time and decreasing group sizes. Scale-invariant correlations are shown between the evolution of impact and size with time and between field impact and sizes at points in time. The recursive or self-similar nature of scale-invariance suggests that any smaller innovation system within the global research system is likely to be complex with scale-invariant properties too.

  14. Dynamic Impregnator Reactor System (Poster)

    SciTech Connect

    Not Available

    2012-09-01

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

  15. NON-CORROSIVE REACTOR FUEL SYSTEM

    DOEpatents

    Herrick, C.C.

    1962-08-14

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

  16. Hybrid Molten Salt Reactor (HMSR) System Study

    SciTech Connect

    Woolley, Robert D; Miller, Laurence F

    2014-04-01

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

  17. Rodded shutdown system for a nuclear reactor

    DOEpatents

    Golden, Martin P.; Govi, Aldo R.

    1978-01-01

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

  18. The CANDU Reactor System: An Appropriate Technology.

    PubMed

    Robertson, J A

    1978-02-10

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

  19. Reactor control rod timing system. [LMFBR

    DOEpatents

    Wu, P.T.K.

    1980-03-18

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

  20. Control system studies for thermionic reactors

    NASA Technical Reports Server (NTRS)

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

    1978-01-01

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

  1. Gaseous fuel reactor systems for aerospace applications

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schwenk, F. C.

    1977-01-01

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

  2. Fundamental Considerations Of Gas Core Reactor Systems

    NASA Astrophysics Data System (ADS)

    Bussard, Robert W.

    1988-04-01

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

  3. Using Innovative Information Systems Techniques To Teach Information Systems.

    ERIC Educational Resources Information Center

    Chimi, Carl J.; Gordon, Gene M.

    This paper discusses a number of innovative techniques that were used to teach courses in Information Systems to undergraduate and graduate students. While none of these techniques is individually innovative, the combination of techniques provides a true "hands-on" environment for students; because of the way that the components of the courses are…

  4. Innovative shotcreting system in Japan

    SciTech Connect

    Okada, Takashi

    1995-12-31

    Robotization of shotcreting has advanced remarkably in Japan in a short span of time for the purpose of avoiding exposure of human workers to adverse conditions. This paper provides an overview of various devices used in shotcreting and discusses the development of improved systems which ensure higher safety and larger productivity.

  5. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

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

  6. Reactor power system deployment and startup

    NASA Technical Reports Server (NTRS)

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

    1985-01-01

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

  7. Fault-tolerant reactor protection system

    DOEpatents

    Gaubatz, Donald C.

    1997-01-01

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

  8. Fault-tolerant reactor protection system

    DOEpatents

    Gaubatz, D.C.

    1997-04-15

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

  9. Parallel reactor systems for bioprocess development.

    PubMed

    Weuster-Botz, Dirk

    2005-01-01

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

  10. Direct conversion nuclear reactor space power systems

    SciTech Connect

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

    1982-08-01

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

  11. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, Philippe

    1994-01-01

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

  12. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, P.

    1994-07-05

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

  13. Integral reactor system and method for fuel cells

    DOEpatents

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

    2013-11-19

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

  14. Static conversion systems. [for space power reactors

    NASA Technical Reports Server (NTRS)

    Ewell, R.; Mondt, J.

    1985-01-01

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

  15. The PCAST Energy Technology Innovation System Study

    NASA Astrophysics Data System (ADS)

    Savitz, M.; Fri, R.

    2010-12-01

    The President's Council of Advisors on Science and Technology (PCAST) recently made recommendations for strengthening the nation's energy innovation system. The PCAST report builds in part on earlier work at the National Research Council (NRC) and elsewhere. For example, PCAST largely adopted the description of the energy innovation system that appeared in the NRC report on 'Limiting the Magnitude of Future Climate Change'. Similarly, the 'Limiting' report provided examples of the importance of social science research in crafting energy policy, a recommendation of the PCAST report. And both the 'Limiting' report and an earlier report on 'America's Energy Future' recommended an aggressive commercial demonstration program for carbon capture and storage and new nuclear power plants. The PCAST report discusses the need for new approaches for federal demonstration projects. This session traces these relationships and suggests how similar synergies might be encouraged in the future.

  16. Microchannel Reactor System for Catalytic Hydrogenation

    SciTech Connect

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

    2010-12-22

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

  17. Staged membrane oxidation reactor system

    SciTech Connect

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

    2013-04-16

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

  18. Staged membrane oxidation reactor system

    SciTech Connect

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

    2014-05-20

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

  19. Staged membrane oxidation reactor system

    DOEpatents

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

    2012-09-11

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

  20. Neutral-beam-injection systems for reactors

    SciTech Connect

    Pyle, R.V.

    1983-06-01

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

  1. Plasma generators, reactor systems and related methods

    DOEpatents

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

    2007-06-19

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

  2. Power conditioning for space nuclear reactor systems

    NASA Technical Reports Server (NTRS)

    Berman, Baruch

    1987-01-01

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

  3. Innovative reactor technology for selective oxidation of toxic organic pollutants in wastewater by ozone.

    PubMed

    Boncz, M A; Bruning, H; Rulkens, W H

    2003-01-01

    Ozonation can be a suitable technique for the pre-treatment of wastewater containing low concentrations of toxic or non-biodegradable compounds that cannot be treated with satisfactory results when only the traditional, less expensive biological techniques are applied. In this case, the oxidation process has to be made as efficient as possible, in order to reduce the coats of ozone addition and use. An efficient oxidation process with ozone can be obtained by focusing the oxidation with ozone selectively on the direct oxidation of toxic pollutants and to minimize ozone losses due to the decay of ozone in water. Supported by data of the rate constants of the reactions involved, a mathematical model was developed. It quantifies the ozone consumption by the process, and the share of ozone consumption by undesired side reactions, in several different reactor systems. Results obtained with this model indicate that a plug flow reactor (PFR) will be the most efficient design for the oxidation reactor. As an alternative, the cascaded tank reactor system (CTR), in which the ozone feed may be realized with less practical problems, might be considered. The traditional continuous flow stirred tank reactor (CFSTR) is shown to be the least efficient system.

  4. Nuclear reactor pressure vessel support system

    DOEpatents

    Sepelak, George R.

    1978-01-01

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

  5. The liquid annular reactor system (LARS) propulsion

    SciTech Connect

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

    1991-01-05

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

  6. NEUTRONIC REACTOR OPERATIONAL METHOD AND CORE SYSTEM

    DOEpatents

    Winters, C.E.; Graham, C.B.; Culver, J.S.; Wilson, R.H.

    1960-07-19

    Homogeneous neutronic reactor systems are described wherein an aqueous fuel solution is continuously circulated through a spherical core tank. The pumped fuel solution-is injected tangentially into the hollow spherical interior, thereby maintaining vigorous rotation of the solution within the tank in the form of a vortex; gaseous radiolytic decomposition products concentrate within the axial vortex cavity. The evolved gas is continuously discharged through a gas- outlet port registering with an extremity of the vortex cavity. and the solution stream is discharged through an annular liquid outlet port concentrically encircling the gas outlet by virtue of which the vortex and its cavity are maintained precisely axially aligned with the gas outlet. A primary heat exchanger extracts useful heat from the hot effluent fuel solution before its recirculation into the core tank. Hollow cylinders and other alternative core- tank configurations defining geometric volumes of revolution about a principal axis are also covered. AEC's Homogeneous Reactor Experiment No. 1 is a preferred embodiment.

  7. Fuel handling system for a nuclear reactor

    DOEpatents

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

    1986-01-01

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  8. Fuel handling system for a nuclear reactor

    DOEpatents

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

    1986-12-02

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  9. Nuclear reactor insulation and preheat system

    DOEpatents

    Wampole, Nevin C.

    1978-01-01

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

  10. The Liquid Annular Reactor System (LARS) propulsion

    NASA Technical Reports Server (NTRS)

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

    1990-01-01

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

  11. Tomotherapy and other innovative IMRT delivery systems.

    PubMed

    Fenwick, John D; Tomé, Wolfgang A; Soisson, Emilie T; Mehta, Minesh P; Rock Mackie, T

    2006-10-01

    Fixed-field treatments, delivered using conventional clinical linear accelerators fitted with multileaf collimators, have rapidly become the standard form of intensity-modulated radiotherapy (IMRT). Several innovative nonstandard alternatives also exist, for which delivery and treatment planning systems are now commercially available. Three of these nonstandard IMRT approaches are reviewed here: tomotherapy, robotic linear accelerators (CyberKnife, Accuray Inc., Sunnyvale, CA), and standard linear accelerators modulated by jaws alone or by their jaws acting together with a tertiary beam-masking device. Rationales for the nonstandard IMRT approaches are discussed, and elements of their delivery system designs are briefly described. Differences between fixed-field IMRT dose distributions and the distributions that can be delivered by using the nonstandard technologies are outlined. Because conventional linear accelerators are finely honed machines, innovative design enhancement of one aspect of system performance often limits another facet of machine capability. Consequently the various delivery systems may prove optimal for different types of treatment, with specific machine designs excelling for disease sites with specific target volume and normal structure topologies. However it is likely that the delivery systems will be distinguished not just by the optimality of the dose distributions they deliver, but also by factors such as the efficiency of their treatment process, the integration of their onboard imaging systems into that process, and their ability to measure and minimize or compensate for target movement, including the effects of respiratory motion.

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

    NASA Astrophysics Data System (ADS)

    Harto, Andang Widi

    2012-06-01

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

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

    SciTech Connect

    Harto, Andang Widi

    2012-06-06

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

  14. Reactor coolant pump monitoring and diagnostic system

    SciTech Connect

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

    1990-01-01

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

  15. Innovative insulation systems for superconducting fusion magnets

    NASA Astrophysics Data System (ADS)

    Humer, K.; Bittner-Rohrhofer, K.; Fillunger, H.; Maix, R. K.; Prokopec, R.; Weber, H. W.

    2006-03-01

    Glass fibre reinforced plastics (GFRPs) are usually employed as insulating materials for the superconducting coils of large fusion magnets, e.g. of the International Thermonuclear Experimental Reactor (ITER). Both the radiation spectrum and the stresses at the magnet location significantly influence the mechanical behaviour of the magnet insulation and, therefore, impose high demands on the material performance. During the last few decades, advanced epoxy based GFRPs with improved mechanical properties and radiation hardness were introduced into fusion technology. More recently, cyanate ester (CE) matrix systems have become of special interest. In this paper, various magnet insulation systems containing boron-free R-glass fibre reinforcements in commercial and new epoxies as well as in pure CE and CE/epoxy blended matrix systems are presented. All systems were irradiated in a fission reactor at ambient temperature (~340 K) to a fast neutron fluence of 1 × 1022 m-2 (E>0.1 MeV). The mechanical properties were assessed at 77 K in tension as well as in interlaminar shear prior to and after irradiation under static and dynamic conditions.

  16. Colliding Beam Fusion Reactor Space Propulsion System

    NASA Astrophysics Data System (ADS)

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

    2004-02-01

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

  17. 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.

  18. Weld monitor and failure detector for nuclear reactor system

    DOEpatents

    Sutton, Jr., Harry G.

    1987-01-01

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

  19. Structural Conditions for Collaboration and Learning in Innovation Networks: Using an Innovation System Performance Lens to Analyse Agricultural Knowledge Systems

    ERIC Educational Resources Information Center

    Hermans, Frans; Klerkx, Laurens; Roep, Dirk

    2015-01-01

    Purpose: We investigate how the structural conditions of eight different European agricultural innovation systems can facilitate or hinder collaboration and social learning in multidisciplinary innovation networks. Methodology: We have adapted the Innovation System Failure Matrix to investigate the main barriers and enablers eight countries…

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

    DOEpatents

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

    1991-01-01

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

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

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1990-01-01

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

  2. [Innovation and the next generation radiotherapy system].

    PubMed

    Tanabe, Eiji

    2013-01-01

    Innovation is the key to future success for Japan that is slowly falling behind. Industries targeted by the "Abenomics" growth strategy include healthcare and medicine. Since cancer is the leading cause of death in Japan, the development of a system that can detect and treat early stage cancers will be very valuable for patient QOL and reducing health care costs. Although the effectiveness of radiation therapy for treating early stage cancer is widely recognized, there has been no system to treat small, moving tumors with sub millimeter accuracy. A project supported by NEDO develops a "Next-Generation Radiation Therapy System" that uses high energy, narrow X-rays beams that can be accurately pinpointed deep inside the body. Performance testing of a prototype system is currently underway at the National Center for Global Health and Medicine in Tokyo. PMID:24893448

  3. Lessons learned from the tokamak Advanced Reactor Innovation and Evaluation Study (ARIES)

    SciTech Connect

    Krakowski, R.A.; Bathke, C.G.; Miller, R.L.; Werley, K.A.

    1994-07-01

    Lessons from the four-year ARIES (Advanced Reactor Innovation and Evaluation Study) investigation of a number of commercial magnetic-fusion-energy (MFE) power-plant embodiments of the tokamak are summarized. These lessons apply to physics, engineering and technology, and environmental, safety, and health (ES&H) characteristics of projected tokamak power plants. Summarized herein are the composite conclusions and lessons developed in the course of four conceptual tokamak power-plant designs. A general conclusion from this extensive investigation of the commercial potential of tokamak power plants is the need for combined, symbiotic advances in both physics, engineering, and materials before economic competitiveness with developing advanced energy sources can be realized. Advances in materials are also needed for the exploitation of environmental advantages otherwise inherent in fusion power.

  4. Integrated systems analysis of the PIUS reactor

    SciTech Connect

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

    1993-11-01

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

  5. Nuclear reactor fuel rod attachment system

    DOEpatents

    Not Available

    1980-09-17

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

  6. A UASB reactor coupled to a hybrid aerobic MBR as innovative plant configuration to enhance the removal of organic micropollutants.

    PubMed

    Alvarino, T; Suárez, S; Garrido, M; Lema, J M; Omil, F

    2016-02-01

    An innovative plant configuration based in an UASB reactor coupled to a hybrid aerobic membrane bioreactor designed for sustainable treatment of municipal wastewater at ambient temperatures and low hydraulic retention time was studied in terms of organic micropollutants (OMPs) removal. OMPs removal mechanisms, as well as the potential influence of biomass activity and physical conformation were assessed. Throughout all periods of operation (150 days) high organic matter removals were maintained (>95%) and, regarding OMPs removal, this innovative system has shown to be more efficient than conventional technologies for those OMPs which are prone to be biotransformed under anaerobic conditions. For instance, sulfamethoxazole and trimethoprim have both shown to be biodegradable under anaerobic conditions with similar efficiencies (removal efficiencies above 84%). OMPs main removal mechanism was found to be biotransformation, except in the case of musk fragrances which showed medium sorption onto sludge. OMPs removal was strongly dependent on the efficiency of the primary metabolism (organic matter degradation and nitrification) and the type of biomass. PMID:26386770

  7. Structural materials challenges for advanced reactor systems

    NASA Astrophysics Data System (ADS)

    Yvon, P.; Carré, F.

    2009-03-01

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

  8. Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model''

    SciTech Connect

    Dmitriy Y. Anistratov; Marvin L. Adams; Todd S. Palmer; Kord S. Smith; Kevin Clarno; Hikaru Hiruta; Razvan Nes

    2003-08-04

    OAK (B204) Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model'' The present generation of reactor analysis methods uses few-group nodal diffusion approximations to calculate full-core eigenvalues and power distributions. The cross sections, diffusion coefficients, and discontinuity factors (collectively called ''group constants'') in the nodal diffusion equations are parameterized as functions of many variables, ranging from the obvious (temperature, boron concentration, etc.) to the more obscure (spectral index, moderator temperature history, etc.). These group constants, and their variations as functions of the many variables, are calculated by assembly-level transport codes. The current methodology has two main weaknesses that this project addressed. The first weakness is the diffusion approximation in the full-core calculation; this can be significantly inaccurate at interfaces between different assemblies. This project used the nodal diffusion framework to implement nodal quasidiffusion equations, which can capture transport effects to an arbitrary degree of accuracy. The second weakness is in the parameterization of the group constants; current models do not always perform well, especially at interfaces between unlike assemblies. The project developed a theoretical foundation for parameterization and homogenization models and used that theory to devise improved models. The new models were extended to tabulate information that the nodal quasidiffusion equations can use to capture transport effects in full-core calculations.

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

    SciTech Connect

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

    2006-07-01

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

  10. Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

    SciTech Connect

    Simon, N.; Lorcet, H.; Beauchamp, F.; Guigues, E.; Lovera, P.; Fleche, J. L.; Lacroix, M.; Carra, O.; Prele, G.

    2012-07-01

    Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO{sub 2} interaction are also presented. Then, in a second part, a functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)

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

    NASA Astrophysics Data System (ADS)

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

    1987-07-01

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

  12. Nuclear reactor cooling system decontamination reagent regeneration

    DOEpatents

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

    1985-01-01

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

  13. Innovative ventilation system for animal anatomy laboratory

    SciTech Connect

    Lacey, D.R.; Smith, D.C.

    1997-04-01

    A unique ventilation system was designed and built to reduce formaldehyde fumes in the large animal anatomy lab at the Vet Medical Center at Cornell University. The laboratory includes four rooms totaling 5,500 ft{sup 2}. The main room has 2,300 ft{sup 2} and houses the laboratory where up to 60 students dissect as many as 12 horses at a time. Other rooms are a cold storage locker, an animal preparation room and a smaller lab for specialized instruction. The large animal anatomy laboratory has a history of air quality complaints despite a fairly high ventilation rate of over 10 air changes/hour. The horses are embalmed, creating a voluminous source of formaldehyde and phenol vapors. Budget constraints and increasingly stringent exposure limits for formaldehyde presented a great challenge to design a ventilation system that yields acceptable air quality. The design solution included two innovative elements: air-to-air heat recovery, and focused ventilation.

  14. Proceedings of a Symposium on Advanced Compact Reactor Systems

    NASA Technical Reports Server (NTRS)

    1983-01-01

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

  15. Systems aspects of a space nuclear reactor power system

    NASA Technical Reports Server (NTRS)

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

    1988-01-01

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

  16. Enhancing the Reflexivity of System Innovation Projects with System Analyses

    ERIC Educational Resources Information Center

    van Mierlo, Barbara; Arkesteijn, Marlen; Leeuwis, Cees

    2010-01-01

    Networks aiming for fundamental changes bring together a variety of actors who are part and parcel of a problematic context. These system innovation projects need to be accompanied by a monitoring and evaluation approach that supports and maintains reflexivity to be able to deal with uncertainties and conflicts while challenging current practices…

  17. Open Innovation, Triple Helix and Regional Innovation Systems: Exploring CATAPULT Centres in the UK

    ERIC Educational Resources Information Center

    Kerry, Christopher; Danson, Michael

    2016-01-01

    Through the lens of UK CATAPULT Centres this conceptual paper presents an examination of the links between open innovation, the Triple Helix model and regional innovation systems. Highlighting the importance of boundary-spanning intermediaries, the combined role of these concepts is explored in detail. A conceptual model is then proposed which…

  18. Pipe Explorer{trademark} surveying system. Innovative technology summary report

    SciTech Connect

    Not Available

    1999-06-01

    The US Department of Energy`s (DOE) Chicago Operations Office and the DOE`s Federal Energy Technology Center (FETC) developed a Large Scale Demonstration Project (LSDP) at the Chicago Pile-5 Research Reactor (CP-5) at Argonne National Laboratory-East (ANL). The objective of the LSDP is to demonstrate potentially beneficial decontamination and decommissioning (D and D) technologies in comparison with current baseline technologies. The Pipe Explorer{trademark} system was developed by Science and Engineering Associates, Inc. (SEA), Albuquerque, NM as a deployment method for transporting a variety of survey tools into pipes and ducts. Tools available for use with the system include alpha, beta and gamma radiation detectors; video cameras; and pipe locator beacons. Different versions of this technology have been demonstrated at three other sites; results of these demonstrations are provided in an earlier Innovative Technology Summary Report. As part of a D and D project, characterization radiological contamination inside piping systems is necessary before pipes can be recycled, remediated or disposed. This is usually done manually by surveying over the outside of the piping only, with limited effectiveness and risk of worker exposure. The pipe must be accessible to workers, and embedded pipes in concrete or in the ground would have to be excavated at high cost and risk of exposure to workers. The advantage of the Pipe Explorer is its ability to perform in-situ characterization of pipe internals.

  19. Fluid sampling system for a nuclear reactor

    DOEpatents

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

    1994-11-22

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

  20. Fluid sampling system for a nuclear reactor

    DOEpatents

    Lau, Louis K.; Alper, Naum I.

    1994-01-01

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

  1. Nuclear reactor fuel rod attachment system

    DOEpatents

    Christiansen, David W.

    1982-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2006-01-01

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

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

    SciTech Connect

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

    2006-01-20

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

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

    NASA Astrophysics Data System (ADS)

    Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim

    2014-05-01

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

  5. Liquid metal cooled nuclear reactors with passive cooling system

    DOEpatents

    Hunsbedt, Anstein; Fanning, Alan W.

    1991-01-01

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

  6. Exhaust system with emissions storage device and plasma reactor

    DOEpatents

    Hoard, John W.

    1998-01-01

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

  7. Cogeneration of Electricity and Potable Water Using The International Reactor Innovative And Secure (IRIS) Design

    SciTech Connect

    Ingersoll, D.T.; Binder, J.L.; Kostin, V.I.; Panov, Y.K.; Polunichev, V.; Ricotti, M.E.; Conti, D.; Alonso, G.

    2004-10-06

    The worldwide demand for potable water has been steadily growing and is projected to accelerate, driven by a continued population growth and industrialization of emerging countries. This growth is reflected in a recent market survey by the World Resources Institute, which shows a doubling in the installed capacity of seawater desalination plants every ten years. The production of desalinated water is energy intensive, requiring approximately 3-6 kWh/m3 of produced desalted water. At current U.S. water use rates, a dedicated 1000 MW power plant for every one million people would be required to meet our water needs with desalted water. Nuclear energy plants are attractive for large scale desalination application. The thermal energy produced in a nuclear plant can provide both electricity and desalted water without the production of greenhouse gases. A particularly attractive option for nuclear desalination is to couple a desalination plant with an advanced, modular, passively safe reactor design. The use of small-to-medium sized nuclear power plants allows for countries with smaller electrical grid needs and infrastructure to add new electrical and water capacity in more appropriate increments and allows countries to consider siting plants at a broader number of distributed locations. To meet these needs, a modified version of the International Reactor Innovative and Secure (IRIS) nuclear power plant design has been developed for the cogeneration of electricity and desalted water. The modular, passively safe features of IRIS make it especially well adapted for this application. Furthermore, several design features of the IRIS reactor will ensure a safe and reliable source of energy and water even for countries with limited nuclear power experience and infrastructure. The IRIS-D design utilizes low-quality steam extracted from the low-pressure turbine to boil seawater in a multi-effect distillation desalination plant. The desalination plant is based on the horizontal

  8. Fuelling a National Innovation System in Colombia

    ERIC Educational Resources Information Center

    Lucio-Arias, Diana

    2006-01-01

    This presentation of the innovation-driven environment in Colombia derives from important national efforts to gather and store pertinent information. Two large surveys have tested the "innovative behaviour" of Colombian manufacturing firms--the more recent of these was in 2005. Another information source is the Scienti platform, an online effort…

  9. Schooling Redesigned: Towards Innovative Learning Systems

    ERIC Educational Resources Information Center

    Istance, David

    2015-01-01

    What does redesigning schools and schooling through innovation mean in practice? How might it be brought about? These questions have inspired an influential international reflection on "Innovative Learning Environments" (ILE) led by the OECD. This reflection has already resulted in publications on core design principles and frameworks…

  10. 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.-

  11. Thermionic switched self-actuating reactor shutdown system

    DOEpatents

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

    1989-01-01

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

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

    SciTech Connect

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

    2009-03-01

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

  13. Innovative self-powered submersible microbial electrolysis cell (SMEC) for biohydrogen production from anaerobic reactors.

    PubMed

    Zhang, Yifeng; Angelidaki, Irini

    2012-05-15

    A self-powered submersible microbial electrolysis cell (SMEC), in which a specially designed anode chamber and external electricity supply were not needed, was developed for in situ biohydrogen production from anaerobic reactors. In batch experiments, the hydrogen production rate reached 17.8 mL/L/d at the initial acetate concentration of 410 mg/L (5 mM), while the cathodic hydrogen recovery ( [Formula: see text] ) and overall systemic coulombic efficiency (CE(os)) were 93% and 28%, respectively, and the systemic hydrogen yield ( [Formula: see text] ) peaked at 1.27 mol-H(2)/mol-acetate. The hydrogen production increased along with acetate and buffer concentration. The highest hydrogen production rate of 32.2 mL/L/d and [Formula: see text] of 1.43 mol-H(2)/mol-acetate were achieved at 1640 mg/L (20 mM) acetate and 100 mM phosphate buffer. Further evaluation of the reactor under single electricity-generating or hydrogen-producing mode indicated that further improvement of voltage output and reduction of electron losses were essential for efficient hydrogen generation. In addition, alternate exchanging the electricity-assisting and hydrogen-producing function between the two cell units of the SMEC was found to be an effective approach to inhibit methanogens. Furthermore, 16S rRNA genes analysis showed that this special operation strategy resulted same microbial community structures in the anodic biofilms of the two cell units. The simple, compact and in situ applicable SMEC offers new opportunities for reactor design for a microbial electricity-assisted biohydrogen production system.

  14. Innovative self-powered submersible microbial electrolysis cell (SMEC) for biohydrogen production from anaerobic reactors.

    PubMed

    Zhang, Yifeng; Angelidaki, Irini

    2012-05-15

    A self-powered submersible microbial electrolysis cell (SMEC), in which a specially designed anode chamber and external electricity supply were not needed, was developed for in situ biohydrogen production from anaerobic reactors. In batch experiments, the hydrogen production rate reached 17.8 mL/L/d at the initial acetate concentration of 410 mg/L (5 mM), while the cathodic hydrogen recovery ( [Formula: see text] ) and overall systemic coulombic efficiency (CE(os)) were 93% and 28%, respectively, and the systemic hydrogen yield ( [Formula: see text] ) peaked at 1.27 mol-H(2)/mol-acetate. The hydrogen production increased along with acetate and buffer concentration. The highest hydrogen production rate of 32.2 mL/L/d and [Formula: see text] of 1.43 mol-H(2)/mol-acetate were achieved at 1640 mg/L (20 mM) acetate and 100 mM phosphate buffer. Further evaluation of the reactor under single electricity-generating or hydrogen-producing mode indicated that further improvement of voltage output and reduction of electron losses were essential for efficient hydrogen generation. In addition, alternate exchanging the electricity-assisting and hydrogen-producing function between the two cell units of the SMEC was found to be an effective approach to inhibit methanogens. Furthermore, 16S rRNA genes analysis showed that this special operation strategy resulted same microbial community structures in the anodic biofilms of the two cell units. The simple, compact and in situ applicable SMEC offers new opportunities for reactor design for a microbial electricity-assisted biohydrogen production system. PMID:22402271

  15. Computer optimization of reactor-thermoelectric space power systems

    NASA Technical Reports Server (NTRS)

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

    1973-01-01

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

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

    Code of Federal Regulations, 2012 CFR

    2012-01-01

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

  17. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

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

  18. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

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

  19. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

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

  20. INNOVATIVE URBAN WET-WEATHER FLOW MANAGEMENT SYSTEMS

    EPA Science Inventory

    This report describes innovative methods to improve wet weather flow (WWF) management systems, that provide drainage services at the same time as decreasing stormwater pollutant discharges, for urban developments of the 21st century. Traditionally, wet-weather collection systems...

  1. Monitoring circuit for reactor safety systems

    DOEpatents

    Keefe, Donald J.

    1976-01-01

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

  2. National Implications: Closed Systems Stifle Innovation, Collaboration and Flexibility

    ERIC Educational Resources Information Center

    Cloud, Michelle; Kritsonis, William Allan

    2006-01-01

    Educational leaders must work to establish organizational structures that help schools achieve and sustain their vision. The intent of this article is to briefly examine how closed systems stifle innovation, collaboration and flexibility in schooling. Innovation, collaboration and flexibility are key ingredients for creating successful…

  3. Gas-cooled reactor for space power systems

    SciTech Connect

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

    1987-05-01

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

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

    SciTech Connect

    Harty, R.B.

    1983-09-30

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

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

    SciTech Connect

    Baumann, N.P.

    1990-01-01

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

  6. Gas-cooled reactor power systems for space

    NASA Astrophysics Data System (ADS)

    Walter, Carl E.

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

  7. Innovative Approach to Establish Root Causes for Cracking in Aggressive Reactor Environments

    SciTech Connect

    Bruemmer, Stephen M.; Thomas, Larry E.; Vetrano, John S.; Simonen, Edward P.

    2003-10-31

    The research focuses on the high-resolution characterization of degradation microstructures and microchemistries in specimens tested under controlled conditions for the environment and for the material where in-service complexities can be minimized. Thermodynamic and kinetic modeling of crack-tip processes is employed to analyze corrosion-induced structures and gain insights into degradation mechanisms. Novel mechanistic ''fingerprinting'' of crack-tip structures is used to isolate causes of environmental cracking in tandem with quantitative measurements of crack growth. Sample preparation methods and advanced analytical techniques are used to characterize corrosion/oxidation reactions and crack-tip structures at near atomic dimensions in order to gain insight into fundamental environmental cracking mechanisms. Reactions at buried interfaces, not accessible by conventional approaches, are being systematically interrogated. Crack-growth experiments in high-temperature water environments are evaluating and isolating the effects of material condition (matrix strength, grain boundary composition and precipitation) on stress corrosion cracking (SCC). The fundamental understanding of crack advance mechanisms will establish the basis to design new corrosion-resistant alloys for current light-water reactors and advanced reactor systems.

  8. Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems

    NASA Astrophysics Data System (ADS)

    Was, Gary S.

    2007-08-01

    The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems.

  9. Transient thermal analysis of a space reactor power system

    SciTech Connect

    Gaeta, M.J.; Best, F.R. . Dept. of Nuclear Engineering)

    1993-07-01

    Space nuclear power systems utilize materials and processes that are completely different from terrestrial reactor systems. Therefore, the tools used to analyze ground-based systems are inappropriate for space reactor design and analysis. The purpose of this study was to develop a space reactor transient analysis tool and to apply this tool to scenarios of interest. The scope of the simulation includes the thermal and neutronic behavior of a liquid-metal-cooled fast reactor, the electrical and thermal performance of the thermoelectric generators, the thermal dynamics of heat pipe radiators, and the thermal behavior of the coolant piping between major components. The thermal model of the system is explicitly coupled to a momentum model of the primary and secondary coolant loops. A one-dimensional conduction model is employed in all solid component models. The reactor model includes an expression for energy generation due to fission and decay heat. The thermoelectric heat exchanger model accounts for thermal energy conversion to useful electrical output. The two-node radiator heat pipe model includes normal operation as well as limited heat pipe operation under sonic limit conditions. The reactor, thermoelectric heat exchanger, and heat pipe models are coupled explicitly by the coolant piping thermal model. The computer program is used to simulate a variety of transients including reactor power changer, degradation of the radiator, and a temporary open circuit condition on the thermoelectrics.

  10. Robotic system for remote maintenance of a pulsed nuclear reactor

    SciTech Connect

    Thunborg, S.

    1986-01-01

    Guidelines recently established for occupational radiation exposure specify that exposure should be as low as reasonably achievable. In conformance with these guidelines, SNL has developed a remote maintenance robot (RMR) system for use in the Sandia Pulse Reactor III (SPR III) facility. The RMR should reduce occupational radiation exposure by a factor of 4 and decrease reactor downtime. Other goals include developing a technology base for a more advanced pulse reactor and for the nuclear fuel cycle programs of the US Department of Energy and US Nuclear Regulatory Commission. The RMR has five major subsystems: (a) a chain-driven cart to bring the system into the reactor room; (b) a Puma 560 robot to perform dextrous operations; (c) a programmable turntable to orient the robot to any of the reactor's four sides; (d) a programmable overhead hoist for lifting components weighing up to 400 lb onto or off of the reactor; and (e) a supervisory control console for the system operator. Figure 1 is a schematic diagram of the turntable, hoist, and robot system in position around the SPR III reactor.

  11. Emergency heat removal system for a nuclear reactor

    DOEpatents

    Dunckel, Thomas L.

    1976-01-01

    A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

  12. Autonomous Control and Diagnostics of Space Reactor Systems

    SciTech Connect

    Upadhyaya, B.R.; Xu, X.; Perillo, S.R.P.; Na, M.G.

    2006-07-01

    This paper describes three key features of the development of an autonomous control strategy for space reactor systems. These include the development of a reactor simulation model for transient analysis, development of model-predictive control as part of the autonomous control strategy, and a fault detection and isolation module. The latter is interfaced with the control supervisor as part of a hierarchical control system. The approach has been applied to the nodal model of the SP-100 reactor with a thermo-electric generator. The results of application demonstrate the effectiveness of the control approach and its ability to reconfigure the control mode under fault conditions. (authors)

  13. Microprocessor tester for the treat upgrade reactor trip system

    SciTech Connect

    Lenkszus, F.R.; Bucher, R.G.

    1984-01-01

    The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.

  14. SUBSURFACE VOLATIZATION AND VENTILATION SYSTEM (SVVS) - INNOVATIVE TECHNOLOGY REPORT

    EPA Science Inventory

    This report summarizes the findings associated with a Demonstration Test of Environmental Improvement Technologies’ (EIT) Subsurface Volatilization and Ventilation System (SVVS) process. The technology was evaluated under the EPA Superfund Innovative Technology Evaluation (SITE) ...

  15. Proposal of Space Reactor for Nuclear Electric Propulsion System

    NASA Astrophysics Data System (ADS)

    Nagata, Hidetaka; Nishiyama, Takaaki; Nakashima, Hideki

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

  16. Autonomous Control of Space Reactor Systems

    SciTech Connect

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

    2007-11-30

    Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

  17. Reactor Lithium Heat Pipes for HP-STMCs Space Reactor Power System

    NASA Astrophysics Data System (ADS)

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2004-02-01

    Design and performance analysis of the nuclear reactor's lithium heat pipes for a 110-kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The evaporator length of the heat pipes is the same as the active core height (0.45 m) and the C-C finned condenser is of the same length as the STMC panels (1.5 m). The C-C finned condenser section is radiatively coupled to the collector shoes of the STMCs placed on both sides. The lengths of the adiabatic section, the values of the power throughput and the evaporator wall temperature depend on the radial location of the heat pipe in the reactor core and the number and dimensions of the potassium heat pipes in the heat rejection radiator. The reactor heat pipes have a total length that varies from 7.57 to 7.73 m, and a 0.2 mm thick Mo-14%Re wick with an average pore radius of 12 μm. The wick is separated from the Mo-14%Re wall by a 0.5 mm annulus filled with liquid lithium, to raise the prevailing capillary limit. The nominal evaporator (or reactor) temperature varies from 1513 to 1591 K and the thermal power of the reactor is 1.6 MW, which averages 12.7 kW for each of the 126 reactor heat pipes. The power throughput per heat pipe increase to a nominal 15.24 kW at the location of the peak power in the core and to 20.31 kW when an adjacent heat pipe fails. The prevailing capillary limit of the reactor heat pipes is 28.3 kW, providing a design margin >= 28%.

  18. Reactor Lithium Heat Pipes for HP-STMCs Space Reactor Power System

    SciTech Connect

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2004-02-04

    Design and performance analysis of the nuclear reactor's lithium heat pipes for a 110-kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The evaporator length of the heat pipes is the same as the active core height (0.45 m) and the C-C finned condenser is of the same length as the STMC panels (1.5 m). The C-C finned condenser section is radiatively coupled to the collector shoes of the STMCs placed on both sides. The lengths of the adiabatic section, the values of the power throughput and the evaporator wall temperature depend on the radial location of the heat pipe in the reactor core and the number and dimensions of the potassium heat pipes in the heat rejection radiator. The reactor heat pipes have a total length that varies from 7.57 to 7.73 m, and a 0.2 mm thick Mo-14%Re wick with an average pore radius of 12 {mu}m. The wick is separated from the Mo-14%Re wall by a 0.5 mm annulus filled with liquid lithium, to raise the prevailing capillary limit. The nominal evaporator (or reactor) temperature varies from 1513 to 1591 K and the thermal power of the reactor is 1.6 MW, which averages 12.7 kW for each of the 126 reactor heat pipes. The power throughput per heat pipe increase to a nominal 15.24 kW at the location of the peak power in the core and to 20.31 kW when an adjacent heat pipe fails. The prevailing capillary limit of the reactor heat pipes is 28.3 kW, providing a design margin {>=} 28%.

  19. Modeling Reactor Coolant Systems Thermal-Hydraulic Transients

    1999-10-05

    RELAP5/MOD3.2* is used to model reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transients without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal-hydraulic systems. Control system and secondary system components are included to allow modeling of themore » plant controls, turbines, condensers, and secondary feedwater systems.« less

  20. A high energy neutral beam system for reactors

    SciTech Connect

    Anderson, O.A.; Chan, C.F.; Cooper, W.S.; Leung, K.N.; Lietzke, A.F.; Kim, C.H.; Kunkel, W.B.; Kwan, J.W.; Purgalis, P.; Schlachter, A.S.

    1988-09-01

    High energy neutral beams provide a promising method of heating and driving current in steady-state tokamak fusion reactors. As an example, we have made a conceptual design of a neutral beam system for current drive on the International Thermonuclear Experimental Reactor (ITER). The system, based on electrostatic acceleration of D/sup /minus// ions, can deliver up to 100 MW of 1.6 MeV D/sup 0/ neutrals through three ports. Radiation protection is provided by locating sensitive beamline components 35 to 50 m from the reactor. In an application to a 3300 MW power reactor, a system delivering 120 MW of 2-2.4 MeV deuterium beams assisted by 21 MW of lower hybrid wave power drives 25 MA and provides an adequate plasma power gain (Q = 24) for a commercial fusion power plant. 8 refs., 1 fig., 2 tabs.

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

    1999-04-21

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

  2. 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.

  3. Gas-cooled reactor power systems for space

    SciTech Connect

    Walter, C.E.

    1987-01-01

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

  4. Reference Reactor Module for the Affordable Fission Surface Power System

    NASA Astrophysics Data System (ADS)

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

    2008-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO2-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important ``affordability'' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  5. Reference Reactor Module for the Affordable Fission Surface Power System

    SciTech Connect

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

    2008-01-21

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO{sub 2}-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important 'affordability' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  6. Passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  7. Natural circulating passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  8. CONCEPTUAL DESIGN OF A LUNAR REGOLITH CLUSTERED-REACTOR SYSTEM

    SciTech Connect

    John Darrell Bess

    2009-06-01

    It is proposed that a fast-fission, heatpipe-cooled, lunar-surface power reactor system be divided into subcritical units that could be launched safely without the incorporation of additional spectral shift absorbers or other complex means of control. The reactor subunits are to be emplaced directly into the lunar regolith utilizing the regolith not just for shielding but as the reflector material to increase the neutron economy of the system. While a single subunit cannot achieve criticality by itself, coordinated placement of additional subunits will provide a critical reactor system for lunar surface power generation. A lunar regolith clustered-reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of a slight increase in launch mass per rated power level and an overall reduction in neutron economy when compared to a single-reactor system. Additional subunits may be launched with future missions to increase the cluster size and power according to desired lunar base power demand and lifetime. The results address the potential uncertainties associated with the lunar regolith material and emplacement of the subunit systems. Physical distance between subunits within the clustered emplacement exhibits the most significant feedback regarding changes in overall system reactivity. Narrow, deep holes will be the most effective in reducing axial neutron leakage from the core. The variation in iron concentration in the lunar regolith can directly influence the overall system reactivity although its effects are less than the more dominant factors of subunit emplacement.

  9. New reactor technology: safety improvements in nuclear power systems.

    PubMed

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

  10. New reactor technology: safety improvements in nuclear power systems.

    PubMed

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems. PMID:18049233

  11. Deployment history and design considerations for space reactor power systems

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.

    2009-05-01

    The history of the deployment of nuclear reactors in Earth orbits is reviewed with emphases on lessons learned and the operation and safety experiences. The former Soviet Union's "BUK" power systems, with SiGe thermoelectric conversion and fast neutron energy spectrum reactors, powered a total of 31 Radar Ocean Reconnaissance Satellites (RORSATs) from 1970 to 1988 in 260 km orbit. Two of the former Soviet Union's TOPAZ reactors, with in-core thermionic conversion and epithermal neutron energy spectrum, powered two Cosmos missions launched in 1987 in ˜800 km orbit. The US' SNAP-10A system, with SiGe energy conversion and a thermal neutron energy spectrum reactor, was launched in 1965 in 1300 km orbit. The three reactor systems used liquid NaK-78 coolant, stainless steel structure and highly enriched uranium fuel (90-96 wt%) and operated at a reactor exit temperature of 833-973 K. The BUK reactors used U-Mo fuel rods, TOPAZ used UO 2 fuel rods and four ZrH moderator disks, and the SNAP-10A used moderated U-ZrH fuel rods. These low power space reactor systems were designed for short missions (˜0.5 kW e and ˜1 year for SNAP-10A, <3.0 kW e and <6 months for BUK, and ˜5.5 kW e and up to 1 year for TOPAZ). The deactivated BUK reactors at the end of mission, which varied in duration from a few hours to ˜4.5 months, were boosted into ˜800 km storage orbit with a decay life of more than 600 year. The ejection of the last 16 BUK reactor fuel cores caused significant contamination of Earth orbits with NaK droplets that varied in sizes from a few microns to 5 cm. Power systems to enhance or enable future interplanetary exploration, in-situ resources utilization on Mars and the Moon, and civilian missions in 1000-3000 km orbits would generate significantly more power of 10's to 100's kW e for 5-10 years, or even longer. A number of design options to enhance the operation reliability and safety of these high power space reactor power systems are presented and discussed.

  12. Dual annular rotating "windowed" nuclear reflector reactor control system

    DOEpatents

    Jacox, Michael G.; Drexler, Robert L.; Hunt, Robert N. M.; Lake, James A.

    1994-01-01

    A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

  13. Innovators.

    ERIC Educational Resources Information Center

    NEA Today, 2001

    2001-01-01

    Describes various innovations that have been developed to enhance education. These innovations include: helping educators help at-risk students succeed; promoting high school journalism; ensuring quality online learning experiences; developing a student performing group that uses theater to address social issues; and having students design their…

  14. BIODEGRADATION OF METHYL TERT-BUTYL ETHER USING AN INNOVATIVE BIOMASS CONCENTRATOR REACTOR

    EPA Science Inventory

    The aerobic biodegradation of methyl tert-butyl ether (MTBE) was investigated using a pilot-scale Biomass Concentrator Reactor (BCR). The reactor was operated for a year at a flow rate of 2500 L/d of Cincinnati dechlorinated tap water and an influent MTBE concentration o...

  15. Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

    NASA Astrophysics Data System (ADS)

    Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

    2014-08-01

    The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

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

    PubMed

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

    2013-05-01

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

  17. Small space reactor power systems for unmanned solar system exploration missions

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

  18. Small space reactor power systems for unmanned solar system exploration missions

    SciTech Connect

    Bloomfield, H.S.

    1987-12-01

    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

  19. 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.

  20. Reference reactor module for NASA's lunar surface fission power system

    SciTech Connect

    Poston, David I; Kapernick, Richard J; Dixon, David D; Werner, James; Qualls, Louis; Radel, Ross

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

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

    DOEpatents

    Isaksson, Juhani M.

    1996-01-01

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

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

    DOEpatents

    Isaksson, J.M.

    1996-06-18

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

  3. Gas-cooled reactor power systems for space

    SciTech Connect

    Walter, C.E.

    1987-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

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

  5. System aspects of a Space Nuclear Reactor Power System

    SciTech Connect

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Kia, T.; Nesmith, B.

    1988-01-01

    Selected systems aspects of a 300 kW nuclear reactor power system for spacecraft have been studied. The approach included examination of two candidate missions and their associated spacecraft, and a number of special topics dealing with the power system design and operation. The missions considered were a reusable orbital transfer vehicle and a space-based radar. The special topics included: power system configuration and scaling, launch vehicle integration, operating altitude, orbital storage, start-up, thawing, control, load following, procedures in case of malfunction, restart, thermal and nuclear radiation to other portions of the spacecraft, thermal stresses between subsystems, boom and cable designs, vibration modes, altitude control, reliability, and survivability. Among the findings are that the stowed length of the power system is important to mission design and that orbital storage for months to years may be needed for missions involving orbital assembly. The power system design evolved during the study and has continued to evolve; the current design differs somewhat from that examined in this paper.

  6. NEUTRONIC REACTOR COUNTER METHOD AND SYSTEM

    DOEpatents

    Graham, C.B.; Spiewak, I.

    1960-05-31

    An improved method is given for controlling the rate of fission in circulating-fuel neutronic reactors in which the fuel is a homogeneous liquid containing fissionable material and a neutron moderator. A change in the rate of flssion is effected by preferentially retaining apart from the circulating fuel a variable amount of either fissionable material or moderator, thereby varying the concentration of fissionable material in the fuel. In the case of an aqueous fuel solution a portion of the water may be continuously vaporized from the circulating solution and the amount of condensate, or condensate plus make-up water, returned to the solution is varied to control the fission rate.

  7. AQUABOX 50 AND MARABU PACKED BIOLOGICAL REACTOR SYSTEM TECHNOLOGY EVALUATION, STADTWERKE DUESSELDORF AG SITE, DUESSELDORF, GERMANY

    EPA Science Inventory

    This ITER summarizes the results of an evaluation of the AQUABOX 50 and MARABU Packed Biological Reactor technologies. The evaluation was conducted under a bilateral agreement between the United States (U.S.) Environmental Protection Agency (EPA) Superfund Innovative Technology ...

  8. Different Mechanisms for Establishing Liquid Walls in Advanced Reactor Systems

    NASA Astrophysics Data System (ADS)

    Hançerlioğullari, Aybaba; Cini, Mesut

    2013-04-01

    The APEX study is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around a fusion plasma. In this study the modeling of APEX hybrid reactor produced by using ARIES-RS hybrid reactor technology, was performed by using the Monte Carlo code and ENF/B-V-VI nuclear data. The most important feature of APEX hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity, good power transformation productivity the magnitude of the reactor's operational duration, low failure percentage, short maintenance time and the inclusion of the system's simple technology and material. Around the fusion chamber, molten salt Li2BeF4 and natural lithium were used as cooling materials. The result of the study indicated that fissile material production UF4 and ThF4 heavy metal salt increased nearly at the same percentage.

  9. Reactor Controllability of 3-Region-Core Molten Salt Reactor System - A Study on Load Following Capability

    SciTech Connect

    Takahisa Yamamoto; Koshi Mitachi; Masatoshi Nishio

    2006-07-01

    The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of fissile fuels (breeding). Thorium (Th) and uranium-233 ({sup 233}U) are fertile and fissile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650 K), that is both the reactor nuclear fuel and the coolant. The MSR system is one of the six advanced reactor concepts identified by the Generation IV International Forum (GIF) as a candidate for cooperative development. In the MSR system, fuel salt flows through a fuel duct constructed around a reactor core and fuel channel of a graphite moderator accompanied by fission reaction and heat generation, and flows out to an external-loop system consisted of a heat exchanger and a circulation pump. Due to the motion of fuel salt, delayed neutron precursors that are one of the source of neutron production make to change their position between the fission reaction and neutron emission events and decay even occur in the external loop system. Hence the reactivity and effective delayed neutron precursor fraction of the MSR system are lower than those of solid fuel reactor systems such as Boiling Water Reactors (BWRs) and Pressurised Water Reactor (PWRs). Since all of the presently operating nuclear power reactors utilize solid fuel, little attention had been paid to the MSR analysis of the reactivity loss and reactor characteristics change caused by the fuel salt circulation. Sides et al. and Shimazu et al. developed MSR analytical models based on the point reactor kinetics model to consider the effect of fuel salt flow. Their models represented a reactor as having six zones for fuel salt and three zones for the graphite moderator. Since their models employed the point reactor kinetics model and the rough temperature approximation, their results were not sufficiently accurate to

  10. Reactor technology assessment and selection utilizing systems engineering approach

    SciTech Connect

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-12

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  11. Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

    SciTech Connect

    Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.

    2011-04-06

    Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

  12. Reactor technology assessment and selection utilizing systems engineering approach

    NASA Astrophysics Data System (ADS)

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-01

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  13. In Situ Underwater Gamma Spectroscopy System. Innovative Technology Summary Report

    SciTech Connect

    2001-06-01

    The baseline approach to characterize underwater objects is to use radiation sensors on an extendable pole or the Remote Underwater Characterization System to detect gamma radiation and to collect samples for laboratory analysis to determine concentrations of specific radionuclides. The In-Situ Underwater Gamma Spectroscopy (ISUGS) System is essentially a submersible In-Situ Object Counting System from Canberra that can determine both the total gamma radiation and quantify the specific radionuclides contributing to the radiation. ISUGS was demonstrated to characterize objects in the canal of the Materials Test Reactor in INEEL's Test Reactor Area. Cost analysis based on demonstration data revealed that ISUGS reduced costs by 80% for characterization of ten objects compared to the baseline approach. Frammatome provides ISUGS as part of its characterization services. Based on these promising results, INEEL and other DOE sites plan to use ISUGS to characterize their reactor fuel pools.

  14. Microalgal reactors: a review of enclosed system designs and performances.

    PubMed

    Carvalho, Ana P; Meireles, Luís A; Malcata, F Xavier

    2006-01-01

    One major challenge to industrial microalgal culturing is to devise and develop technical apparata, cultivation procedures and algal strains susceptible of undergoing substantial increases in efficiency of use of solar energy and carbon dioxide. Despite several research efforts developed to date, there is no such thing as "the best reactor system"- defined, in an absolute fashion, as the one able to achieve maximum productivity with minimum operation costs, irrespective of the biological and chemical system at stake. In fact, choice of the most suitable system is situation-dependent, as both the species of alga available and the final purpose intended will play a role. The need of accurate control impairs use of open-system configurations, so current investigation has focused mostly on closed systems. In this review, several types of closed bioreactors described in the technical literature as able to support production of microalgae are comprehensively presented and duly discussed, using transport phenomenon and process engineering methodological approaches. The text is subdivided into subsections on: reactor design, which includes tubular reactors, flat plate reactors and fermenter-type reactors; and processing parameters, which include gaseous transfer, medium mixing and light requirements. PMID:17137294

  15. Politics of innovation in multi-level water governance systems

    NASA Astrophysics Data System (ADS)

    Daniell, Katherine A.; Coombes, Peter J.; White, Ian

    2014-11-01

    Innovations are being proposed in many countries in order to support change towards more sustainable and water secure futures. However, the extent to which they can be implemented is subject to complex politics and powerful coalitions across multi-level governance systems and scales of interest. Exactly how innovation uptake can be best facilitated or blocked in these complex systems is thus a matter of important practical and research interest in water cycle management. From intervention research studies in Australia, China and Bulgaria, this paper seeks to describe and analyse the behind-the-scenes struggles and coalition-building that occurs between water utility providers, private companies, experts, communities and all levels of government in an effort to support or block specific innovations. The research findings suggest that in order to ensure successful passage of the proposed innovations, champions for it are required from at least two administrative levels, including one with innovation implementation capacity, as part of a larger supportive coalition. Higher governance levels can play an important enabling role in facilitating the passage of certain types of innovations that may be in competition with currently entrenched systems of water management. Due to a range of natural biases, experts on certain innovations and disciplines may form part of supporting or blocking coalitions but their evaluations of worth for water system sustainability and security are likely to be subject to competing claims based on different values and expertise, so may not necessarily be of use in resolving questions of "best courses of action". This remains a political values-based decision to be negotiated through the receiving multi-level water governance system.

  16. Nuclear reactor heat transport system component low friction support system

    DOEpatents

    Wade, Elman E.

    1980-01-01

    A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

  17. Software reliability and safety in nuclear reactor protection systems

    SciTech Connect

    Lawrence, J.D.

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  18. Reactor protection system with automatic self-testing and diagnostic

    DOEpatents

    Gaubatz, Donald C.

    1996-01-01

    A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically "identical" values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic.

  19. Reactor protection system with automatic self-testing and diagnostic

    DOEpatents

    Gaubatz, D.C.

    1996-12-17

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ``identical`` values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs.

  20. Small reactor power systems for manned planetary surface bases

    SciTech Connect

    Bloomfield, H.S.

    1987-12-01

    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

  1. Small reactor power systems for manned planetary surface bases

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

  2. Research in Modeling and Simulation for Airspace Systems Innovation

    NASA Technical Reports Server (NTRS)

    Ballin, Mark G.; Kimmel, William M.; Welch, Sharon S.

    2007-01-01

    This viewgraph presentation provides an overview of some of the applied research and simulation methodologies at the NASA Langley Research Center that support aerospace systems innovation. Risk assessment methodologies, complex systems design and analysis methodologies, and aer ospace operations simulations are described. Potential areas for future research and collaboration using interactive and distributed simula tions are also proposed.

  3. Analysis of reactor trips originating in balance of plant systems

    SciTech Connect

    Stetson, F.T.; Gallagher, D.W.; Le, P.T.; Ebert, M.W. )

    1990-09-01

    This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of these events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs.

  4. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

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

  5. 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)

  6. System Study: Reactor Core Isolation Cooling 1998–2012

    SciTech Connect

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  7. System Study: Reactor Core Isolation Cooling 1998-2014

    SciTech Connect

    Schroeder, John Alton

    2015-12-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  8. System Study: Reactor Core Isolation Cooling 1998–2013

    SciTech Connect

    Schroeder, John Alton

    2015-01-31

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  9. Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

    DOEpatents

    Hunsbedt, Anstein; Busboom, Herbert J.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

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

    PubMed

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

    2013-05-01

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

  11. Westinghouse Small Modular Reactor nuclear steam supply system design

    SciTech Connect

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

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

    SciTech Connect

    Grogan, Brandon R; Mihalczo, John T

    2009-01-01

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

  13. Sustainability Analysis of Innovative Transport System

    NASA Astrophysics Data System (ADS)

    Meiere, Ieva; Bazbauers, Gatis

    2011-01-01

    The focus of the research is to develop a new approach to transport solution based on the use of a conveyortype system and to compare the environmental impact of the new system with the existing ones. The new transport system consists of a conveyor driven by an electric motor, with a wind power plant supplying electricity, hydrogen storage and a fuel cell for matching the wind power production with the motor load. The research tasks included the evaluation of the consumption of fossil fuels and the associated environmental impact of existing transport system and a comparison with energy consumption and associated environmental impact of the new system. The energy balance of the conveyor transport system was modelled on an hourly basis by using the EnergyPLAN computer program [1] which allows to analyze a combination of intermittent renewable energy technologies, storage and transport systems. The results show that the existing transport system has greater impact on the environment than the proposed one.

  14. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    SciTech Connect

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

    2005-09-27

    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

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

    NASA Astrophysics Data System (ADS)

    Buden, D.

    1993-08-01

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

  16. ANDES Measurements for Advanced Reactor Systems

    NASA Astrophysics Data System (ADS)

    Plompen, A. J. M.; Hambsch, F.-J.; Kopecky, S.; Nyman, M.; Rouki, C.; Salvador Castiñeira, P.; Schillebeeckx, P.; Belloni, F.; Berthoumieux, E.; Gunsing, F.; Lampoudis, C.; Calviani, M.; Guerrero, C.; Cano-Ott, D.; Gonzalez Romero, E.; Aïche, M.; Jurado, B.; Mathieu, L.; Derckx, X.; Farget, F.; Rodrigues Tajes, C.; Bacquias, A.; Dessagne, Ph.; Kerveno, M.; Borcea, C.; Negret, A.; Colonna, N.; Goncalves, I.; Penttilä, H.; Rinta-Antila, S.; Kolhinen, V. S.; Jokinen, A.

    2014-05-01

    A significant number of new measurements was undertaken by the ANDES “Measurements for advanced reactor systems” initiative. These new measurements include neutron inelastic scattering from 23Na, Mo, Zr, and 238U, neutron capture cross sections of 238U, 241Am, neutron induced fission cross sections of 240Pu, 242Pu, 241Am, 243Am and 245Cm, and measurements that explore the limits of the surrogate technique. The latter study the feasibility of inferring neutron capture cross sections for Cm isotopes, the neutron-induced fission cross section of 238Pu and fission yields and fission probabilities through full Z and A identification in inverse kinematics for isotopes of Pu, Am, Cm and Cf. Finally, four isotopes are studied which are important to improve predictions for delayed neutron precursors and decay heat by total absorption gamma-ray spectrometry (88Br, 94Rb, 95Rb, 137I). The measurements which are performed at state-of-the-art European facilities have the ambition to achieve the lowest possible uncertainty, and to come as close as is reasonably achievable to the target uncertainties established by sensitivity studies. An overview is presented of the activities and achievements, leaving detailed expositions to the various parties contributing to the conference.

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

    PubMed

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

    2011-10-15

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

  18. Development and Assessment of Advanced Reactor Core Protection System

    NASA Astrophysics Data System (ADS)

    in, Wang-Kee; Park, Young-Ho; Baeg, Seung-Yeob

    An advanced core protection system for a pressurized water reactor, Reactor Core Protection System(RCOPS), was developed by adopting a high performance hardware platform and optimal system configuration. The functional algorithms of the core protection system were also improved to enhance the plant availability by reducing unnecessary reactor trips and increasing operational margin. The RCOPS consists of four independent safety channels providing a two-out-of-four trip logic. The reliability analysis using the reliability block diagram method showed the unavailability of the RCOPS to be lower than the conventional system. The failure mode and effects analysis demonstrated that the RCOPS does not lose its intended safety functions for most failures. New algorithms for the RCOPS functional design were implemented in order to avoid unnecessary reactor trips by providing auxiliary pre-trip alarms and signal validation logic for the control rod position. The new algorithms in the RCOPS were verified by comparing the RCOPS calculations with reference results. The new thermal margin algorithm for the RCOPS was expected to increase the operational margin to the limit for Departure from Nucleate Boiling Ratio (DNBR) by approximately 1%.

  19. Deployment of remote dismantlement systems at the CP-5 reactor

    SciTech Connect

    Black, D.B.; Ditch, R.W.; Henley, D.R.; Seifert, L.S.

    1997-06-01

    The Chicago Pile 5 (CP-5) Reactor Facility is currently undergoing decontamination and decommissioning (D&D) at the Argonne National Laboratory (ANL) Illinois site. CP-5 was the principal nuclear reactor used to produce neutrons for scientific research at Argonne from 1954 to 1979. The CP-5 reactor was a heavy-water moderated, enriched uranium-fueled reactor with a graphite reflector. The CP-5 D&D project includes the disassembly and removal of all radioactive components, equipment, and structures associated with the CP-5 facility. The Department of Energy`s Robotics Technology Development Program along with the Federal Energy Technology Center, Morgantown Office, have provided teleoperated, remote systems for use in the dismantlement of the CP-5 reactor structure for tasks requiring remote dismantlement. These systems include the dual-arm work platform, the Rosie mobile D&D vehicle, the swing-reduced crane control system, and a remotely-operated crane control system. The dual-arm work platform is a robotic dismantlement system that includes a pair of Schilling Titan III hydraulic manipulators mounted on a special platform, a hydraulic power unit and an operator console. The Rosie mobile D&D work system developed by RedZone Robotics, Inc. is an electro-hydraulic omni-directional locomotor platform with a heavy manipulator mounted on its deck. The Rosie vehicle moves about the floor around the CP-5 reactor block and is operated from a console in the control room. The swing-reduced crane control system has been installed on the CP-5 polar crane, and allows a load suspended from the crane hook to be moved while reducing the induced swing in the load. A remote control system and a rotating crane hook have also been added to the CP-5 polar crane. This paper discusses the status of these remote systems at CP-5 and the facility changes made to allow for their use in the dismantlement of the reactor structure internals. 4 refs., 3 figs.

  20. Innovations in an Accounting Information Systems Course.

    ERIC Educational Resources Information Center

    Shaoul, Jean

    A new approach to teaching an introductory accounting information systems course is outlined and the potential of this approach for integrating computers into the accounting curriculum at Manchester University (England) is demonstrated. Specifically, the use of a small inventory recording system and database in an accounting information course is…

  1. 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)

  2. 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.

  3. 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

  4. Creating a Framework for Research on Systemic Technology Innovations

    ERIC Educational Resources Information Center

    Fishman, Barry; Marx, Ronald W.; Blumenfeld, Phyllis; Krajcik, Joseph; Soloway, Elliot

    2004-01-01

    This article examines why cognitively oriented technology innovations, designed to foster deep thinking and learning, have not become widespread in K-12 schools. We argue a key reason is that most design-based research does not explicitly address systemic issues of usability, scalability and sustainability. This limitation must be overcome if…

  5. Small Business Innovation Research Award Success Story: Proton Energy Systems

    SciTech Connect

    2011-04-01

    This success story describes Proton Energy Systems, a small business that designs and manufactures proton exchange membrane (PEM) electrolysis sytems to produce hydrogen from water. The U.S. Department of Energy's Fuel Cell Technologies Program has supported much of Proton's technology development through Small Business Innovation Research (SBIR) Awards and other non-SBIR funding.

  6. Innovations and Neutralizations in the Warlpiri Pronominal System.

    ERIC Educational Resources Information Center

    Bavin, E. L.; Shopen, T.

    1987-01-01

    Discusses the progress in the number of innovations and neutralizations in the pronominal system of Warlpiri, an aboriginal language spoken in central Australia. The changes are analyzed by age-group usage, and patterns of the changes are suggested. Part of a sample interview in presented. (Author/LMO)

  7. Gap and impact of LMR (Liquid Metal Reactor) piping systems and reactor components

    SciTech Connect

    Ma, D.C.; Gvildys, J.; Chang, Y.W.

    1987-01-01

    Because of high operation temperature, the LMR (Liquid Metal Reactor) plant is characterized by the thin-walled piping and components. Gaps are often present to allow free thermal expansion during normal plant operation. Under dynamic loadings, such as seismic excitation, if the relative displacement between the components exceeds the gap distance, impacts will occur. Since the components and piping become brittle over their design lifetime, impact is of important concern for it may lead to fractures of components and other serious effects. This paper deals with gap and impact problems in the LMR reactor components and piping systems. Emphasis is on the impacts due to seismic motion. Eight sections are contained in this paper. The gap and impact problems in LMR piping systems are described and a parametric study is performed on the effects of gap-induced support nonlinearity on the dynamics characteristics of the LMR piping systems. Gap and impact problems in the LMR reactor components are identified and their mathematical models are illustrated, and the gap and impact problems in the seismic reactor scram are discussed. The mathematical treatments of various impact models are also described. The uncertainties in the current seismic impact analyses of LMR components and structures are presented. An impact test on a 1/10-scale LMR thermal liner is described. The test results indicated that several clusters of natural modes can be excited by the impact force. The frequency content of the excited modes depends on the duration of the impact force; the shorter the duration, the higher the frequency content.

  8. Modification of the Core Cooling System of TRIGA 2000 Reactor

    NASA Astrophysics Data System (ADS)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  9. Modification of the Core Cooling System of TRIGA 2000 Reactor

    SciTech Connect

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-22

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24 deg. C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  10. Neutron Cross Section Covariances: Recent Workshop and Advanced Reactor Systems

    NASA Astrophysics Data System (ADS)

    Oblozinsky, Pavel

    2008-10-01

    The recent Workshop on Neutron Cross Section Covariances, organized by BNL and attended by more than 50 scientists, responded to demands of many user groups, including advanced reactor systems, for uncertainty and correlation information. These demands can be explained by considerable progress in advanced neutronics simulation that probe covariances and their impact on design and operational margins of nuclear systems. The Workshop addressed evaluation methodology, recent evaluations as well as user's perspective, marking era of revival of covariance development that started some two years ago. We illustrate urgent demand for covariances in the case of advanced reactor systems, including fast actinide burner under GNEP, new generation of power reactors, Gen-IV, and reactors under AFCI. A common feature of many of these systems is presence of large amount of minor actinides and fission products that require improved nuclear data. Advanced simulation codes rely on quality input, to be obtained by adjusting the data library, such as the new ENDF/B-VII.0, by considering integral experiments as currently pursued by GNEP. To this end the nuclear data community is developing covariances for formidable amount of 112 materials (isotopes).

  11. Space-reactor electric systems: subsystem technology assessment

    SciTech Connect

    Anderson, R.V.; Bost, D.; Determan, W.R.

    1983-03-29

    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

  12. Services Systems to Leverage Innovators' Knowledge: The Telecoms Industry Case

    NASA Astrophysics Data System (ADS)

    Bugeaud, Florie; Soulier, Eddie

    Today, telecoms operators have to prove their innovation capacity. They address corporate customers which are involved in collaborative value networks. They implement a process of new services research in order to create adapted solutions. This process is linked to a collaborative and complex practice between diverse innovators. However there is no structured method to leverage this community’s knowledge. Based on the SSME discussions, we suggest enriching the telecoms innovation approaches by the representation and the simulation of "services systems". This paper also proposes the development of a collaborative tool to support this suggestion and the innovators’ practices. Our approach is currently tested on a specific telecom service in the e-health domain which presents an interesting business ecosystem.

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

    NASA Technical Reports Server (NTRS)

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

    1974-01-01

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

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

    DOEpatents

    Kelly, Sean M

    2016-09-27

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

  15. A Gas-Cooled Reactor Surface Power System

    SciTech Connect

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

    1998-11-09

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

  16. A gas-cooled reactor surface power system

    NASA Astrophysics Data System (ADS)

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

    1999-01-01

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

  17. A gas-cooled reactor surface power system

    SciTech Connect

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

    1999-01-01

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

  18. A gas-cooled reactor surface power system

    SciTech Connect

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

    1999-01-22

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

  19. Innovation in Extraterrestrial Service Systems - A Challenge for Service Science

    NASA Technical Reports Server (NTRS)

    Bergner, David

    2010-01-01

    This presentation was prepared at the invitation of Professor Yukio Ohsawa, Department of Systems Innovation, School of Engineering, The University of Tokyo, for delivery at the International Workshop on Innovating Service Systems, sponsored by the Japanese Society of Artificial Intelligence (JSAI) as part of the JSAI Internation Symposium on AI, 2010. It offers several challenges for Service Science and Service Innovation. the goal of the presentation is to stimulate thinking about how service systems viII evolve in the future, as human society advances from its terrestrial base toward a permanent presence in space. First we will consider the complexity of the International Space Station (ISS) as it is today, with particular emphasis of its research facilities, and focus on a current challenge - to maximize the utilization of ISS research facilities for the benefit of society. After briefly reviewing the basic principles of Service Science, we will discuss the potential application of Service Innovation methodology to this challenge. Then we viII consider how game-changing technologies - in particular Synthetic Biology - could accelerate the pace of sociocultural evolution and consequently, the progression of human society into space. We will use this provocative vision to advance thinking about how the emerging field of Service Science, Management, and Engineering (SSME) might help us anticipate and better handle the challenges of this inevitable evolutionary process.

  20. The influenza vaccine innovation system and lessons for PDPs.

    PubMed

    Huzair, Farah

    2012-03-01

    As Product Development Partnerships (PDPs) emerge and evolve in response to the need for vaccines, this paper re-examines the oldest and most successful PDP in the vaccine field; that which year after year, produces and reinvents influenza vaccines. This paper describes the influenza vaccine production and innovation system and reviews some of its most recent major innovations. Innovation in this system is a result of collaborative partnerships between various actors from both the public and private sector. It is argued that the influenza vaccine innovation system is a Product Development Partnership (PDP), be it an unconventional one, with a central coordination role allocated to the WHO rather than a private company or charitable/not for profit entity. The unusual structure of this PDP overcomes some of the organizational issues surrounding vaccine research and production faced by other documented PDPs. These are first, the need to coordinate knowledge flow via an effective knowledge broker. Second, the need to build in-house capacity and fund essential research and elements of production where private partners find involvement too risky or costly. PMID:22327495

  1. The influenza vaccine innovation system and lessons for PDPs.

    PubMed

    Huzair, Farah

    2012-03-01

    As Product Development Partnerships (PDPs) emerge and evolve in response to the need for vaccines, this paper re-examines the oldest and most successful PDP in the vaccine field; that which year after year, produces and reinvents influenza vaccines. This paper describes the influenza vaccine production and innovation system and reviews some of its most recent major innovations. Innovation in this system is a result of collaborative partnerships between various actors from both the public and private sector. It is argued that the influenza vaccine innovation system is a Product Development Partnership (PDP), be it an unconventional one, with a central coordination role allocated to the WHO rather than a private company or charitable/not for profit entity. The unusual structure of this PDP overcomes some of the organizational issues surrounding vaccine research and production faced by other documented PDPs. These are first, the need to coordinate knowledge flow via an effective knowledge broker. Second, the need to build in-house capacity and fund essential research and elements of production where private partners find involvement too risky or costly.

  2. MOLECULAR BONDING SYSTEM - INNOVATIVE TECHNOLOGY EVALUATION REPORT

    EPA Science Inventory

    This document presents an evaluation of the Molecular Bonding System (MBS) and its ability to chemically stabilize three metals-contaminated wstes/soils during a SITe demo. The MBS process treated approximately 500 tons each of soil/Fill, Slag, and Miscellaneous Smelter Waste wit...

  3. Automating large-scale reactor systems

    SciTech Connect

    Kisner, R.A.

    1985-01-01

    This paper conveys a philosophy for developing automated large-scale control systems that behave in an integrated, intelligent, flexible manner. Methods for operating large-scale systems under varying degrees of equipment degradation are discussed, and a design approach that separates the effort into phases is suggested. 5 refs., 1 fig.

  4. A Review of Innovation Systems Framework as a Tool for Gendering Agricultural Innovations: Exploring Gender Learning and System Empowerment

    ERIC Educational Resources Information Center

    Kingiri, Ann N.

    2013-01-01

    Purpose: To reflect on the opportunities that a systems understanding of innovation provides for addressing gender issues relevant to women, and to provide some insight on how these might be tackled. Approach: Review of literature relating to gender issues and how they relate to achieving, on the one hand, equity and efficiency goals, and on the…

  5. Operation of staged membrane oxidation reactor systems

    DOEpatents

    Repasky, John Michael

    2012-10-16

    A method of operating a multi-stage ion transport membrane oxidation system. The method comprises providing a multi-stage ion transport membrane oxidation system with at least a first membrane oxidation stage and a second membrane oxidation stage, operating the ion transport membrane oxidation system at operating conditions including a characteristic temperature of the first membrane oxidation stage and a characteristic temperature of the second membrane oxidation stage; and controlling the production capacity and/or the product quality by changing the characteristic temperature of the first membrane oxidation stage and/or changing the characteristic temperature of the second membrane oxidation stage.

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

    SciTech Connect

    Conklin, J.C. )

    1990-04-01

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

  7. Expert system for online surveillance of nuclear reactor coolant pumps

    DOEpatents

    Gross, Kenny C.; Singer, Ralph M.; Humenik, Keith E.

    1993-01-01

    An expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

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

    SciTech Connect

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

    2005-07-15

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

  9. Teams as innovative systems: multilevel motivational antecedents of innovation in R&D teams.

    PubMed

    Chen, Gilad; Farh, Jiing-Lih; Campbell-Bush, Elizabeth M; Wu, Zhiming; Wu, Xin

    2013-11-01

    Integrating theories of proactive motivation, team innovation climate, and motivation in teams, we developed and tested a multilevel model of motivators of innovative performance in teams. Analyses of multisource data from 428 members of 95 research and development (R&D) teams across 33 Chinese firms indicated that team-level support for innovation climate captured motivational mechanisms that mediated between transformational leadership and team innovative performance, whereas members' motivational states (role-breadth self-efficacy and intrinsic motivation) mediated between proactive personality and individual innovative performance. Furthermore, individual motivational states and team support for innovation climate uniquely promoted individual innovative performance, and, in turn, individual innovative performance linked team support for innovation climate to team innovative performance.

  10. Innovative technology summary report: Transportable vitrification system

    SciTech Connect

    1998-09-01

    At the end of the cold war, many of the Department of Energy`s (DOE`s) major nuclear weapons facilities refocused their efforts on finding technically sound, economic, regulatory compliant, and stakeholder acceptable treatment solutions for the legacy of mixed wastes they had produced. In particular, an advanced stabilization process that could effectively treat the large volumes of settling pond and treatment sludges was needed. Based on this need, DOE and its contractors initiated in 1993 the EM-50 sponsored development effort required to produce a deployable mixed waste vitrification system. As a consequence, the Transportable Vitrification System (TVS) effort was undertaken with the primary requirement to develop and demonstrate the technology and associated facility to effectively vitrify, for compliant disposal, the applicable mixed waste sludges and solids across the various DOE complex sites. After 4 years of development testing with both crucible and pilot-scale melters, the TVS facility was constructed by Envitco, evaluated and demonstrated with surrogates, and then successfully transported to the ORNL ETTP site and demonstrated with actual mixed wastes in the fall of 1997. This paper describes the technology, its performance, the technology applicability and alternatives, cost, regulatory and policy issues, and lessons learned.

  11. Nuclear design of the burst power ultrahigh temperature UF4 vapor core reactor system

    NASA Astrophysics Data System (ADS)

    Kahook, Samer D.; Dugan, Edward T.

    1991-01-01

    Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF4 Ultrahigh Temperature Vapor Core Reactor (UTVR)/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MWe for thousands of seconds). The fuel/working fluid is a mixture of UF4 and metal fluoride. Preliminary calculations indicate high overall system efficiencies (≊20%), small radiator size (≊5 m2/MWe), and high specific power (≊5 kWe/kg). Neutronic analysis has revealed a number of attractive features for this novel reactor concept. These include some unique and very effective inherent negative reactivity control mechanisms such as the vapor-fuel density power coefficient of reactivity, the direct neutronic coupling among the multiple fissioning core regions (the central vapor core and the surrounding boiler columns), and the mass flow coupling feedback between the fissioning cores.

  12. Innovative camera system developed for Sprint vehicle

    SciTech Connect

    Not Available

    1985-04-01

    A new inspection system for the Sprint 101 ROV eliminates parallax errors because all three camera modules use a single lens for viewing. Parallax is the apparent displacement of an object when it is viewed from two points not in the same line of sight. The central camera is a Pentax 35-mm single lens reflex with a 28-mm lens. It comes with 250-shot film cassettes, an automatic film wind-on, and a data chamber display. An optical transfer assembly on the stills camera viewfinder transmits the image to one of the two video camera modules. The video picture transmitted to the surface is exactly the same as the stills photo. The surface operator can adjust the focus by viewing the video display.

  13. The radon monitoring system in Daya Bay Reactor Neutrino Experiment

    NASA Astrophysics Data System (ADS)

    Chu, M. C.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Leung, J. K. C.; Leung, K. Y.; Lin, Y. C.; Luk, K. B.; Pun, C. S. J.

    2016-02-01

    We developed a highly sensitive, reliable and portable automatic system (H3) to monitor the radon concentration of the underground experimental halls of the Daya Bay Reactor Neutrino Experiment. H3 is able to measure radon concentration with a statistical error less than 10% in a 1-h measurement of dehumidified air (R.H. 5% at 25 °C) with radon concentration as low as 50 Bq/m3. This is achieved by using a large radon progeny collection chamber, semiconductor α-particle detector with high energy resolution, improved electronics and software. The integrated radon monitoring system is highly customizable to operate in different run modes at scheduled times and can be controlled remotely to sample radon in ambient air or in water from the water pools where the antineutrino detectors are being housed. The radon monitoring system has been running in the three experimental halls of the Daya Bay Reactor Neutrino Experiment since November 2013.

  14. TFE fast driver reactor system for low-power applications

    NASA Astrophysics Data System (ADS)

    Van Hagan, Thomas H.; Lewis, Bryan R.; Bellis, Elizabeth A.; Fisher, Mike V.

    1992-01-01

    This paper addresses reactor design considerations for an in-core thermionic system concept proposed for emerging space power applications with electric power requirements in the 10 to 50-kWe range. At this power level an in-core thermionic rector core requires a combination of thermionic fuel elements (TFEs) and driver fuel elements to achieve nuclear criticality. A pumped liquid-metal loop cools the reactor, transporting the reject heat to a heat pipe radiator. The system concept is a straightforward derivative of the thermionic 100-kWe system designed during the SP-100 Phase 1 program. Combining existing thermionic technology with LMFBR fuel technology and pumped-loop waste heat removal defines a concept that has the advantages of reliability, scalability, technology maturity, and design flexibility.

  15. Kalman filter application for distributed parameter estimation in reactor systems

    SciTech Connect

    Martin, R.P.; Edwards, R.M.

    1996-07-01

    An application of the Kalman filter has been developed for the real-time identification of a distributed parameter in a nuclear power plant. This technique can be used to improve numerical method-based best-estimate simulation of complex systems such as nuclear power plants. The application to a reactor system involves a unique modal model that approximates physical components, such as the reactor, as a coupled oscillator, i.e., a modal model with coupled modes. In this model both states and parameters are described by an orthogonal expansion. The Kalman filter with the sequential least-squares parameter estimation algorithm was used to estimate the modal coefficients of all states and one parameter. Results show that this state feedback algorithm is an effective way to parametrically identify a distributed parameter system in the presence of uncertainties.

  16. A passive automated personnel accountability system for reactor emergency preparedness

    SciTech Connect

    Zimmerman, R.O.; DeLisle, G.V.; Hickey, E.E.

    1988-04-01

    In 1985 a project was undertaken at the N Reactor on the Hanford Site to develop an automated personnel accountability system to ensure accountability of all personnel within 30 minutes of a site evacuation. The decision to develop such a system was made after a full-scale evacuation drill showed that the manual accountability system in use at the time was inadequate to meet the 30-minute requirement. Accountability systems at commercial nuclear power plants were evaluated, but found to be unsuitable because they were not passive, that is, they required action on part of the user for the system to work. Approximately 2500 people could be required to evacuate the 100-N Area. Therefore, a card key system or badge exchange system was judged not to be feasible. A passive accountability system was desired for N Reactor to allow personnel to enter and leave the site in a more timely manner. To meet the need for an automated accountability system at N Reactor, a special Evacuation Accountability System (EVACS) was designed and developed. The EVACS system has three basic components: the transponder, a credit card-sized device worn with the security badge; portal monitors, which are electronically activated by the transponder; and a computer information system that contains the personnel data base. Each person wearing a transponder is accounted for automatically by walking through a portal. In this paper, a description of the hardware and software will be presented, together with problems encountered and lessons learned while adapting an existing technology to this particular use. The system is currently installed and requires acceptance testing before becoming operational.

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

    SciTech Connect

    Buden, D.

    1993-08-11

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

  18. Sulfide-oxidizing bacteria establishment in an innovative microaerobic reactor with an internal silicone membrane for sulfur recovery from wastewater.

    PubMed

    Valdés, F; Camiloti, P R; Rodriguez, R P; Delforno, T P; Carrillo-Reyes, J; Zaiat, M; Jeison, D

    2016-06-01

    A novel bioreactor, employing a silicone membrane for microaeration, was studied for partial sulfide oxidation to elemental sulfur. The objective of this study was to assess the feasibility of using an internal silicone membrane reactor (ISMR) to treat dissolved sulfide and to characterize its microbial community. The ISMR is an effective system to eliminate sulfide produced in anaerobic reactors. Sulfide removal efficiencies reached 96 % in a combined anaerobic/microaerobic reactor and significant sulfate production did not occur. The oxygen transfer was strongly influenced by air pressure and flow. Pyrosequencing analysis indicated various sulfide-oxidizing bacteria (SOB) affiliated to the species Acidithiobacillus thiooxidans, Sulfuricurvum kujiense and Pseudomonas stutzeri attached to the membrane and also indicated similarity between the biomass deposited on the membrane wall and the biomass drawn from the material support, supported the establishment of SOB in an anaerobic sludge under microaerobic conditions. Furthermore, these results showed that the reactor configuration can develop SOB under microaerobic conditions and can improve and reestablish the sulfide conversion to elemental sulfur. PMID:27003697

  19. 75 FR 70741 - Innovative Energy Systems, LLC; Supplemental Notice That Initial Market-Based Rate Filing...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-18

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Innovative Energy Systems, LLC; Supplemental Notice That Initial Market... supplemental notice in the above-referenced proceeding Innovative Energy Systems, LLC's application for...

  20. Systems effects on family planning innovativeness.

    PubMed

    Lee, S B

    1983-12-01

    Data from Korea were used to explore the importance of community level variables in explaining family planning adoption at the individual level. An open system concept was applied, assuming that individual family planning behavior is influenced by both environmental and individual factors. The environmental factors were measured at the village level and designated as community characteristics. The dimension of communication network variables was introduced. Each individual was characterized in terms of the degree of her involvement in family planning communication with others in her village. It was assumed that the nature of the communication network linking individuals with each other effects family planning adoption at the individual level. Specific objectives were to determine 1) the relative importance of the specific independent variables in explaining family planning adoption and 2) the relative importance of the community level variables in comparison with the individual level variables in explaining family planning adoption at the individual level. The data were originally gathered in a 1973 research project on Korea's mothers' clubs. 1047 respondents were interviewed, comprising all married women in 25 sample villages having mothers' clubs. The dependent variable was family planning adoption behavior, defined as current use of any of the modern methods of family planning. The independent variables were defined at 3 levels: individual, community, and at a level intermediate between them involving communication links between individuals. More of the individual level independent variables were significantly correlated with the dependent variables than the community level variables. Among those variables with statistically significant correlations, the correlation coefficients were consistently higher for the individual level than for the community level variables. More of the variance in the dependent variable was explained by individual level than by

  1. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs)....

  2. Performance of Magnet Insulation Systems at Low Temperature and After Reactor Irradiation

    SciTech Connect

    Bittner-Rohrhofer, K.; Humer, K.; Fillunger, H.; Maix, R.K.; Weber, H.W.

    2004-06-28

    Advanced composite materials reinforced with boron-free glass fibers are candidate insulation materials for fusion magnets, in particular for ITER. Thus, these systems require an excellent performance and mechanical integrity after irradiation. The present innovative organic insulation system consists of R-glass fiber reinforced tapes impregnated with an advanced cyanate-ester/epoxy resin. This composite is suitable for vacuum-pressure impregnation. In order to assess the radiation resistance of the mechanical properties, the laminate was irradiated in the TRIGA reactor (Vienna) to the ITER design fluence level of 1x1022 m-2 (E>0.1 MeV). The blend was screened at 77 K using the static tensile and short-beam-shear test prior to and after irradiation. In addition, tension-tension fatigue measurements were done in order to investigate the material performance under pulsed operating conditions.

  3. Performance of Magnet Insulation Systems at Low Temperature and After Reactor Irradiation

    NASA Astrophysics Data System (ADS)

    Bittner-Rohrhofer, K.; Humer, K.; Fillunger, H.; Maix, R. K.; Weber, H. W.

    2004-06-01

    Advanced composite materials reinforced with boron-free glass fibers are candidate insulation materials for fusion magnets, in particular for ITER. Thus, these systems require an excellent performance and mechanical integrity after irradiation. The present innovative organic insulation system consists of R-glass fiber reinforced tapes impregnated with an advanced cyanate-ester/epoxy resin. This composite is suitable for vacuum-pressure impregnation. In order to assess the radiation resistance of the mechanical properties, the laminate was irradiated in the TRIGA reactor (Vienna) to the ITER design fluence level of 1×1022 m-2 (E>0.1 MeV). The blend was screened at 77 K using the static tensile and short-beam-shear test prior to and after irradiation. In addition, tension-tension fatigue measurements were done in order to investigate the material performance under pulsed operating conditions.

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

    DOEpatents

    DeVolpi, Alexander

    1987-01-01

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

  5. Completely modular Thermionic Reactor Ion Propulsion System (TRIPS)

    NASA Technical Reports Server (NTRS)

    Peelgren, M. L.; Kikin, G. M.; Sawyer, C. D.

    1972-01-01

    The nuclear reactor powered ion propulsion system described is an advanced completely modularized system which lends itself to development of prototype and/or flight type components without the need for complete system tests until late in the development program. This modularity is achieved in all of the subsystems and components of the electric propulsion system including (1) the thermionic fuel elements, (2) the heat rejection subsystem (heat pipes), (3) the power conditioning modules, and (4) the ion thrusters. Both flashlight and external fuel type in-core thermionic reactors are considered as the power source. The thermionic fuel elements would be useful over a range of reactor power levels. Electrical heated acceptance testing in their flight configuration is possible for the external fuel case. Nuclear heated testing by sampling methods could be used for acceptance testing of flashlight fuel elements. The use of heat pipes for cooling the collectors and as a means of heat transport to the radiator allows early prototype or flight configuration testing of a small module of the heat rejection subsystem as opposed to full scale liquid metal pumps and radiators in a large vacuum chamber. The power conditioner (p/c) is arranged in modules with passive cooling.

  6. Course Management Systems and Blended Learning: An Innovative Learning Approach

    ERIC Educational Resources Information Center

    Chou, Amy Y.; Chou, David C.

    2011-01-01

    This article utilizes Rogers' innovation-decision process model (2003) and Beckman and Berry's innovation process model (2007) to create an innovative learning map that illustrates three learning methods (i.e., face-to-face learning, online learning, and blended learning) in two types of innovation (i.e., incremental innovation and radical…

  7. Systems and methods for dismantling a nuclear reactor

    DOEpatents

    Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

    2014-10-28

    Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

  8. Spreading nursing unit innovation in large hospital systems.

    PubMed

    Pearson, Marjorie L; Upenieks, Valda V; Yee, Tracy; Needleman, Jack

    2008-03-01

    Healthcare administrators increasingly face the challenge of how to spread innovation throughout their organizations. The authors present the results of an evaluation of the efforts of 3 major hospital systems to internally disseminate nursing unit change among medical-surgical units. The findings show that all 3 organizations carefully planned, coordinated, and implemented a spread process; none left dissemination to chance. Although clear differences were evident in the way they engineered their spread, many similarities also were found.

  9. Designing visual displays and system models for safe reactor operations

    SciTech Connect

    Brown-VanHoozer, S.A.

    1995-12-31

    The material presented in this paper is based on two studies involving the design of visual displays and the user`s prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator`s perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors.

  10. Slow control systems of the Reactor Experiment for Neutrino Oscillation

    NASA Astrophysics Data System (ADS)

    Choi, J. H.; Jang, H. I.; Choi, W. Q.; Choi, Y.; Jang, J. S.; Jeon, E. J.; Joo, K. K.; Kim, B. R.; Kim, H. S.; Kim, J. Y.; Kim, S. B.; Kim, S. Y.; Kim, W.; Kim, Y. D.; Ko, Y. J.; Lee, J. K.; Lim, I. T.; Pac, M. Y.; Park, I. G.; Park, J. S.; Park, R. G.; Seo, H. K.; Seo, S. H.; Shin, C. D.; Siyeon, K.; Yeo, I. S.; Yu, I.

    2016-02-01

    The RENO experiment has been in operation since August 2011 to measure reactor antineutrino disappearance using identical near and far detectors. For accurate measurements of neutrino mixing parameters and efficient data taking, it is crucial to monitor and control the detector in real time. Environmental conditions also need to be monitored for stable operation of detectors as well as for safety reasons. In this paper, we report the design, hardware, operation, and performance of the slow control system.

  11. PWR full-reactor coolant system decontamination

    SciTech Connect

    Aspden, R.G.; Pessall, N.; Grand, T.F. )

    1992-01-01

    The overall objective of the current program is to identify and address all aspects of full system decontamination with the purpose of qualifying at least one process for PWR use. The objective of the current study is to provide baseline data on the performance of materials on the primary side after exposure to one cycle of the LOMI fault testing. This data supplements prior information obtained after exposure to three cycles of LOMI testing. The technical significance of this excursion will be determined in a subsequent task. The general corrosion characteristics of over 39 materials were evaluated for some combinations of material, type of specimen (coupon and creviced coupons), and loop velocity (0, 5, 20 and 150 ft/sec). At velocities of less than or equal to 20 ft/sec, sixteen types of specimens were employed to evaluate localized corrosion and stress corrosion cracking. Specimens were examined after one cycle. Also included in this exposure were specimens added to provide more information on the effect of LOMI fault exposure one: (1) surface roughening of Stellite 156; (2) crevice corrosion of chromium plated 304 stainless steel with the open end gap increased from 3 to {approximately} 9 mils; (3) susceptibility of Inconel X-750 (HTH) to subsequent stress corrosion cracking, (4) loss of chromium plate from threads of 304 stainless steel bolts torqued into stainless steel collars; (5) crack initiation in an Alloy 600 tube known to be susceptible to primary water stress corrosion cracking; and (6) surface alternation of stressed Inconel X-750 springs with the spring temper.

  12. Real-time, multitasking control system for reactor inspection robots

    SciTech Connect

    Byrne, T.J.; Jenkins, J.B.; Lewis, W.I.; Park, L.R.; Reeves, G.E.

    1988-01-01

    The Equipment Engineering Division of the Department of Energy's Savannah River Laboratory in Aiken, South Carolina has developed a remote system to perform ultrasonic (UT) and eddy current (ET) wall weld inspections inside the nuclear reactors at the site. The basic components of the inspection system include an inspection robot and control hardware, a supervisory computer, and ultrasonic and eddy current data collection and analysis computers. The ultrasonic and eddy current systems are responsible for driving the transducers, and digitizing, displaying, and storing the information. 7 figs.

  13. Reactor Subsystem Simulation for Nuclear Hybrid Energy Systems

    SciTech Connect

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

    2012-09-01

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

  14. Removal of copper in an integrated sulfate reducing bioreactor-crystallization reactor system.

    PubMed

    Sierra-Alvarez, Reyes; Hollingsworth, Jeremy; Zhou, Michael S

    2007-02-15

    Removal of copper was investigated using an innovative water treatment system integrating a sulfidogenic bioreactor with a fluidized-bed crystallization reactor containing fine sand to facilitate the recovery of copper as a purified copper-sulfide mineral. The performance of the system was tested using a simulated semiconductor manufacturing wastewater containing high levels of Cu2+ (4-66 mg/L), sulfate, and a mixture of citrate, isopropanol, and polyethylene glycol (Mn 300). Soluble copper removal efficiencies exceeding 99% and effluent copper concentrations averaging 89 micog/L were demonstrated in the two-stage system, with near complete metal removal occurring in the crystallizer. Copper crystals deposited on sand grains were identified as covellite (CuS). The removal of organic constituents did not exceed 70% of the initial chemical oxygen demand due to incomplete degradation of isopropanol and its breakdown product (acetone). Taken as a whole, these results indicate the potential of this novel reactor configuration for the simultaneous removal of heavy metals and organic constituents. The ability of this process to recover heavy metals in a purified form makes it particularly attractive for the treatment of contaminated aqueous streams, including industrial wastewaters and acid mine drainage. PMID:17593752

  15. A small, 1400 K, reactor for Brayton space power systems.

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Mayo, W.

    1972-01-01

    An investigation was conducted to determine minimum dimensions and minimum weight obtainable in a design for a reactor using uranium-233 nitride or plutonium-239 nitride as fuel. Such a reactor had been considered by Krasner et al. (1971). Present space power status is discussed, together with questions of reactor design and power distribution in the reactor. The characteristics of various reactor types are compared, giving attention also to a zirconium hydride reactor.

  16. 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

  17. Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

    SciTech Connect

    Caira, M.; Gramiccia, L.; Naviglio, A.; Sorabella, L.; Bandini, G.

    1996-07-01

    The MARS nuclear plant is a 600 MWth PWR with completely passive core safeguards. The most relevant innovative safety system is the Emergency Core Cooling System (ECCS), which is based on natural circulation, and on a passive-type activation that follows a core flow decrease, whatever was the cause (only one component, 400% redundant, is not static). The main thermal hydraulic transients occurring as a consequence of design basis accidents for the MARS plant were presented at the ICONE 3 Conference. Those transients were analyzed in the first stage, with the aim at pointing out the capability of the innovative ECCS to intervene. So, they included only a short-time analysis (extended for a few hundreds of seconds) and the well known RELAP 5 computer program was used for this purpose. In the present paper, the long-term analyses (extended for several thousands of seconds) of the same transients are shown. These analyses confirmed that the performance of the Emergency Core Cooling System of the MARS reactor is guaranteed also in long-term scenarios.

  18. Enabling autonomous control for space reactor power systems

    SciTech Connect

    Wood, R. T.

    2006-07-01

    The application of nuclear reactors for space power and/or propulsion presents some unique challenges regarding the operations and control of the power system. Terrestrial nuclear reactors employ varying degrees of human control and decision-making for operations and benefit from periodic human interaction for maintenance. In contrast, the control system of a space reactor power system (SRPS) employed for deep space missions must be able to accommodate unattended operations due to communications delays and periods of planetary occlusion while adapting to evolving or degraded conditions with no opportunity for repair or refurbishment. Thus, a SRPS control system must provide for operational autonomy. Oak Ridge National Laboratory (ORNL) has conducted an investigation of the state of the technology for autonomous control to determine the experience base in the nuclear power application domain, both for space and terrestrial use. It was found that control systems with varying levels of autonomy have been employed in robotic, transportation, spacecraft, and manufacturing applications. However, autonomous control has not been implemented for an operating terrestrial nuclear power plant nor has there been any experience beyond automating simple control loops for space reactors. Current automated control technologies for nuclear power plants are reasonably mature, and basic control for a SRPS is clearly feasible under optimum circumstances. However, autonomous control is primarily intended to account for the non optimum circumstances when degradation, failure, and other off-normal events challenge the performance of the reactor and near-term human intervention is not possible. Thus, the development and demonstration of autonomous control capabilities for the specific domain of space nuclear power operations is needed. This paper will discuss the findings of the ORNL study and provide a description of the concept of autonomy, its key characteristics, and a prospective

  19. Reactor system for conversion of alkanes to alkenes in a external fcc catalyst cooler

    SciTech Connect

    Harandi, M.N.; Owen, H.

    1991-05-28

    This patent describes a fluid catalytic cracking reactor system. It comprises: pressurized regenerator vessel means; first valved conduit means; dehydrogenation reactor means; fluid handling means; second valved conduit means; and third conduit means.

  20. Autonomous Control Capabilities for Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

    2004-02-01

    The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

  1. Autonomous Control Capabilities for Space Reactor Power Systems

    SciTech Connect

    Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

    2004-02-04

    The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

  2. 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

  3. Pressurized water nuclear reactor system with hot leg vortex mitigator

    DOEpatents

    Lau, Louis K. S.

    1990-01-01

    A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

  4. Scale Effects on Magnet Systems of Heliotron-Type Reactors

    NASA Astrophysics Data System (ADS)

    S, Imagawa; A, Sagara

    2005-02-01

    For power plants heliotron-type reactors have attractive advantages, such as no current-disruptions, no current-drive, and wide space between helical coils for the maintenance of in-vessel components. However, one disadvantage is that a major radius has to be large enough to obtain large Q-value or to produce sufficient space for blankets. Although the larger radius is considered to increase the construction cost, the influence has not been understood clearly, yet. Scale effects on superconducting magnet systems have been estimated under the conditions of a constant energy confinement time and similar geometrical parameters. Since the necessary magnetic field with a larger radius becomes lower, the increase rate of the weight of the coil support to the major radius is less than the square root. The necessary major radius will be determined mainly by the blanket space. The appropriate major radius will be around 13 m for a reactor similar to the Large Helical Device (LHD).

  5. A Novel Approach to Material Development for Advanced Reactor Systems

    SciTech Connect

    Was, G.S.; Atzmon, M.; Wang, L.

    1999-12-22

    OAK B188 A Novel Approach to Material Development for Advanced Reactor Systems. Year one of this project had three major goals. First, to specify, order and install a new high current ion source for more rapid and stable proton irradiation. Second, to assess the use low temperature irradiation and chromium pre-enrichment in an effort to isolate a radiation damage microstructure in stainless steels without the effects of RIS. Third, to prepare for the irradiation of reactor pressure vessel steel and Zircaloy. In year 1 quarter 1, the project goal was to order the high current ion source and to procure and prepare samples of stainless steel for low temperature proton irradiation.

  6. A Novel Approach to Material Development for Advanced Reactor Systems

    SciTech Connect

    Was, G.S.; Atzmon, M.; Wang, L.

    2000-06-27

    OAK B188 A Novel Approach to Material Development for Advanced Reactor Systems. Year one of this project had three major goals. First, to specify, order and install a new high current ion source for more rapid and stable proton irradiation. Second, to assess the use of low temperature irradiation and chromium pre-enrichment in an effort to isolate a radiation damage microstructure in stainless steel without the effects of RIS. Third, to initiate irradiation of reactor pressure vessel steel and Zircaloy. In year 1 quarter 3, the project goal was to complete irradiation of model alloys of RPV steels for a range of doses and begin sample characterization. We also planned to prepare samples for microstructure isolation in stainless steels, and to identify sources of Zircaloy for irradiation and characterization.

  7. 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.

  8. System for fuel rod removal from a reactor module

    DOEpatents

    Matchett, Richard L.; Roof, David R.; Kikta, Thomas J.; Wilczynski, Rosemarie; Nilsen, Roy J.; Bacvinskas, William S.; Fodor, George

    1990-01-01

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system.

  9. System for fuel rod removal from a reactor module

    DOEpatents

    Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

    1988-07-28

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

  10. Ongoing Development of a Series Bosch Reactor System

    NASA Technical Reports Server (NTRS)

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

    2013-01-01

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

  11. Ongoing Development of a Series Bosch Reactor System

    NASA Technical Reports Server (NTRS)

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

    2013-01-01

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

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

    SciTech Connect

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

    1995-12-31

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

  13. Incorporation of Passive Safety Systems in the Generation-IV Multi-Application Small Light Water Reactor (MASLWR)

    SciTech Connect

    Modro, S. Michael; Ougouag, Abderrafi M.; Fisher, James; Weaver, Kevan

    2002-07-01

    The Idaho National Engineering and Environmental Laboratory (INEEL), Nexant Inc. and the Oregon State University (OSU) developed an innovative Multi-Application Small Light Water Reactor (MASLWR) concept. The MASLWR is a small, modular, safe, and economic natural circulation light water reactor developed with the primary goal of producing electric power, but with the flexibility to be used for water desalination or district heating with deployment in a variety of locations. The MASLWR was developed, by design, to be a safe and economic reactor concept that can be deployed in the near term by utilizing current experience and capabilities of the industry. The key features of the MASLWR concept are the extreme simplicity of the design and its passive safety systems. This paper provides an overview of safety analyses performed for the MASLWR concept and explores potential for the increase in passive safety via the implementation of new features. The results of these safety studies demonstrate that the reactor core will be provided with a stable cooling source adequate to remove decay heat without significant cladding heatup under all credible scenarios. The response of the system to accident conditions is a controlled depressurization, whereby most of the primary system blowdown occurs via the submerged ADS blowdown pathway. (authors)

  14. The detector system of the Daya Bay reactor neutrino experiment

    DOE PAGES

    An, F. P.

    2015-12-15

    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin 22θ13 and the effective mass splitting Δm2ee. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrinomore » mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors’ baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This study describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.« less

  15. The detector system of the Daya Bay reactor neutrino experiment

    NASA Astrophysics Data System (ADS)

    An, F. P.; Bai, J. Z.; Balantekin, A. B.; Band, H. R.; Beavis, D.; Beriguete, W.; Bishai, M.; Blyth, S.; Brown, R. L.; Butorov, I.; Cao, D.; Cao, G. F.; Cao, J.; Carr, R.; Cen, W. R.; Chan, W. T.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chasman, C.; Chen, H. Y.; Chen, H. S.; Chen, M. J.; Chen, Q. Y.; Chen, S. J.; Chen, S. M.; Chen, X. C.; Chen, X. H.; Chen, X. S.; Chen, Y. X.; Chen, Y.; Cheng, J. H.; Cheng, J.; Cheng, Y. P.; Cherwinka, J. J.; Chidzik, S.; Chow, K.; Chu, M. C.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dong, L.; Dove, J.; Draeger, E.; Du, X. F.; Dwyer, D. A.; Edwards, W. R.; Ely, S. R.; Fang, S. D.; Fu, J. Y.; Fu, Z. W.; Ge, L. Q.; Ghazikhanian, V.; Gill, R.; Goett, J.; Gonchar, M.; Gong, G. H.; Gong, H.; Gornushkin, Y. A.; Grassi, M.; Greenler, L. S.; Gu, W. Q.; Guan, M. Y.; Guo, R. P.; Guo, X. H.; Hackenburg, R. W.; Hahn, R. L.; Han, R.; Hans, S.; He, M.; He, Q.; He, W. S.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hinrichs, P.; Ho, T. H.; Hoff, M.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, L. M.; Hu, L. J.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. Z.; Huang, H. X.; Huang, P. W.; Huang, X.; Huang, X. T.; Huber, P.; Hussain, G.; Isvan, Z.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiang, H. J.; Jiang, W. Q.; Jiao, J. B.; Johnson, R. A.; Joseph, J.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Lai, C. Y.; Lai, W. C.; Lai, W. H.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, M. K. P.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Lewis, C. A.; Li, B.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, J.; Li, N. Y.; Li, Q. J.; Li, S. F.; Li, S. C.; Li, W. D.; Li, X. B.; Li, X. N.; Li, X. Q.; Li, Y.; Li, Y. F.; Li, Z. B.; Liang, H.; Liang, J.; Lin, C. J.; Lin, G. L.; Lin, P. Y.; Lin, S. X.; Lin, S. K.; Lin, Y. C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, B. J.; Liu, C.; Liu, D. W.; Liu, H.; Liu, J. L.; Liu, J. C.; Liu, S.; Liu, S. S.; Liu, X.; Liu, Y. B.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, A.; Luk, K. B.; Luo, T.; Luo, X. L.; Ma, L. H.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Mayes, B.; McDonald, K. T.; McFarlane, M. C.; McKeown, R. D.; Meng, Y.; Mitchell, I.; Mohapatra, D.; Monari Kebwaro, J.; Morgan, J. E.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Newsom, C.; Ngai, H. Y.; Ngai, W. K.; Nie, Y. B.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pagac, A.; Pan, H.-R.; Patton, S.; Pearson, C.; Pec, V.; Peng, J. C.; Piilonen, L. E.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, B.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Sands, W. R.; Seilhan, B.; Shao, B. B.; Shih, K.; Song, W. Y.; Steiner, H.; Stoler, P.; Stuart, M.; Sun, G. X.; Sun, J. L.; Tagg, N.; Tam, Y. H.; Tanaka, H. K.; Tang, W.; Tang, X.; Taychenachev, D.; Themann, H.; Torun, Y.; Trentalange, S.; Tsai, O.; Tsang, K. V.; Tsang, R. H. M.; Tull, C. E.; Tung, Y. C.; Viaux, N.; Viren, B.; Virostek, S.; Vorobel, V.; Wang, C. H.; Wang, L. S.; Wang, L. Y.; Wang, L. Z.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, T.; Wang, W.; Wang, W. W.; Wang, X. T.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Webber, D. M.; Wei, H. Y.; Wei, Y. D.; Wen, L. J.; Wenman, D. L.; Whisnant, K.; White, C. G.; Whitehead, L.; Whitten, C. A.; Wilhelmi, J.; Wise, T.; Wong, H. C.; Wong, H. L. H.; Wong, J.; Wong, S. C. F.; Worcester, E.; Wu, F. F.; Wu, Q.; Xia, D. M.; Xia, J. K.; Xiang, S. T.; Xiao, Q.; Xing, Z. Z.; Xu, G.; Xu, J. Y.; Xu, J. L.; Xu, J.; Xu, W.; Xu, Y.; Xue, T.; Yan, J.; Yang, C. G.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Yeh, M.; Yeh, Y. S.; Yip, K.; Young, B. L.; Yu, G. Y.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, F. H.; Zhang, H. H.; Zhang, J. W.; Zhang, K.; Zhang, Q. X.; Zhang, Q. M.; Zhang, S. H.; Zhang, X. T.; Zhang, Y. C.; Zhang, Y. H.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhao, Q. W.; Zhao, Y. F.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhou, Z. Y.; Zhuang, H. L.; Zimmerman, S.; Zou, J. H.

    2016-03-01

    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of νbare oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin2 2θ13 and the effective mass splitting Δ mee2. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors' baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This paper describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.

  16. The detector system of the Daya Bay reactor neutrino experiment

    SciTech Connect

    An, F. P.

    2015-12-15

    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin 213 and the effective mass splitting Δm2ee. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors’ baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This study describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.

  17. Nuclear plant-aging research on reactor protection systems

    SciTech Connect

    Meyer, L.C.

    1988-01-01

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed.

  18. Designing a SCADA system simulator for fast breeder reactor

    NASA Astrophysics Data System (ADS)

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

    2016-04-01

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

  19. Comparison of suspended and fixed photocatalytic reactor systems.

    PubMed

    Geissen, S U; Xi, W; Weidemeyer, A; Vogelpohl, A; Bousselmi, L; Ghrab, A; Nnabi, A E

    2001-01-01

    Photocatalysis is a promising technology for the purification of pretreated wastewaters in sun-rich countries if an economically applicable reactor system is available. Within this project the catalyst separation as an essential process step of suspended reactor systems was investigated. For the separation of suspended catalyst a sedimentation basin with and without lamella and a membrane filtration were investigated. The sedimentation was found to be very sensitive to the kind of the ion background of wastewater, the pH, the TiO2 influent concentration as well as the hydrodynamics in the clarifier. Under optimized conditions effluent concentrations of less than 5 mg SS/L and a clear water without turbidity could be reached with a specific flow rate of up to 0.7 m3/m2/h. The best performance for P25 was achieved with a TiO2 influent concentration of 5 g/L. Membrane filtration was the only method to guarantee a complete retention of the TiO2 as well as a rejection of microorganisms and high molecular compounds. With cross-flow velocities of 3 m/s and a transmembrane pressure of 100 kPa flux rates up to 1,200 L/m2/h were achieved. A flow-film-reactor (FFR) was operated with the model compound DCA under identical conditions with fixed and suspended TiO2. Whereas the fixed system has the advantage that no separation step is necessary and a simple construction can be used, suspended systems offer a three times higher reaction velocity for a catalyst concentration of 10 g/L, but are also characterized by higher investment costs.

  20. The application of an innovative continuous multiple tube reactor as a strategy to control the specific organic loading rate for biohydrogen production by dark fermentation.

    PubMed

    Gomes, Simone D; Fuess, Lucas T; Penteado, Eduardo D; Lucas, Shaiane D M; Gotardo, Jackeline T; Zaiat, Marcelo

    2015-12-01

    Biohydrogen production in fixed-bed reactors often leads to unstable and decreasing patterns because the excessive accumulation of biomass in the bed negatively affects the specific organic loading rate (SOLR) applied to the reactor. In this context, an innovative reactor configuration, i.e., the continuous multiple tube reactor (CMTR), was assessed in an attempt to better control the SOLR for biohydrogen production. The CMTR provides a continuous discharge of biomass, preventing the accumulation of solids in the long-term. Sucrose was used as the carbon source and mesophilic temperature conditions (25°C) were applied in three continuous assays. The reactor showed better performance when support material was placed in the outlet chamber to enhance biomass retention within the reactor. Although the SOLR could not be effectively controlled, reaching values usually higher than 10gsucroseg(-1)VSSd(-1), the volumetric hydrogen production and molar hydrogen production rates peaked, respectively, at 1470mLH2L(-1)d(-1) and 45mmolH2d(-1), indicating that the CMTR was a suitable configuration for biohydrogen production.

  1. Conceptual Design of Passive Safety System for Lead-Bismuth Cooled Fast Reactor

    NASA Astrophysics Data System (ADS)

    Abdullah, A. G.; Nandiyanto, A. B. D.

    2016-04-01

    This paper presents the results of the conceptual design of passive safety systems for reactor power 225 MWth using Pb-Bi coolant. Main purpose of this research is to design of heat removal system from the reactor wall. The heat from the reactor wall is removed by RVACS system using the natural circulation from the atmosphere around the reactor at steady state. The calculation is performed numerically using Newton-Raphson method. The analysis involves the heat transfer systems in a radiation, conduction and natural convection. Heat transfer calculations is performed on the elements of the reactor vessel, outer wall of guard vessel and the separator plate. The simulation results conclude that the conceptual design is able to remove heat 1.33% to 4.67% from the thermal reactor power. It’s can be hypothesized if the reactor had an accident, the system can still overcome the heat due to decay.

  2. Innovation Approaches to Development and Ground Testing of Advanced Bimodal Space Power and Propulsion Systems

    SciTech Connect

    Hill, T.; Noble, C.; Martinell, J.; Borowski, S.

    2000-07-14

    The last major development effort for nuclear power and propulsion systems ended in 1993. Currently, there is not an initiative at either the National Aeronautical and Space Administration (NASA) or the U.S. Department of Energy (DOE) that requires the development of new nuclear power and propulsion systems. Studies continue to show nuclear technology as a strong technical candidate to lead the way toward human exploration of adjacent planets or provide power for deep space missions, particularly a 15,000 lbf bimodal nuclear system with 115 kW power capability. The development of nuclear technology for space applications would require technology development in some areas and a major flight qualification program. The last major ground test facility considered for nuclear propulsion qualification was the U.S. Air Force/DOE Space Nuclear Thermal Propulsion Project. Seven years have passed since that effort, and the questions remain the same, how to qualify nuclear power and propulsion systems for future space flight. It can be reasonably assumed that much of the nuclear testing required to qualify a nuclear system for space application will be performed at DOE facilities as demonstrated by the Nuclear Rocket Engine Reactor Experiment (NERVA) and Space Nuclear Thermal Propulsion (SNTP) programs. The nuclear infrastructure to support testing in this country is aging and getting smaller, though facilities still exist to support many of the technology development needs. By renewing efforts, an innovative approach to qualifying these systems through the use of existing facilities either in the U.S. (DOE's Advance Test Reactor, High Flux Irradiation Facility and the Contained Test Facility) or overseas should be possible.

  3. Innovative Approaches to Development and Ground Testing of Advanced Bimodal Space Power and Propulsion Systems

    SciTech Connect

    Hill, Thomas Johnathan; Noble, Cheryl Ann; Noble, C.; Martinell, John Stephen; Borowski, S.

    2000-07-01

    The last major development effort for nuclear power and propulsion systems ended in 1993. Currently, there is not an initiative at either the National Aeronautical and Space Administration (NASA) or the U.S. Department of Energy (DOE) that requires the development of new nuclear power and propulsion systems. Studies continue to show nuclear technology as a strong technical candidate to lead the way toward human exploration of adjacent planets or provide power for deep space missions, particularly a 15,000 lbf bimodal nuclear system with 115 kW power capability. The development of nuclear technology for space applications would require technology development in some areas and a major flight qualification program. The last major ground test facility considered for nuclear propulsion qualification was the U.S. Air Force/DOE Space Nuclear Thermal Propulsion Project. Seven years have passed since that effort, and the questions remain the same, how to qualify nuclear power and propulsion systems for future space flight. It can be reasonable assumed that much of the nuclear testing required to qualify a nuclear system for space application will be performed at DOE facilities as demonstrated by the Nuclear Rocket Engine Reactor Experiment (NERVA) and Space Nuclear Thermal Propulsion (SNTP) programs. The nuclear infrastructure to support testing in this country is aging and getting smaller, though facilities still exist to support many of the technology development needs. By renewing efforts, an innovative approach to qualifying these systems through the use of existing facilities either in the U.S. (DOE's Advance Test Reactor, High Flux Irradiation Facility and the Contained Test Facility) or overseas should be possible.

  4. Aging study of boiling water reactor high pressure injection systems

    SciTech Connect

    Conley, D.A.; Edson, J.L.; Fineman, C.F.

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.

  5. Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

  6. An innovative algorithm for panoramic representation in observation systems

    NASA Astrophysics Data System (ADS)

    Luison, Cristian; Aquilanti, Valeria; Riccobono, Aldo; Liberace, Claudio

    2013-06-01

    This document presents the study and test carried out for the development of an innovative algorithm designed to create a panoramic representation of the scene scanned by observation systems operating with passive sensors. The purpose of the algorithm is to represent 360° of scene using staring sensors mounted on stabilized or semi-stabilized platforms, without requirements on video output, both in terms of the transmission format and in terms of frame rate. The algorithm is real-time and does not require step-and-stare technique or special systems to scan the scene. The architecture of the algorithm requires a very low computational cost for the electronics contained in a Multi-Functional Display (MDP) used in defense applications. The algorithm has been implemented and tested on the JANUS NAVAL system, where the results were very satisfactory. Today, a patent is pendent.

  7. Co-simulation of innovative integrated HVAC systems in buildings

    SciTech Connect

    Trcka, Marija; Hensena, Jan L.M.; Wetter, Michael

    2010-06-21

    Integrated performance simulation of buildings HVAC systems can help in reducing energy consumption and increasing occupant comfort. However, no single building performance simulation (BPS) tool offers sufficient capabilities and flexibilities to analyze integrated building systems and to enable rapid prototyping of innovative building and system technologies. One way to alleviate this problem is to use co-simulation, as an integrated approach to simulation. This article elaborates on issues important for co-simulation realization and discusses multiple possibilities to justify the particular approach implemented in the here described co-simulation prototype. The prototype is validated with the results obtained from the traditional simulation approach. It is further used in a proof-of-concept case study to demonstrate the applicability of the method and to highlight its benefits. Stability and accuracy of different coupling strategies are analyzed to give a guideline for the required coupling time step.

  8. Defining Innovation: Using Soft Systems Methodology to Approach the Complexity of Innovation in Educational Technology

    ERIC Educational Resources Information Center

    Cox, Glenda

    2010-01-01

    This paper explores what educational technologists in one South African Institution consider innovation to be. Ten educational technologists in various faculties across the university were interviewed and asked to define and answer questions about innovation. Their answers were coded and the results of the overlaps in coding have been assimilated…

  9. Role of Brokerage in Evolving Innovation Systems: A Case of the Fodder Innovation Project in Nigeria

    ERIC Educational Resources Information Center

    Madzudzo, Elias

    2011-01-01

    This paper looks at brokerage functions in a project on building innovation capacity through improved networking. Innovation capacity influences how actors respond to changes in their environments. In such dynamic environments well connected sets of actors are at an advantage in that they can combine skills to address the emerging opportunities…

  10. Indigenous African Knowledge Systems and Innovation in Higher Education in South Africa

    ERIC Educational Resources Information Center

    Higgs, P.; Higgs, L. G.; Venter, E.

    2003-01-01

    The importance of innovation in higher education is recognised in South African educational discourse. The South African White Paper on Science and Technology, issued in September 1996 and entitled, "Preparing for the 21st Century", states that, "the White Paper is built upon the twin concepts of "innovation" and a "national system of innovation"…

  11. An evaluation of some innovative fragmentation systems for oil shale

    SciTech Connect

    Hieta, M.; Hustrulid, W.A.

    1991-06-01

    This report describes a large-scale underground mining method, large-hole stoping, using some innovative fragmentation systems (buffer blasting, continuous loading/hauling, and mechanical miners for development). This study includes a literature review and an experimental study of one of the key design factors--buffer blasting. The purpose of the buffer-blasting experiments is to examine the swell that is necessary to achieve satisfactory fragmentation results. The study also includes a technical and economic evaluation of the new mining method compared with conventional room and pillar mining. The purpose of this study is to examine innovative methods that exist today and may provide a more efficient mining system than that currently used. Note that this is a conceptual study, and that the mining for the two mine designs were compared using a daily production rate of 75,000 tones per day. This amount was chosen because it is the maximum amount possible for a rubber-tired room and pillar operation with only a one-shaft complex.

  12. A highly innovative global broadband mobile communication system concept

    NASA Astrophysics Data System (ADS)

    Wittig, Manfred

    2010-04-01

    Global mobile communication systems deployed in the 1990s (Iridium and Globalstar) failed to achieve the expected market success. However, recently a second generation of these two LEO constellations have been or will be procured. This shows evidence that global mobile satellite communications are attractive for a certain class of users. Basic requirements for mobile satellite communication networks are global coverage at high elevation angles, continuous availability and most important attractive service offers. For the first two reasons LEO constellations were designed and deployed. The large number of required satellites (remember the original Teledesic concept was designed containing 860 satellites) increases the CAPEX and consequently the service cost. This is probably the major reason for the limited commercial success of Iridium and Globalstar. An ideal global mobile (broadband) communication system shall provide global coverage at high elevation angles (not below e.g. 30°) with the lowest number of satellites. This seems to be contradictory. However, a solution will be presented that achieves global coverage which requires only five satellites and achieves high elevation angles even at polar locations. This orbit constellation is to be complemented with innovative payloads to achieve cost efficient broadband mobile services all over the Globe. Innovative L/S-band payload concepts will be introduced as well as Ka-band payloads.

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

    SciTech Connect

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

    2003-09-01

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

  14. Chemical Looping Combustion System-Fuel Reactor Modeling

    SciTech Connect

    Gamwo, I.K.; Jung, J.; Anderson, R.R.; Soong, Y.

    2007-04-01

    Chemical looping combustion (CLC) is a process in which an oxygen carrier is used for fuel combustion instead of air or pure oxygen as shown in the figure below. The combustion is split into air and fuel reactors where the oxidation of the oxygen carrier and the reduction of the oxidized metal occur respectively. The CLC system provides a sequestration-ready CO2 stream with no additional energy required for separation. This major advantage places combustion looping at the leading edge of a possible shift in strict control of CO2 emissions from power plants. Research in this novel technology has been focused in three distinct areas: techno-economic evaluations, integration of the system into power plant concepts, and experimental development of oxygen carrier metals such as Fe, Ni, Mn, Cu, and Ca. Our recent thorough literature review shows that multiphase fluid dynamics modeling for CLC is not available in the open literature. Here, we have modified the MFIX code to model fluid dynamic in the fuel reactor. A computer generated movie of our simulation shows bubble behavior consistent with experimental observations.

  15. High Flux Isotope Reactor system RELAP5 input model

    SciTech Connect

    Morris, D.G.; Wendel, M.W.

    1993-01-01

    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

  16. Nuclear reactor cooling system decontamination reagent regeneration. [PWR; BWR

    DOEpatents

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P. Jr.

    1980-06-06

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

  17. Modular Code and Data System for Fast Reactor Neutronics Analyses

    2008-06-30

    Version 00. The European Reactor ANalysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R&Dmore » organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core follow-up modules (connected

  18. Reactor/Brayton power systems for nuclear electric spacecraft

    NASA Technical Reports Server (NTRS)

    Layton, J. P.

    1980-01-01

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

  19. Technological implications of SNAP reactor power system development on future space nuclear power systems

    SciTech Connect

    Anderson, R.V.

    1982-11-16

    Nuclear reactor systems are one method of satisfying space mission power needs. The development of such systems must proceed on a path consistent with mission needs and schedules. This path, or technology roadmap, starts from the power system technology data base available today. Much of this data base was established during the 1960s and early 1970s, when government and industry developed space nuclear reactor systems for steady-state power and propulsion. One of the largest development programs was the Systems for Nuclear Auxiliary Power (SNAP) Program. By the early 1970s, a technology base had evolved from this program at the system, subsystem, and component levels. There are many implications of this technology base on future reactor power systems. A review of this base highlights the need for performing a power system technology and mission overview study. Such a study is currently being performed by Rockwell's Energy Systems Group for the Department of Energy and will assess power system capabilities versus mission needs, considering development, schedule, and cost implications. The end product of the study will be a technology roadmap to guide reactor power system development.

  20. Optics designs for an innovative LED lamp family system

    NASA Astrophysics Data System (ADS)

    Weiss, Herbert; Muschaweck, Julius; Hadrath, Stefan; Kudaev, Sergey

    2011-10-01

    On the general lighting market of LED lamps for professional applications there are still mainly products for single purpose solutions existing. There is a lack of standardised lamp systems like they are common for conventional lighting technologies. Therefore, an LED lamp family system was studied using high power LED with the objective to entirely substitute standard conventional lamp families in general lighting applications in the professional market segment. This comprises the realization of sets of lamp types with compact and linear shapes as well as with light distribution characteristics ranging from diffuse to extreme collimation and exceptionally high candle power. Innovative secondary optics concepts are discussed which allow both, the design of lamps with non-bulky shape and to obtain sufficient colour mixing when using multicolour LED combinations in order to achieve a very high colour rendering quality.

  1. Innovative Detection System of Ochratoxin A by Thin Film Photodiodes

    PubMed Central

    Caputo, Domenico; de Cesare, Giampiero; Fanelli, Corrado; Nascetti, Augusto; Ricelli, Alessandra; Scipinotti, Riccardo

    2007-01-01

    In this work we present, for the first time, a rapid, compact and innovative method for detection of Ochratoxin A (OTA) based on hydrogenated amorphous silicon (a-Si:H) sensors. 2 μl of acidified toluene containing OTA at different concentrations were spotted on the silica side of a High Performance Thin Layer Cromatography plate and aligned with a a-Si:H p-i-n photodiode deposited by Plasma Enhanced Chemical Vapor Deposition on a different glass substrate. As an UV radiation excites the mycotoxin, the re-emitted light is detected by the a-Si:H sensor. Results show a very good linearity between OTA concentration and the sensor photocurrent over almost three orders of magnitude. The minimum detected OTA concentration is equal to 0.1ng, showing that the presented system has the potential for a low cost system suitable for the early detection of toxins in foods.

  2. Innovative production system goes in off Ivory Coast

    SciTech Connect

    Childers, M.; Barnes, J. |

    1995-07-17

    The phased field development of the Lion and Panthere fields, offshore the Ivory Coast, includes a small floating production, storage, and offloading (FPSO) tanker with minimal processing capability as an early oil production system (EPS). For the long-term production scheme, the FPSO will be replaced by a converted jack up mobile offshore production system (MOPS) with full process equipment. The development also includes guyed-caisson well platforms, pipeline export for natural gas to fuel an onshore power plant, and a floating storage and offloading (FSO) tanker for oil export. Pipeline export for oil is a future possibility. This array of innovative strategies and techniques seldom has been brought together in a single project. The paper describes the development plan, early oil, jack up MOPS, and transport and installation.

  3. Innovation in sodium erythorbate production: the use of membrane-reactors.

    PubMed

    Spigno, G; Pagella, C; De Faveri, M D

    2001-01-01

    Isoascorbic or erythorbic acid is a stereoisomer of ascorbic acid acting as a preservative against oxidation and decoloration, recently approved for food use also in the European Market. Actually erythorbate is produced by means of a complex and long process after bio-fermentation of dextrose. In order to simplify and improve this process a membrane system was conceived able to induce acid 2-ketogluconic diffusion from fermentation broth directly into methanol where it esterificates. Acid 2-ketogluconic methyl ester is the intermediate for erythorbic acid. The principal problem was to perfectly separate water from methanol to avoid saponification instead of esterification.

  4. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    NASA Astrophysics Data System (ADS)

    Malathi, N.; Sahoo, P.; Ananthanarayanan, R.; Murali, N.

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ˜100 Hz/mm, ˜1 s, and ˜0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

  5. Level monitoring system with pulsating sensor--application to online level monitoring of dashpots in a fast breeder reactor.

    PubMed

    Malathi, N; Sahoo, P; Ananthanarayanan, R; Murali, N

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation. PMID:25725884

  6. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    SciTech Connect

    Malathi, N.; Sahoo, P. Ananthanarayanan, R.; Murali, N.

    2015-02-15

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control and Safety Rod Drive Mechanism during reactor operation.

  7. Level monitoring system with pulsating sensor--application to online level monitoring of dashpots in a fast breeder reactor.

    PubMed

    Malathi, N; Sahoo, P; Ananthanarayanan, R; Murali, N

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

  8. FFTF (Fast Flux Test Facility) reactor shutdown system reliability reevaluation

    SciTech Connect

    Pierce, B.F.

    1986-07-01

    The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations.

  9. FAFTRCS: an experiment in computerized reactor safety systems

    SciTech Connect

    Chisholm, G.H.

    1985-01-01

    Nuclear Power Plant availability and reliability could be improved by the integration of computers into the control environment. However, computer-based systems are historically viewed as being unreliable. This places a burden upon the designer to demonstrate adequate reliability and availability for the computer. The complexity associated with computers coupled with the manual nature of these demonstrations results in a high cost which typically has been justified for critical applications only. This paper investigates a methodology for automating this process and discusses a project which intends to apply this methodology to design verification and validation for a control system which will be installed and tested in an actual reactor control environment. 7 refs., 4 figs., 1 tab.

  10. 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

  11. Innovation, Innovation, Innovation

    ERIC Educational Resources Information Center

    Schuller, Tom

    2007-01-01

    Innovation, Universities and Skills. The new title of the department offers much food for thought. The title is indeed an intriguing and important one. Bringing the idea of innovation right to the fore is, to use an overworked term, challenging. Pinning down what innovation means is not at all easy. There are three different lines of argument. The…

  12. The IRIS Spool-Type Reactor Coolant Pump

    SciTech Connect

    Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

    2002-07-01

    IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)

  13. Design and Implementation of an Innovative Residential PV System

    NASA Astrophysics Data System (ADS)

    Najm, Elie Michel

    This work focuses on the design and implementation of an innovative residential PV system. In chapter one, after an introduction related to the rapid growth of solar systems' installations, the most commonly used state of the art solar power electronics' configurations are discussed, which leads to introducing the proposed DC/DC parallel configuration. The advantages and disadvantages of each of the power electronics' configurations are deliberated. The scope of work in the power electronics is defined in this chapter to be related to the panel side DC/DC converter. System integration and mechanical proposals are also within the scope of work and are discussed in later chapters. Operation principle of a novel low cost PV converter is proposed in chapter 2. The proposal is based on an innovative, simplified analog implementation of a master/slave methodology resulting in an efficient, soft-switched interleaved variable frequency flybacks, operating in the boundary conduction mode (BCM). The scheme concept and circuit configuration, operation principle and theoretical waveforms, design equations, and design considerations are presented. Furthermore, design examples are also given, illustrating the significance of the newly derived frequency equation for flybacks operating in BCM. In chapters 3, 4, and 5, the design implementation and optimization of the novel DC/DC converter illustrated in chapter 2 are discussed. In chapter 3, a detailed variable frequency BCM flyback design model leading to optimizing the component selections and transformer design, detailed in chapter 4, is presented. Furthermore, in chapter 4, the method enabling the use of lower voltage rating switching devices is also discussed. In chapter 5, circuitry related to Start-UP, drive for the main switching devices, zero-voltage-switching (ZVS) as well as turn OFF soft switching and interleaving control are fully detailed. The experimental results of the proposed DC/DC converter are presented in

  14. 77 FR 15812 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-16

    ... COMMISSION Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors AGENCY: Nuclear...-1265, ``Initial Test Program of Condensate and Feedwater Systems for Light- Water Reactors.'' DG-1265... plant startup, and power ascension tests for the condensate and feedwater systems in all types of...

  15. Supervisory Control System Architecture for Advanced Small Modular Reactors

    SciTech Connect

    Cetiner, Sacit M; Cole, Daniel L; Fugate, David L; Kisner, Roger A; Melin, Alexander M; Muhlheim, Michael David; Rao, Nageswara S; Wood, Richard Thomas

    2013-08-01

    This technical report was generated as a product of the Supervisory Control for Multi-Modular SMR Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (SMR) Research and Development Program of the U.S. Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular advanced SMR (AdvSMR) plants. This research activity advances the state-of-the art by incorporating decision making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides a brief history of hierarchical functional architectures and the current state-of-the-art, describes a reference AdvSMR to show the dependencies between systems, presents a hierarchical structure for supervisory control, indicates the importance of understanding trip setpoints, applies a new theoretic approach for comparing architectures, identifies cyber security controls that should be addressed early in system design, and describes ongoing work to develop system requirements and hardware/software configurations.

  16. Ultrasonic processing of dairy systems in large scale reactors.

    PubMed

    Zisu, Bogdan; Bhaskaracharya, Raman; Kentish, Sandra; Ashokkumar, Muthupandian

    2010-08-01

    High intensity low frequency ultrasound was used to process dairy ingredients to improve functional properties. Based on a number of lab-scale experiments, several experimental parameters were optimised for processing large volumes of whey and casein-based dairy systems in pilot scale ultrasonic reactors. A continuous sonication process at 20 kHz capable of delivering up to 4 kW of power with a flow-through reactor design was used to treat dairy ingredients at flow rates ranging from 200 to 6000 mL/min. Dairy ingredients treated by ultrasound included reconstituted whey protein concentrate (WPC), whey protein and milk protein retentates and calcium caseinate. The sonication of solutions with a contact time of less than 1 min and up to 2.4 min led to a significant reduction in the viscosity of materials containing 18% to 54% (w/w) solids. The viscosity of aqueous dairy ingredients treated with ultrasound was reduced by between 6% and 50% depending greatly on the composition, processing history, acoustic power and contact time. A notable improvement in the gel strength of sonicated and heat coagulated dairy systems was also observed. When sonication was combined with a pre-heat treatment of 80 degrees C for 1 min or 85 degrees C for 30s, the heat stability of the dairy ingredients containing whey proteins was significantly improved. The effect of sonication was attributed mainly to physical forces generated through acoustic cavitation as supported by particle size reduction in response to sonication. As a result, the gelling properties and heat stability aspects of sonicated dairy ingredients were maintained after spray drying and reconstitution. Overall, the sonication procedure for processing dairy systems may be used to improve process efficiency, improve throughput and develop value added ingredients with the potential to deliver economical benefits to the dairy industry.

  17. Advanced Recyclable Media System{reg_sign}. Innovative technology summary report

    SciTech Connect

    1998-12-01

    The objective of the Large-Scale Demonstration Project (LSDP) is to select and demonstrate potentially beneficial technologies at the Argonne National Laboratory East`s (ANL) Chicago Pile-5 (CP-5) Research Reactor. The purpose of the LSDP is to demonstrate that using innovative and improved deactivation and decommissioning (D and D) technologies from various sources can result in significant benefits, such as decreased cost and increased health and safety, as compared with baseline D and D technologies. This report describes a demonstration of the Advanced Recyclable Media System{reg_sign} technology which was employed by Surface Technology Systems, Inc. to remove coatings from a concrete floor. This demonstration is part of the CP-5 LSDP sponsored by the US Department of Energy (DOE) Office of Science and Technology Deactivation and Decommissioning Focus Area (DDFA). The Advanced Recyclable Media System{reg_sign} (ARMS) technology is an open blast technology which uses a soft recyclable media. The patented ARMS Engineered Blast Media consists of a fiber-reinforced polymer matrix which can be manufactured in various grades of abrasiveness. The fiber media can be remade and/or reused up to 20 times and can clean almost any surface (e.g., metal, wood, concrete, lead) and geometry including corners and the inside of air ducts.

  18. A New Approach for Analysing National Innovation Systems in Emerging and Developing Countries

    ERIC Educational Resources Information Center

    Seidel, Uwe; Muller, Lysann; Meier zu Kocker, Gerd; Filho, Guajarino de Araujo

    2013-01-01

    This paper presents a tool for the indicator-based analysis of national innovation systems (ANIS). ANIS identifies the economic strengths and weaknesses of a country-wide, regional or local system and includes a comprehensive examination and evaluation of the status of existing innovation systems. The use of a particular form of expert interviews…

  19. Testing of an advanced thermochemical conversion reactor system

    SciTech Connect

    Not Available

    1990-01-01

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

  20. Compatibility of refractory materials for nuclear reactor poison control systems

    NASA Technical Reports Server (NTRS)

    Sinclair, J. H.

    1974-01-01

    Metal-clad poison rods have been considered for the control system of an advanced space power reactor concept studied at the NASA Lewis Research Center. Such control rods may be required to operate at temperatures of about 140O C. Selected poison materials (including boron carbide and the diborides of zirconium, hafnium, and tantalum) were subjected to 1000-hour screening tests in contact with candidate refractory metal cladding materials (including tungsten and alloys of tantalum, niobium, and molybdenum) to assess the compatibility of these materials combinations at the temperatures of interest. Zirconium and hafnium diborides were compatible with refractory metals at 1400 C, but boron carbide and tantalum diboride reacted with the refractory metals at this temperature. Zirconium diboride also showed promise as a reaction barrier between boron carbide and tungsten.

  1. Silicon doping system at the research reactor FRM II.

    PubMed

    Li, X; Gerstenberg, H; Neuhaus, I

    2009-01-01

    Silicon doping has being carried out at FRM II since 2 years. During the commissioning of our new reactor, a simple test rig was used to determine the neutron flux profile at the irradiation position and optimise a nickel absorber liner, which is equipped at the irradiation position for vertical smoothing of the neutron flux profile. MCNP-code was used during the design of the liner. The final automatic doping system is designed to allow the irradiation of cylindrical silicon single crystals 500mm high and up to 200mm in diameter. Silicon ingots are additionally rotated continuously about their own cylinder axis during irradiation. The neutron flux density is measured online by using self-powered-neutron (SPN) detectors. The necessary doping homogeneity of +/-5% is achieved. The doping procedure and doping quality of ingots with high target resistivity are also discussed. PMID:19324563

  2. An Innovative Ceramic Corrosion Protection System for Zircaloy Cladding

    SciTech Connect

    Ronald H. Baney, Dr. D. Butt, Dr. P. Demkowicz, Dr. G. Fuchs Department of Materials Science; James S. Tulenko, Department of Nuclear and Radiological Engineering; University of Florida.

    2003-02-19

    Light Water reactor (LWR) fuel performance is currently limited by thermal, chemical and mechanical constraints associated with the design, fabrication, and operation of the fuel in incore operation. Corrosion of the zirconium based (Zircaloy-4) alloy cladding of the fuel is a primary limiting factor. Recent success at the University of Florida in developing thin ceramic films with great adhesive properties for metal substrates offers an innovative breakthrough for eliminating a major weakness of the Zircaloy clad. ?The University of Florida proposes to coat the existing Zircaloy clad tubes with a ceramic coating for corrosion protection. An added bonus of this approach would be the implementation of a boron-containing burnable poison outer layer will also be demonstrated as part of the ceramic coating development. In this proposed effort, emphasis will be on the ceramic coating with only demonstration of feasibility on the burnable outer coating approach. This proposed program i s expected to give a step change (approximately a doubling) in clad lifetime before failure due to corrosion. In the development of ceramic coatings for Zircaloy-4 clad, silicon carbide and zirconium carbide coatings will first be applied to Zircaloy-4 coupons and cladding samples by thermal assisted chemical vapor deposition, plasma assisted chemical vapor deposition or by laser ablation deposition. All of these processes are in use at the University of Florida and have shown great potential. The questions of adhesion and thermal expansion mismatch of the ceramic coating to the Zircaloy substrate will be addressed. Several solutions to these conditions will be examined, if needed. These solutions include the use of a zirconium oxide compliant layer, employment of a laser roughened surface and the use of a gradient composition interlayer. These solutions have already been shown to be effective for other high modulus coatings on metal substrates. Mechanical properties and adhesion of the

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

    NASA Technical Reports Server (NTRS)

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

    2009-01-01

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

  4. Innovative tephra studies in the East African Rift System

    NASA Astrophysics Data System (ADS)

    WoldeGabriel, Giday; Hart, William K.; Heiken, Grant

    Geosciences investigations form the foundation for paleoanthropological research in the East African Rift System. However, innovative applications of tephra studies for constraining spatial and temporal relations of diverse geological processes, biostratigraphic records, and paleoenvironmental conditions within the East African Rift System were fueled by paleoanthropological investigations into the origin and evolution of hominids and material culture. Tephra is a collective, size-independent term used for any material ejected during an explosive volcanic eruption.The East African Rift System has become a magnet for paleoanthropological research ever since the discovery of the first hominids at Olduvai Gorge, in Tanzania, in the 1950s [Leakey et al., 1961]. Currently, numerous multidisciplinary scientific teams from academic institutions in the United States and Western Europe make annual pilgrimages for a couple of months to conduct paleoanthropological field research in the fossil-rich sedimentary deposits of the East African Rift System in Ethiopia, Kenya, and Tanzania. The field expedition consists of geological, paleontological, archaeological, and paleoenvironmental investigations.

  5. Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

    SciTech Connect

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

    The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values.

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

    SciTech Connect

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

    1991-12-31

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

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

    PubMed

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

    1991-01-01

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

  8. Code System for Supercritical Water Cooled Reactor LOCA Analysis.

    1999-10-13

    Version 00 The new SCRELA code was developed to analyze the LOCA of the supercritical water cooled reactor. Since the currently available LWR codes for LOCA analysis could not analyze the significant differences in reactor characteristics between the supercritical-water cooled reactor and the current LWR, the first objective of this code development was to analyze the uniqueness of this reactor. The behavior of the supercritical water in the blowdown phase and the reflood phase ismore » modeled.« less

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

    SciTech Connect

    Karpov, A. S.

    2013-01-15

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

  10. 78 FR 28896 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-16

    ... COMMISSION Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components... Regulatory Commission (NRC) is issuing Revision 2 to Regulatory Guide (RG) 1.57, ``Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components,'' in which there are no...

  11. Automated Microdosing System for Integration With a Miniaturized High-pressure Reactor System

    PubMed Central

    Hawali, Ihsan; Thurow, Kerstin

    2005-01-01

    We present a new automated dosing system developed by the Institute for Automation of the University of Rostock, Germany. The new system is designed for the dosing of chemical liquids in the range of 50 μL–2.5 mL. It is integrated into a miniaturized reactor system to be used in the field of combinatorial synthesis. The reactor system can be pressurized up to 150 bar and tempered up to 200^C. A wide range of liquids with different physical properties can be handled with the new dosing system. A detailed description of the new dosing system in terms of function and operation as well as the relevant features and potential benefits is provided. PMID:18924738

  12. Automated Microdosing System for Integration With a Miniaturized High-pressure Reactor System.

    PubMed

    Stoll, Norbert; Hawali, Ihsan; Thurow, Kerstin

    2005-01-01

    We present a new automated dosing system developed by the Institute for Automation of the University of Rostock, Germany. The new system is designed for the dosing of chemical liquids in the range of 50 muL-2.5 mL. It is integrated into a miniaturized reactor system to be used in the field of combinatorial synthesis. The reactor system can be pressurized up to 150 bar and tempered up to 200(;)C. A wide range of liquids with different physical properties can be handled with the new dosing system. A detailed description of the new dosing system in terms of function and operation as well as the relevant features and potential benefits is provided.

  13. Laser in vessel-viewing system for nuclear fusion reactors

    NASA Astrophysics Data System (ADS)

    Bartolini, Luciano; Bordone, Andrea; Coletti, Alberto; Ferri De Collibus, Mario; Fornetti, Giorgio G.; Lupini, S.; Neri, Carlo; Poggi, Claudio; Riva, Marco; Semeraro, Luigi; Talarico, Carlo

    2000-11-01

    An amplitude modulated laser radar has been developed by ENEA (Italian Agency for New Technologies, Energy and Environment) for periodic in-vessel inspection in large fusion machines. Its overall optical design has been developed taking into account the extremely high radiation levels and operating temperatures foreseen in large European fusion machines such as JET (Joint European Torus) and ITER (International Thermo- nuclear Experimental Reactor). The viewing system is based on a transceiving optical radar using a RF modulated single mode 840 nm wavelength laser beam. The sounding beam is transmitted through a coherent optical fiber and a focusing optic to the inner part of the nuclear reactor vessel by a stainless steel probe on the tip of which a suitable scanning silica prism steers the laser beam along a linear raster spanning a -90 degree(s) to +60 degree(s) in elevation and 360 degree(s) in azimuth for a complete mapping of the vessel itself. All the electronics, including the laser source, avalanche photodiode and all the active components are located outside the bioshield, while passive components (receiving optics, transmitting collimator, fiber optics), located in the torus hall, are made of fused silica so that the overall laser radar is radiation resistant. The signal is acquired, the raster lines being synchronized with the aid of optical encoders linked to the scanning prism, thus yielding a TV like image. Preliminary results have been obtained scanning large sceneries including several real targets having different backscattering properties, colors and surface reflectivity ranging over several decades to simulate the expected dynamic range of the video signals incoming from the vessel.

  14. Instrumentation Needs for Integral Primary System Reactors (IPSRs) - Task 1 Final Report

    SciTech Connect

    Gary D. Storrick; Bojan Petrovic; Luca Oriani; Lawrence E. Conway; Diego Conti

    2005-09-30

    This report presents the results of the Westinghouse work performed under Task 1 of this Financial Assistance Award and satisfies a Level 2 Milestone for the project. While most of the signals required for control of IPSRs are typical of other PWRs, the integral configuration poses some new challenges in the design or deployment of the sensors/instrumentation and, in some cases, requires completely new approaches. In response to this consideration, the overall objective of Task 1 was to establish the instrumentation needs for integral reactors, provide a review of the existing solutions where available, and, identify research and development needs to be addressed to enable successful deployment of IPSRs. The starting point for this study was to review and synthesize general characteristics of integral reactors, and then to focus on a specific design. Due to the maturity of its design and availability of design information to Westinghouse, IRIS (International Reactor Innovative and Secure) was selected for this purpose. The report is organized as follows. Section 1 is an overview. Section 2 provides background information on several representative IPSRs, including IRIS. A review of the IRIS safety features and its protection and control systems is used as a mechanism to ensure that all critical safety-related instrumentation needs are addressed in this study. Additionally, IRIS systems are compared against those of current advanced PWRs. The scope of this study is then limited to those systems where differences exist, since, otherwise, the current technology already provides an acceptable solution. Section 3 provides a detailed discussion on instrumentation needs for the representative IPSR (IRIS) with detailed qualitative and quantitative requirements summarized in the exhaustive table included as Appendix A. Section 3 also provides an evaluation of the current technology and the instrumentation used for measurement of required parameters in current PWRs. Section 4

  15. A Systems Dynamics Model of Implementation of an Innovation.

    ERIC Educational Resources Information Center

    Gaynor, Alan K.; And Others

    The research presented in this report investigated the critical factors that affected the decision to abandon or replace a curricular innovation in one elementary school. The specific innovation examined in this research is called developing mathematical processes, which emphasizes process and induction rather than computational skills. Although…

  16. Role of Universities in the National Innovation System. Discussion Paper

    ERIC Educational Resources Information Center

    Group of Eight (NJ1), 2011

    2011-01-01

    Over recent years governments have been placing more emphasis on innovation as a source of national competitiveness. Governments now assess their investments across many areas in terms of the contribution that such investments make to increasing innovation. This has been especially significant for education and in particular for the development of…

  17. Innovation in Open Systems: A Comparative Study of Banks.

    ERIC Educational Resources Information Center

    Young, Robert L.; And Others

    1981-01-01

    Explains the extent to which the innovativeness of banks (as measured by the adoption of credit cards and computers) is affected by competition, growth, size, and departmentalization. Notes that size and growth are more significantly related to innovation than are departmentalization and competition. (SB)

  18. System simulation of a multicell thermionic space power reactor

    NASA Astrophysics Data System (ADS)

    von Arx, Alan Vincent

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

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

    NASA Astrophysics Data System (ADS)

    Determan, William R.; Otting, William D.

    1992-01-01

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

  20. Heat stress monitoring system. Innovative technology summary report

    SciTech Connect

    1998-11-01

    The US Department of Energy`s (DOE) nuclear facility decontamination and decommissioning (D and D) program involves the need to decontaminate and decommission buildings expeditiously and cost-effectively. Simultaneously, the health and safety of personnel involved in the D and D activities is of primary concern. Often, D and D workers must perform duties in inclement weather, and because they also frequently work in contaminated areas, they must wear personal protective clothing and/or respirators. Monitoring the health status of workers under these conditions is an important component of ensuring their safety. The MiniMitter VitalSense Telemetry System`s heat stress monitoring system (HSMS) is designed to monitor the vital signs of individual workers as they perform work in conditions that might be conducive to heat exhaustion or heat stress. The HSMS provides real-time data on the physiological condition of workers which can be monitored to prevent heat stress or other adverse health situations. This system is particularly useful when workers are wearing personal protective clothing or respirators that make visual observation of their condition more difficult. The MiniMitter VitalSense Telemetry System can monitor up to four channels (e.g., heart rate, body activity, ear canal, and skin temperature) and ten workers from a single supervisory station. The monitors are interfaced with a portable computer that updates and records information on individual workers. This innovative technology, even though it costs more, is an attractive alternative to the traditional (baseline) technology, which measures environmental statistics and predicts the average worker`s reaction to those environmental conditions without taking the physical condition of the individual worker into consideration. Although use of the improved technology might be justified purely on the basis of improved safety, it has the potential to pay for itself by reducing worker time lost caused by heat

  1. Implementation of a Technological Innovation: Factors Influencing the Adoption of a New Student Information System

    ERIC Educational Resources Information Center

    Johnston, Katrina M.

    2013-01-01

    School information systems (SIS) have the potential to cause a change in a school's technical, structural, psycho-social, and managerial systems. Implementation of a technological innovation such as an SIS is not a one-step occurrence; it is a process that occurs over time. Implementing any technological innovation involves active learning…

  2. The Practice Teaching Platform and System for Innovating Talents and Performing Research

    ERIC Educational Resources Information Center

    Wang, Boyu; Hao, Ziqiang

    2014-01-01

    The practice education is an important part of higher educational system and an important approach to cultivating applied innovative talents. This paper studies practice of teaching platform and practical teaching system for the research and practice based on the objective basis of applied innovative talents and the basic characteristics of the…

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

    SciTech Connect

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

    1983-06-30

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

  4. Technological innovations in the development of cardiovascular clinical information systems.

    PubMed

    Hsieh, Nan-Chen; Chang, Chung-Yi; Lee, Kuo-Chen; Chen, Jeen-Chen; Chan, Chien-Hui

    2012-04-01

    Recent studies have shown that computerized clinical case management and decision support systems can be used to assist surgeons in the diagnosis of disease, optimize surgical operation, aid in drug therapy and decrease the cost of medical treatment. Therefore, medical informatics has become an extensive field of research and many of these approaches have demonstrated potential value for improving medical quality. The aim of this study was to develop a web-based cardiovascular clinical information system (CIS) based on innovative techniques, such as electronic medical records, electronic registries and automatic feature surveillance schemes, to provide effective tools and support for clinical care, decision-making, biomedical research and training activities. The CIS developed for this study contained monitoring, surveillance and model construction functions. The monitoring layer function provided a visual user interface. At the surveillance and model construction layers, we explored the application of model construction and intelligent prognosis to aid in making preoperative and postoperative predictions. With the use of the CIS, surgeons can provide reasonable conclusions and explanations in uncertain environments.

  5. A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system

    SciTech Connect

    Bartram, B.W.; Dougherty, D.K.

    1987-01-01

    This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs. (TEM)

  6. Best Estimate Code System to Calculate Thermal & Hydraulic Phenomena in a Nuclear Reactor or Related System.

    1999-05-19

    Version 00 RELAP4/MOD7/101 performs best estimate analyses of nuclear reactors or related systems undergoing a transient. Transient thermal-hydraulic, two-phase phenomena are calculated from formulations of one-dimensional, homogeneous, equilibrium conservation equations for water mass, momentum, and energy. Heat structures are modeled using a transient one-dimensional heat conduction solution that is coupled to the fluid through heat transfer relations. Various explicit models are used to calculate nonhomogeneous, nonequilibrium behavior including a phase separation model, a vertical slipmore » model, and a nonequilibrium model. Other models are used to represent critical flow, reactor kinetics, pressurized water reactor reflood behavior, nuclear fuel rod swelling and blockage, and components such as pumps, valves, and accumulators.« less

  7. Shielding considerations for advanced space nuclear reactor systems

    SciTech Connect

    Angelo, J.P. Jr.; Buden, D.

    1982-01-01

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

  8. Hanging core support system for a nuclear reactor. [LMFBR

    DOEpatents

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-04-26

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

  9. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    SciTech Connect

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  10. Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems.

    1999-07-23

    AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energymore » deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations.« less

  11. Health Systems Innovation at Academic Health Centers: Leading in a New Era of Health Care Delivery.

    PubMed

    Ellner, Andrew L; Stout, Somava; Sullivan, Erin E; Griffiths, Elizabeth P; Mountjoy, Ashlin; Phillips, Russell S

    2015-07-01

    Challenged by demands to reduce costs and improve service delivery, the U.S. health care system requires transformational change. Health systems innovation is defined broadly as novel ideas, products, services, and processes-including new ways to promote healthy behaviors and better integrate health services with public health and other social services-which achieve better health outcomes and/or patient experience at equal or lower cost. Academic health centers (AHCs) have an opportunity to focus their considerable influence and expertise on health systems innovation to create new approaches to service delivery and to nurture leaders of transformation. AHCs have traditionally used their promotions criteria to signal their values; creating a health systems innovator promotion track could be a critical step towards creating opportunities for innovators in academic medicine. In this Perspective, the authors review publicly available promotions materials at top-ranked medical schools and find that while criteria for advancement increasingly recognize systems innovation, there is a lack of specificity on metrics beyond the traditional yardstick of peer-reviewed publications. In addition to new promotions pathways and alternative evidence for the impact of scholarship, other approaches to fostering health systems innovation at AHCs include more robust funding for career development in health systems innovation, new curricula to enable trainees to develop skills in health systems innovation, and new ways for innovators to disseminate their work. AHCs that foster health systems innovation could meet a critical need to contribute both to the sustainability of our health care system and to AHCs' continued leadership role within it.

  12. Health Systems Innovation at Academic Health Centers: Leading in a New Era of Health Care Delivery.

    PubMed

    Ellner, Andrew L; Stout, Somava; Sullivan, Erin E; Griffiths, Elizabeth P; Mountjoy, Ashlin; Phillips, Russell S

    2015-07-01

    Challenged by demands to reduce costs and improve service delivery, the U.S. health care system requires transformational change. Health systems innovation is defined broadly as novel ideas, products, services, and processes-including new ways to promote healthy behaviors and better integrate health services with public health and other social services-which achieve better health outcomes and/or patient experience at equal or lower cost. Academic health centers (AHCs) have an opportunity to focus their considerable influence and expertise on health systems innovation to create new approaches to service delivery and to nurture leaders of transformation. AHCs have traditionally used their promotions criteria to signal their values; creating a health systems innovator promotion track could be a critical step towards creating opportunities for innovators in academic medicine. In this Perspective, the authors review publicly available promotions materials at top-ranked medical schools and find that while criteria for advancement increasingly recognize systems innovation, there is a lack of specificity on metrics beyond the traditional yardstick of peer-reviewed publications. In addition to new promotions pathways and alternative evidence for the impact of scholarship, other approaches to fostering health systems innovation at AHCs include more robust funding for career development in health systems innovation, new curricula to enable trainees to develop skills in health systems innovation, and new ways for innovators to disseminate their work. AHCs that foster health systems innovation could meet a critical need to contribute both to the sustainability of our health care system and to AHCs' continued leadership role within it. PMID:25738387

  13. Remote Underwater Characterization System - Innovative Technology Summary Report

    SciTech Connect

    Willis, W D

    1999-04-01

    Characterization and inspection of water-cooled and moderated nuclear reactors and fuel storage pools requires equipment capable of operating underwater. Similarly, the deactivation and decommissioning of older nuclear facilities often requires the facility owner to accurately characterize underwater structures and equipment which may have been sitting idle for years. The Remote Underwater Characterization System (RUCS) is a small, remotely operated submersible vehicle intended to serve multiple purposes in underwater nuclear operations. It is based on the commercially-available "Scallop" vehicle 1 , but has been modified by the Department of Energy's Robotics Technology Development Program to add auto-depth control, and vehicle orientation and depth monitoring at the operator control panel. The RUCS is designed to provide visual and gamma radiation characterization, even in confined or limited access areas. It was demonstrated in August 1998 at the Idaho National Engineering and Environmental Laboratory (INEEL) as part of the INEEL Large Scale Demonstration and Deployment Project. During the demonstration it was compared in a "head-to-head" fashion with the baseline characterization technology. This paper summarizes the results of the demonstration and lessons learned; comparing and contrasting both technologies in the areas of cost, visual characterization, radiological characterization, and overall operations.

  14. Reactor system for conversion of alcohols to ether-rich gasoline

    SciTech Connect

    Harandi, M.N.; Owen, H.

    1992-07-14

    This patent describes a continuous feedstock separation and etherification reactor system for converting crude methanol feedstock to methyl t-alkyl ether. It comprises extractor means; first catalytic reactor means operatively connected; effluent separation means; and second catalytic reactor means operatively connected, means for recovering and charging at least a portion of the C{sub 4}-C{sub 5} olefinic hydrocarbons rich in iso-olefin from the second reactor means to the extractor means as the liquid-olefinic hydrocarbon extraction stream.

  15. An innovative forecasting and dashboard system for Malaysian dengue trends

    NASA Astrophysics Data System (ADS)

    Jamil, Jastini Mohd; Shaharanee, Izwan Nizal Mohd

    2016-08-01

    Dengue fever has been recognized in over 100 countries and 2.5 billion people live in areas where dengue is endemic. It is currently a serious arthropod-borne disease, affecting around 50-100 million people worldwide every year. Dengue fever is also prevalent in Malaysia with numerous cases including mortality recorded over the past year. In 2012, a total of 21,900 cases of dengue fever were reported with 35 deaths. Dengue, a mosquito-transmitted virus, causes a high fever accompanied by significant pain in afflicted patient and the Aedes Aegypti mosquito is the primary disease carrier. Knowing the dangerous effect of dengue fever, thus one of the solutions is to implement an innovative forecasting and dashboard system of dengue spread in Malaysia, with emphasize on an early prediction of dengue outbreak. Specifically, the model developed will provide with a valuable insight into strategically managing and controlling the future dengue epidemic. Importantly, this research will deliver the message to health policy makers such as The Ministry of Health Malaysia (MOH), practitioners, and researchers of the importance to integrate their collaboration in exploring the potential strategies in order to reduce the future burden of the increase in dengue transmission cases in Malaysia.

  16. Development of high-fidelity multiphysics system for light water reactor analysis

    NASA Astrophysics Data System (ADS)

    Magedanz, Jeffrey W.

    There has been a tendency in recent years toward greater heterogeneity in reactor cores, due to the use of mixed-oxide (MOX) fuel, burnable absorbers, and longer cycles with consequently higher fuel burnup. The resulting asymmetry of the neutron flux and energy spectrum between regions with different compositions causes a need to account for the directional dependence of the neutron flux, instead of the traditional diffusion approximation. Furthermore, the presence of both MOX and high-burnup fuel in the core increases the complexity of the heat conduction. The heat transfer properties of the fuel pellet change with irradiation, and the thermal and mechanical expansion of the pellet and cladding strongly affect the size of the gap between them, and its consequent thermal resistance. These operational tendencies require higher fidelity multi-physics modeling capabilities, and this need is addressed by the developments performed within this PhD research. The dissertation describes the development of a High-Fidelity Multi-Physics System for Light Water Reactor Analysis. It consists of three coupled codes -- CTF for Thermal Hydraulics, TORT-TD for Neutron Kinetics, and FRAPTRAN for Fuel Performance. It is meant to address these modeling challenges in three ways: (1) by resolving the state of the system at the level of each fuel pin, rather than homogenizing entire fuel assemblies, (2) by using the multi-group Discrete Ordinates method to account for the directional dependence of the neutron flux, and (3) by using a fuel-performance code, rather than a Thermal Hydraulics code's simplified fuel model, to account for the material behavior of the fuel and its feedback to the hydraulic and neutronic behavior of the system. While the first two are improvements, the third, the use of a fuel-performance code for feedback, constitutes an innovation in this PhD project. Also important to this work is the manner in which such coupling is written. While coupling involves combining

  17. Passive cooling system for top entry liquid metal cooled nuclear reactors

    DOEpatents

    Boardman, Charles E.; Hunsbedt, Anstein; Hui, Marvin M.

    1992-01-01

    A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

  18. TESTING AND PERFORMANCE EVALUATION OF AN INNOVATIVE INTERNAL PIPE SEALING SYSTEM FOR WASTEWATER MAIN REHABILITATION

    EPA Science Inventory

    Many utilities are seeking emerging and innovative rehabilitation technologies to extend the service life of their infrastructure systems. This report describes the testing and performance evaluation of an internal pipe sealing system, which provides a permanent physical seal fo...

  19. Innovative Distributed Power Grid Interconnection and Control Systems: Final Report, December 11, 2000 - August 30, 2005

    SciTech Connect

    DePodesta, K.; Birlingmair, D.; West, R.

    2006-03-01

    The contract goal was to further advance distributed generation in the marketplace by making installations more cost-effective and compatible with existing systems. This was achieved by developing innovative grid interconnection and control systems.

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

    SciTech Connect

    Dobranich, D.

    1992-07-01

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

  1. Building America Top Innovations 2012: Basement Insulation Systems

    SciTech Connect

    none,

    2013-01-01

    This Building America Top Innovations profile describes research on basement insulation, which identifies the wall installation methods and materials that perform best in terms of insulation and water resistance.

  2. Improvement of Algorithms for Pressure Maintenance Systems in Drum-Separators of RBMK-1000 Reactors

    SciTech Connect

    Aleksakov, A. N. Yankovskiy, K. I.; Dunaev, V. I.; Kushbasov, A. N.

    2015-05-15

    The main tasks and challenges for pressure regulation in the drum-separators of RBMK-1000 reactors are described. New approaches to constructing algorithms for pressure control in drum-separators by electro-hydraulic turbine control systems are discussed. Results are provided from tests of the operation of modernized pressure regulators during fast transients with reductions in reactor power.

  3. Design of GA thermochemical water-splitting process for the Mirror Advanced Reactor System

    SciTech Connect

    Brown, L.C.

    1983-04-01

    GA interfaced the sulfur-iodine thermochemical water-splitting cycle to the Mirror Advanced Reactor System (MARS). The results of this effort follow as one section and part of a second section to be included in the MARS final report. This section describes the process and its interface to the reactor. The capital and operating costs for the hydrogen plant are described.

  4. System Requirements Document for the Molten Salt Reactor Experiment

    SciTech Connect

    Aigner, R.D.

    2000-04-01

    The purpose of the conversion process is to convert the {sup 233}U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019.

  5. Liquid metal systems development: reactor vessel support structure evaluation

    SciTech Connect

    McEdwards, J.A.

    1981-01-01

    Results of an evaluation of support structures for the reactor vessel are reported. The U ring, box ring, integral ring, tee ring and tangential beam supports were investigated. The U ring is the recommended vessel support structure configuration.

  6. Agricultural Innovation Systems (AIS): A Study of Stakeholders and Their Relations in System of Rice Intensification (SRI)

    ERIC Educational Resources Information Center

    Suchiradipta, Bhattacharjee; Raj, Saravanan

    2015-01-01

    Purpose: This paper identifies the stakeholders of System of Rice Intensification (SRI), their roles and actions and the supporting and enabling environment of innovation in the state as the elements of the Agricultural Innovation Systems (AIS) in SRI in Tripura state of India and studies the relationship matrix among the stakeholders.…

  7. The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World's First Radioactive Fusion Research Reactor for Burial

    SciTech Connect

    Keith Rule; Erik Perry; Jim Chrzanowski; Mike Viola; Ron Strykowsky

    2003-09-15

    Original estimates stated that the amount of radioactive waste that will be generated during the dismantling of the Tokamak Fusion Test Reactor will approach two million kilograms with an associated volume of 2,500 cubic meters. The materials were activated by 14 MeV neutrons and were highly contaminated with tritium, which present unique challenges to maintain integrity during packaging and transportation. In addition, the majority of this material is stainless steel and copper structural metal that were specifically designed and manufactured for this one-of-a-kind fusion research reactor. This provided further complexity in planning and managing the waste. We will discuss the engineering concepts, innovative practices, and technologies that were utilized to size reduce, stabilize, and package the many unique and complex components of this reactor. This waste was packaged and shipped in many different configurations and methods according to the transportation regulations and disposal facility requirements. For this particular project, we were able to utilize two separate disposal facilities for burial. This paper will conclude with a complete summary of the actual results of the waste management costs, volumes, and best practices that were developed from this groundbreaking and successful project.

  8. Hybrid energy systems (HESs) using small modular reactors (SMRs)

    SciTech Connect

    S. Bragg-Sitton

    2014-10-01

    Large-scale nuclear reactors are traditionally operated for a singular purpose: steady-state production of dispatchable baseload electricity that is distributed broadly on the electric grid. While this implementation is key to a sustainable, reliable energy grid, small modular reactors (SMRs) offer new opportunities for increased use of clean nuclear energy for both electric and thermal ap plications in more locations – while still accommodating the desire to support renewable production sources.

  9. Using an Interactive Systems Framework to Expand Telepsychology Innovations in Underserved Communities

    PubMed Central

    Alaniz, Angela B.

    2016-01-01

    Literature indicates that the use of promising innovations in mental health care can be improved. The advancement of telepsychology is one innovation that has been utilized as a method to reduce rural health disparities and increase the number of people with access to mental health services. This paper describes a successful pilot telepsychology program implemented in a rural community to increase access to mental health services and the model's replication and expansion into four additional communities using concepts described in an Interactive Systems Framework. The Interactive Systems Framework highlights how building local capacity specific to organizational functioning and innovations are necessary to support, deliver, and disseminate innovations within new settings. Based on the knowledge gained from this telepsychology innovation, the application of an Interactive Systems Framework and funding mechanisms are discussed. PMID:27403374

  10. 75 FR 7464 - Energy Efficient Building Systems Regional Innovation Cluster Initiative-Joint Federal Funding...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-19

    ... Announcement Information Session AGENCY: Department of Energy (DOE). ACTION: Notice of public meeting and... innovation in energy efficient building technologies and systems design. The DOE funded Energy Efficient... (the ``Energy Regional Innovation Cluster'' or ``E-RIC'') and will work to disseminate new...

  11. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    SciTech Connect

    Rohanda, Anis; Waris, Abdul

    2015-04-16

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  12. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    NASA Astrophysics Data System (ADS)

    Rohanda, Anis; Waris, Abdul

    2015-04-01

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on 16O(n,p)16N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  13. Remote Underwater Characterization System - Innovative Technology Summary Report

    SciTech Connect

    Willis, Walter David

    1999-04-01

    Characterization and inspection of water-cooled and moderated nuclear reactors and fuel storage pools requires equipment capable of operating underwater. Similarly, the deactivation and decommissioning of older nuclear facilities often requires the facility owner to accurately characterize underwater structures and equipment which may have been sitting idle for years. The underwater characterization equipment is often required to operate at depths exceeding 20 ft (6.1 m) and in relatively confined or congested spaces. The typical baseline approach has been the use of radiation detectors and underwater cameras mounted on long poles, or stationary cameras with pan and tilt features mounted on the sides of the underwater facility. There is a perceived need for an inexpensive, more mobile method of performing close-up inspection and radiation measurements in confined spaces underwater. The Remote Underwater Characterization System (RUCS) is a small, remotely operated submersible vehicle intended to serve multiple purposes in underwater nuclear operations. It is based on the commercially-available “Scallop” vehicle1, but has been modified by Department of Energy’s Robotics Technology Development Program to add auto-depth control, and vehicle orientation and depth monitoring at the operator control panel. The RUCS is designed to provide visual and gamma radiation characterization, even in confined or limited access areas. It was demonstrated in August 1998 at Idaho National Engineering and environmental Laboratory (INEEL) as part of the INEEL Large Scale Demonstration and Deployment Project. During the demonstration it was compared in a “head-tohead” fashion with the baseline characterization technology. This paper summarizes the results of the demonstration and lessons learned; comparing and contrasting both technologies in the areas of cost, visual characterization, radiological characterization, and overall operations.

  14. Innovative systems for sustainable nuclear energy generation and waste management

    NASA Astrophysics Data System (ADS)

    Loiseaux, Jm; David, S.

    2006-05-01

    The limited amount of fossil resources, the impact of green-house gas emissions on the world climate, the rising demand of primary energy projected to 2050, lead to a potentially critical situation for the world energy supply. The need for alternative (to fossil energies) massive energy production is evaluated to 10 Gtoe. The potential of Nuclear Energy generation at the level of 5 Gtoe is examined. Such a sustainable production can only be met by a breeder reactor fleet for which a deployment scenario is described with the associated constraints. Waste management is discussed in connection with different nuclear energy development scenarios according to the point in time when breeder reactors are started. At the world level, it appears that the optimal handling of today's wastes rests on an early decision to develop tomorrow's breeder reactors.

  15. Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor

    DOEpatents

    Heinold, Mark R.; Berger, John F.; Loper, Milton H.; Runkle, Gary A.

    2015-12-29

    Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

  16. Encouraging innovation.

    PubMed

    Hyman, Anthony A

    2014-02-01

    Innovation is central to the scientific endeavor, and yet the current system of funding in the United States discourages innovation, especially in the young. Subtle alterations to the funding system, guided in part by the success of the European Research Council, could have major effects on encouraging innovation.

  17. Physical modelling of the composting environment: A review. Part 1: Reactor systems

    SciTech Connect

    Mason, I.G. . E-mail: ian.mason@canterbury.ac.nz; Milke, M.W.

    2005-07-01

    In this paper, laboratory- and pilot-scale reactors used for investigation of the composting process are described and their characteristics and application reviewed. Reactor types were categorised by the present authors as fixed-temperature, self-heating, controlled temperature difference and controlled heat flux, depending upon the means of management of heat flux through vessel walls. The review indicated that fixed-temperature reactors have significant applications in studying reaction rates and other phenomena, but may self-heat to higher temperatures during the process. Self-heating laboratory-scale reactors, although inexpensive and uncomplicated, were shown to typically suffer from disproportionately large losses through the walls, even with substantial insulation present. At pilot scale, however, even moderately insulated self-heating reactors are able to reproduce wall losses similar to those reported for full-scale systems, and a simple technique for estimation of insulation requirements for self-heating reactors is presented. In contrast, controlled temperature difference and controlled heat flux laboratory reactors can provide spatial temperature differentials similar to those in full-scale systems, and can simulate full-scale wall losses. Surface area to volume ratios, a significant factor in terms of heat loss through vessel walls, were estimated by the present authors at 5.0-88.0 m{sup 2}/m{sup 3} for experimental composting reactors and 0.4-3.8 m{sup 2}/m{sup 3} for full-scale systems. Non-thermodynamic factors such as compression, sidewall airflow effects, channelling and mixing may affect simulation performance and are discussed. Further work to investigate wall effects in composting reactors, to obtain more data on horizontal temperature profiles and rates of biological heat production, to incorporate compressive effects into experimental reactors and to investigate experimental systems employing natural ventilation is suggested.

  18. Sliding mode control of the space nuclear reactor system TOPAZ II

    SciTech Connect

    Shtessel, Y.B.; Wyant, F.J.

    1996-03-01

    The Automatic Control System (ACS) of the space nuclear reactor power system TOPAZ II that generates electricity from nuclear heat using in-core thermionic converters is considered. Sliding Mode Control Technique was applied to the reactor system controller design in order to provide the robust high accuracy following of a neutron (thermal) power reference profile in a start up regime and a payload electric power (current) reference profile following in an operation regime. Extensive simulations of the TOPAZ II reactor system with the designed sliding mode controllers showed improved accuracy and robustness of the reactor system performances in a start up regime and in an electric power supply regime as well. {copyright} {ital 1996 American Institute of Physics.}

  19. Hanging core support system for a nuclear reactor

    DOEpatents

    Burelbach, James P.; Kann, William J.; Pan, Yen-Cheng; Saiveau, James G.; Seidensticker, Ralph W.

    1987-01-01

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform. Motion or radiation sensing detectors can be provide at the lower ends of the tension rods for obtaining pertinent readings proximate the core.

  20. Radiogenic lead with dominant content of {sup 208}Pb: New coolant and neutron moderator for innovative nuclear reactors

    SciTech Connect

    Shmelev, A. N.; Kulikov, G. G.; Kryuchkov, E. F.; Apse, V. A.; Kulikov, E. G.

    2012-07-01

    The advantages of radiogenic lead with dominant content of {sup 208}Pb as a reactor coolant with respect to natural lead are caused by unique nuclear properties of {sup 208}Pb which is a double-magic nucleus with closed proton and neutron shells. This results in significantly lower micro cross section and resonance integral of radiative neutron capture by {sup 208}Pb than those for numerous light neutron moderators. The extremely weak ability of {sup 208}Pb to absorb neutrons results in the following effects. Firstly, neutron moderating factor (ratio of scattering to capture cross sections) is larger than that for graphite and light water. Secondly, age and diffusion length of thermal neutrons are larger than those for graphite, light and heavy water. Thirdly, neutron lifetime in {sup 208}Pb is comparable with that for graphite, beryllium and heavy water what could be important for safe reactor operation. The paper presents some results obtained in neutronics and thermal-hydraulics evaluations of the benefits from the use of radiogenic lead with dominant content of {sup 208}Pb instead of natural lead as a coolant of fast breeder reactors. The paper demonstrates that substitution of radiogenic lead for natural lead can offer the following benefits for operation of fast breeder reactors. Firstly, improvement of the reactor safety thanks to the better values of coolant temperature reactivity coefficient and, secondly, improvement of some thermal-hydraulic reactor parameters. Radiogenic lead can be extracted from thorium sludge without isotope separation as {sup 208}Pb is a final isotope in the decay chain of {sup 232}Th. (authors)

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

    NASA Technical Reports Server (NTRS)

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

    1974-01-01

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

  2. Investigation of Anaerobic Fluidized Bed Reactor/ Aerobic Moving Bed Bio Reactor (AFBR/MMBR) System for Treatment of Currant Wastewater

    PubMed Central

    JAFARI, Jalil; MESDAGHINIA, Alireza; NABIZADEH, Ramin; FARROKHI, Mehrdad; MAHVI, Amir Hossein

    2013-01-01

    Background: Anaerobic treatment methods are more suitable for the treatment of concentrated wastewater streams, offer lower operating costs, the production of usable biogas product. The aim of this study was to investigate the performance of an Anaerobic Fluidized Bed Reactor (AFBR)-Aerobic Moving Bed Bio Reactor (MBBR) in series arrangement to treat Currant wastewater. Methods: The bed materials of AFBR were cylindrical particles made of PVC with a diameter of 2–2.3 mm, particle density of 1250 kg/m3. The volume of all bed materials was 1.7 liter which expanded to 2.46 liters in fluidized situation. In MBBR, support media was composed of 1.5 liters Bee-Cell 2000 having porosity of 87% and specific surface area of 650m2/m3. Results: When system operated at 35 ºC, chemical oxygen demand (COD) removal efficiencies were achieved to 98% and 81.6% for organic loading rates (OLR) of 9.4 and 24.2 g COD/l.d, and hydraulic retention times (HRT) of 48 and 18 h, in average COD concentration feeding of 18.4 g/l, respectively. Conclusion: The contribution of AFBR in total COD removal efficiency at an organic loading rate (OLR) of 9.4 g COD/l.d was 95%, and gradually decreased to 76.5% in OLR of 24.2 g COD/l.d. Also with increasing in organic loading rate the contribution of aerobic reactor in removing COD gradually decreased. In this system, the anaerobic reactor played the most important role in the removal of COD, and the aerobic MBBR was actually needed to polish the anaerobic treated wastewater. PMID:26056640

  3. Physics Characterization of a Heterogeneous Sodium Fast Reactor Transmutation System

    SciTech Connect

    Samuel 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 mass 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 a non-flattened and a pancake core geometry. 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 the same size.

  4. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    SciTech Connect

    Qualls, A L; Cetiner, M S; Wilson, Jr, T L

    2012-04-30

    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

  5. 77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-12

    ... COMMISSION Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES:...

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

    SciTech Connect

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

    1984-01-01

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

  7. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  8. Passive decay heat removal system for water-cooled nuclear reactors

    DOEpatents

    Forsberg, Charles W.

    1991-01-01

    A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

  9. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

    1994-05-03

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  10. Progress in space nuclear reactor power systems technology development - The SP-100 program

    NASA Technical Reports Server (NTRS)

    Davis, H. S.

    1984-01-01

    Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

  11. Sodium leak detection system for liquid metal cooled nuclear reactors

    DOEpatents

    Modarres, Dariush

    1991-01-01

    A light source is projected across the gap between the containment vessel and the reactor vessel. The reflected light is then analyzed with an absorption spectrometer. The presence of any sodium vapor along the optical path results in a change of the optical transmissivity of the media. Since the absorption spectrum of sodium is well known, the light source is chosen such that the sensor is responsive only to the presence of sodium molecules. The optical sensor is designed to be small and require a minimum of amount of change to the reactor containment vessel.

  12. Nuclear reactor descriptions for space power systems analysis

    NASA Technical Reports Server (NTRS)

    Mccauley, E. W.; Brown, N. J.

    1972-01-01

    For the small, high performance reactors required for space electric applications, adequate neutronic analysis is of crucial importance, but in terms of computational time consumed, nuclear calculations probably yield the least amount of detail for mission analysis study. It has been found possible, after generation of only a few designs of a reactor family in elaborate thermomechanical and nuclear detail to use simple curve fitting techniques to assure desired neutronic performance while still performing the thermomechanical analysis in explicit detail. The resulting speed-up in computation time permits a broad detailed examination of constraints by the mission analyst.

  13. System Thermal Model for the S-Prime Thermionic Reactor

    NASA Astrophysics Data System (ADS)

    Arx, Alan V. Von

    1994-07-01

    A model has been developed which numerically simulates heat transfer and flow characteristics of the thermal-hydraulic loop of the S-PRIME thermionic reactor. The components for which detailed models have been included are: the thermionic fuel elements (TFEs), heat pipe panels, flow loop and pumps. The reactor start-up operation was then modeled from zero to full power. It includes modelling of the melting of the heat pipe working fluid as well as correlations for the performance of the thermionic cells. The results show that there is stable operation during this period.

  14. Tritium Formation and Mitigation in High-Temperature Reactor Systems

    SciTech Connect

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  15. From Professional Development to System Change: Teacher Leadership and Innovation

    ERIC Educational Resources Information Center

    Frost, David

    2012-01-01

    This article argues for a particular conceptualisation of teacher leadership; one that enables us to overcome the limitations of established approaches to continuing professional development as a strategy for school improvement by mobilising the massive untapped potential of teachers as leaders of innovation. In this article, teacher leadership is…

  16. Building America Top Innovations 2012: Advanced Framing Systems and Packages

    SciTech Connect

    none,

    2013-01-01

    This Building America Top Innovations profile describes Building America research showing advanced 2x6, 24-inch on-center framing, single top plates, open headers, and 2-stud corners reduced board feet of lumber by more than 1,000 feet, cut energy use by 13%, and cut material and labor costs by more than $1,000 on a typical home.

  17. Demonstration of Innovative Sewer System Inspection Technology SewerBatt

    EPA Science Inventory

    The overall objective of this EPA-funded study was to demonstrate innovative a sewer line assessment technology that is designed for rapid deployment using portable equipment. This study focused on demonstration of a technology that is suitable for smaller diameter pipes (less th...

  18. Demonstration of Innovative Sewer System Inspection Technology: SL-RAT

    EPA Science Inventory

    The overall objective of this EPA-funded study was to demonstrate innovative sewer line assessment technologies that are designed for rapid deployment using portable equipment. This study focused on demonstration of technologies that are suitable for smaller diameter pipes (less ...

  19. Development of a novel integrated continuous reactor system for biocatalytic production of biodiesel.

    PubMed

    Chattopadhyay, Soham; Sen, Ramkrishna

    2013-11-01

    A novel integrated immobilized enzyme-reactor system involving a continuous stirred tank reactor with two packed bed reactors in series was developed for the continuous production of biodiesel. The problem of methanol solubility into oil was solved by introducing a stirred tank reactor to dissolve methanol into partially converted oil. This step made the process perfectly continuous without requiring any organic solvent and intermittent methanol addition in the process. The substrate feeding rate of 0.74 mL/min and enzyme loading of 0.75 g per reactor were determined to be optimum for maximum biodiesel yield. The integrated continuous process was stable up to 45 cycles with biodiesel productivity of 137.2 g/L/h, which was approximately 5 times higher than solvent free batch process. In comparison with the processes reported in literature using expensive Novozyme 435 and hazardous organic solvent, the present process is completely green and perfectly continuous with economic and environmental advantages.

  20. IN SITU STEAM ENHANCED RECOVERY PROCESS - HUGHES ENVIRONMENTAL SYSTEMS, INC. - INNOVATIVE TECHNOLOGY EVALUATION REPORT

    EPA Science Inventory

    This Innovative Technology Evaluation report summarizes the findings of an evaluation of the in situ Steam Enhanced Recovery Process (SERP) operated by Hughes Environmental Systems, Inc. at the Rainbow Disposal facility in Huntington Beach, California. he technology demonstration...

  1. Innovation of laboratory exercises in course Distributed systems and computer networks

    NASA Astrophysics Data System (ADS)

    Souček, Pavel; Slavata, Oldřich; Holub, Jan

    2013-09-01

    This paper is focused on innovation of laboratory exercises in course Distributed Systems and Computer Networks. These exercises were introduced in November of 2012 and replaced older exercises in order to reflect real life applications.

  2. Building America Top Innovations 2013 Profile – Quality Management System Guidelines

    SciTech Connect

    none,

    2013-09-01

    This Top Innovation profile describes quality management system tools that were customized for residential construction by BSC, IBACOS, and PHI, for use by builders, trades, and designers to help eliminate mistakes that would require high-cost rework.

  3. Building America Top Innovations 2013 Profile – HPXML: A Standardized Home Performance Data Sharing System

    SciTech Connect

    none,

    2013-09-01

    This Top Innovation profile describes the Standard for Home Performance-Related Data Transfer (known as HPXML), developed by the National Renewable Energy Laboratory, which facilitates smooth communication between program tracking systems and energy upgrade analysis software,

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

    NASA Astrophysics Data System (ADS)

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

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

  5. Heat transfer in inertial confinement fusion reactor systems

    SciTech Connect

    Hovingh, J.

    1980-04-23

    The short time and deposition distance for the energy from inertial fusion products results in local peak power densities on the order of 10/sup 18/ watts/m/sup 3/. This paper presents an overview of the various inertial fusion reactor designs which attempt to reduce these peak power intensities and describes the heat transfer considerations for each design.

  6. Heat insulating system for a fast reactor shield slab

    DOEpatents

    Kotora, J. Jr.; Groh, E.F.; Kann, W.J.; Burelbach, J.P.

    1984-04-10

    Improved thermal insulation for a nuclear reactor deck comprises many helical coil springs disposed in generally parallel, side-by-side laterally overlapping or interfitted relationship to one another so as to define a three-dimensional composite having both metal and voids between the metal, and enclosure means for holding the composite to the underside of the deck.

  7. 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.

  8. Radionuclide inventories for short run-time space nuclear reactor systems

    SciTech Connect

    Coats, R.L.

    1992-10-22

    Space Nuclear Reactor Systems, especially those used for propulsion, often have expected operation run times much shorter than those for land-based nuclear power plants. This produces substantially different radionuclide inventories to be considered in the safety analyses of space nuclear systems. This presentation describes an analysis utilizing ORIGEN2 and DKPOWER to provide comparisons among representative land-based and space systems. These comparisons enable early, conceptual considerations of safety issues and features in the preliminary design phases of operational systems, test facilities, and operations by identifying differences between the requirements for space systems and the established practice for land-based power systems. Early indications are that separation distance is much more effective as a safety measure for space nuclear systems than for power reactors because greater decay of the radionuclide activity occurs during the time to transport the inventory a given distance. In addition, the inventories of long-lived actinides are very low for space reactor systems.

  9. Anaerobic digestion of ice-cream wastewater: A comparison of single and two-phase reactor systems

    SciTech Connect

    Borja, R.; Banks, C.J.

    1995-03-01

    The anaerobic digestion of ice-cream wastewater, a complex substrate which includes milk proteins, carbohydrates, and lipids, has received little attention. Work using an aerobic contact system showed that at a 7.5-d hydraulic retention time (HRT), with an organic loading rate of 1.7 g COD/Ld and influent TSS (total suspended solids) of 5870 mg/L, the effluent COD was 628 mg/L, BOD was 91 mg/L and TSS was 674. Anaerobic filters have also been used at organic loadings of 6 kg COD/m{sup 3}d applied at a HRT of 0.42 day, with COD removals of 80%. Goodwing showed that this waste was capable of being treated by the UASB process with granulation commencing after 60-70 days, and gas production ranging between 0.73 and 0.93 L CH{sub 4}/g COD removed with loading rates between 0.7 and 3.0 g TOC/Ld. Two-phase anaerobic digestion is an innovative fermentation mode that has recently received increased attention. The kinetically dissimilar fermentation phases, hydrolysis-acidification and acetogenesis-methanation are operated in two separate reactors; the first of which is maintained at a very short HRT. The effluent from the first, acid-forming, phase is used as the substrate for the methane-phase reactor which has a longer HRT or cell immobilization. The aim of this work was to compare the methane production capability and performance of a single-phase upflow fixed bed reactor with a two-phase digestion system. The two-phase digestion system consists of a completely mixed reactor for the acidogenic reaction and an upflow fixed bed reactor for the methanogenic reaction. Because of the high lipid content and COD of ice cream wastewater off site disposal has proved to be both expensive and poses problems to the receiving effluent treatment plant. For this reason the potential for a rapid anaerobic stabilization of the waste, with energy recovery in the form of methane gas, has been investigated in an attempt to minimize plant size and maximize gas production. 9 refs., 2 tabs.

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

    NASA Astrophysics Data System (ADS)

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

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

  11. Hydraulic characterization of an activated sludge reactor with recycling system by tracer experiment and analytical models.

    PubMed

    Sánchez, F; Viedma, A; Kaiser, A S

    2016-09-15

    Fluid dynamic behaviour plays an important role in wastewater treatment. An efficient treatment requires the inexistence of certain hydraulic problems such as dead zones or short-circuiting flows. Residence time distribution (RTD) analysis is an excellent technique for detecting these inefficiencies. However, many wastewater treatment installations include water or sludge recycling systems, which prevent us from carrying out a conventional tracer pulse experiment to obtain the RTD curve of the installation. This paper develops an RTD analysis of an activated sludge reactor with recycling system. A tracer experiment in the reactor is carried out. Three analytical models, derived from the conventional pulse model, are proposed to obtain the RTD curve of the reactor. An analysis of the results is made, studying which model is the most suitable for each situation. This paper is useful to analyse the hydraulic efficiency of reactors with recycling systems. PMID:27288672

  12. Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

    SciTech Connect

    Sweeney, F.J. ); Carroll, D.G. ); Chen, C. ); Crane, C.; Dalton, R. ); Taylor, J.R. ); Tosunoglu, S. )

    1993-01-01

    One of the most important safety systems in General Electric's (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE/NE Robotics for Advanced Reactors program formed a team to evaluate the ALMR design for remote inspection of the RVACS. Conceptual designs for robots to perform the required inspection tasks were developed by the team. Design criteria for these remote systems included robot deployment, power supply, navigation, environmental hardening of components, tether management, communication with an operator, sensing, and failure recovery. The operation of the remote inspection concepts were tested using 3-D simulation models of the ALMR. In addition, the team performed an extensive technology review of robot components that could survive the environmental conditions in the RVACS.

  13. Hydraulic characterization of an activated sludge reactor with recycling system by tracer experiment and analytical models.

    PubMed

    Sánchez, F; Viedma, A; Kaiser, A S

    2016-09-15

    Fluid dynamic behaviour plays an important role in wastewater treatment. An efficient treatment requires the inexistence of certain hydraulic problems such as dead zones or short-circuiting flows. Residence time distribution (RTD) analysis is an excellent technique for detecting these inefficiencies. However, many wastewater treatment installations include water or sludge recycling systems, which prevent us from carrying out a conventional tracer pulse experiment to obtain the RTD curve of the installation. This paper develops an RTD analysis of an activated sludge reactor with recycling system. A tracer experiment in the reactor is carried out. Three analytical models, derived from the conventional pulse model, are proposed to obtain the RTD curve of the reactor. An analysis of the results is made, studying which model is the most suitable for each situation. This paper is useful to analyse the hydraulic efficiency of reactors with recycling systems.

  14. A coupled nuclear reactor thermal energy storage system for enhanced load following operation

    NASA Astrophysics Data System (ADS)

    Alameri, Saeed A.

    Nuclear power plants usually provide base-load electric power and operate most economically at a constant power level. In an energy grid with a high fraction of renewable energy sources, future nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling the reactor to a large Thermal Energy Storage (TES) block will allow the reactor to better respond to variable power demands. In the system described in this thesis, a Prismatic-core Advanced High Temperature Reactor (PAHTR) operates at constant power with heat provided to a TES block that supplies power as needed to a secondary energy conversion system. The PAHTR is designed to have a power rating of 300 MW th, with 19.75 wt% enriched Tri-Structural-Isotropic UO 2 fuel and a five year operating cycle. The passive molten salt TES system will operate in the latent heat region with an energy storage capacity of 150 MWd. Multiple smaller TES blocks are used instead of one large block to enhance the efficiency and maintenance complexity of the system. A transient model of the coupled reactor/TES system is developed to study the behavior of the system in response to varying load demands. The model uses six-delayed group point kinetics and decay heat models coupled to thermal-hydraulic and heat transfer models of the reactor and TES system. Based on the transient results, the preferred TES design consists of 1000 blocks, each containing 11000 LiCl phase change material tubes. A safety assessment of major reactor events demonstrates the inherent safety of the coupled system. The loss of forced circulation study determined the minimum required air convection heat removal rate from the reactor core and the lowest possible reduced primary flow rate that can maintain the reactor in a safe condition. The loss of ultimate heat sink study demonstrated the ability of the TES

  15. Reforming America's health system through innovation and entrepreneurship.

    PubMed

    Reece, Richard L

    2005-01-01

    America's attempts for healthcare reform are gridlocked. Healthcare special interests are reluctant to abandon profitable activities, and American culture-distrust of centralized federal power, belief in self-improvement, desire for choice, and belief in equal access to medical technologies-is slow to change. Physician entrepreneurship and innovation, coupled with consumer-driven healthcare and public-private partnerships, may break the present gridlock.

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

    DOEpatents

    Hunsbedt, Anstein

    1996-01-01

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

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

    DOEpatents

    Hunsbedt, A.

    1996-03-12

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

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

    SciTech Connect

    Dobranich, D.

    1992-07-01

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

  19. Biopharmaceutical Innovation System in China: System Evolution and Policy Transitions (Pre-1990s-2010s)

    PubMed Central

    Hu, Hao; Chung, Chao-Chen

    2015-01-01

    Background: This article sets up the initial discussion of the evolution of biopharmaceutical innovation in China through the perspective of sectoral innovation system (SIS). Methods: Two data sources including archival documentary data and field interviews were used in this study. Archival documentary data was collected from China Food and Drug Administration (CFDA) and Chinese National Knowledge Infrastructure (CNKI). In addition, industrial practitioners and leading researchers in academia were interviewed. Results: Biopharmaceutical in China was established through international knowledge transfer. The firms played more active role in commercializing biopharmaceutical in China though universities and research institutes were starting to interact with local firms and make contribution to biopharmaceutical industrialization. The transition of the Chinese government’s policies continuously shapes the evolution of biopharmaceutical sector. Policies have been dramatic changes before and after 1980s to encourage developing biopharmaceutical as a competitive industry for China. Conclusion: A SIS for biopharmaceutical has been shaped in China. However, currently biopharmaceutical is still a small sector in China, and for the further growth of the industry more synthetic policies should be implemented. Not only the policy supports towards the research and innovation of biopharmaceuticals in the early stage of development should be attended, but also commercialization of biopharmaceutical products in the later stage of sales. PMID:26673466

  20. Dual annular rotating [open quotes]windowed[close quotes] nuclear reflector reactor control system

    DOEpatents

    Jacox, M.G.; Drexler, R.L.; Hunt, R.N.M.; Lake, J.A.

    1994-03-29

    A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core. 4 figures.

  1. Sodium coolant purification systems for a nuclear power station equipped with a BN-1200 reactor

    NASA Astrophysics Data System (ADS)

    Alekseev, V. V.; Kovalev, Yu. P.; Kalyakin, S. G.; Kozlov, F. A.; Kumaev, V. Ya.; Kondrat'ev, A. S.; Matyukhin, V. V.; Pirogov, E. P.; Sergeev, G. P.; Sorokin, A. P.; Torbenkova, I. Yu.

    2013-05-01

    Both traditional coolant purification methods (by means of traps and sorbents for removing cesium), the use of which supported successful operation of nuclear power installations equipped with fast-neutron reactors with a sodium coolant, and the possibility of removing oxygen from sodium through the use of hot traps are analyzed in substantiating the purification system for a nuclear power station equipped with a BN-1200 reactor. It is shown that a cold trap built into the reactor vessel must be a mandatory component of the reactor plant primary coolant circuit's purification system. The use of hot traps allows oxygen to be removed from the sodium coolant down to permissible concentrations when the nuclear power station operates in its rated mode. The main lines of works aimed at improving the performance characteristics of cold traps are suggested based on the results of performed investigations.

  2. A CAMAC based real-time noise analysis system for nuclear reactors

    NASA Astrophysics Data System (ADS)

    Ciftcioglu, Özer

    1987-05-01

    A CAMAC based real-time noise analysis system was designed for the TRIGA MARK II nuclear reactor at the Institute for Nuclear Energy, Istanbul. The input analog signals obtained from the radiation detectors are introduced to the system through CAMAC interface. The signals converted into digital form are processed by a PDP-11 computer. The fast data processing based on auto/cross power spectral density computations is carried out by means of assembly written FFT algorithms in real-time and the spectra obtained are displayed on a CAMAC driven display system as an additional monitoring device. The system has the advantage of being software programmable and controlled by a CAMAC system so that it is operated under program control for reactor surveillance, anomaly detection and diagnosis. The system can also be used for the identification of nonstationary operational characteristics of the reactor in long term by comparing the noise power spectra with the corresponding reference noise patterns prepared in advance.

  3. Fossil-fuel processing technical/professional services: comparison of Fischer-Tropsch reactor systems. Phase I, final report

    SciTech Connect

    Thompson, G.J.; Riekena, M.L.; Vickers, A.G.

    1981-09-01

    The Fischer-Tropsch reaction was commercialized in Germany and used to produce military fuels in fixed bed reactors. It was recognized from the start that this reactor system had severe operating and yield limitations and alternative reactor systems were sought. In 1955 the Sasol I complex, using an entrained bed (Synthol) reactor system, was started up in South Africa. Although this reactor was a definite improvement and is still operating, the literature is filled with proponents of other reactor systems, each claiming its own advantages. This report provides a summary of the results of a study to compare the development potential of three of these reactor systems with the commercially operating Synthol-entrained bed reactor system. The commercial Synthol reactor is used as a benchmark against which the development potential of the other three reactors can be compared. Most of the information on which this study is based was supplied by the M.W. Kellogg Co. No information beyond that in the literature on the operation of the Synthol reactor system was available for consideration in preparing this study, nor were any details of the changes made to the original Synthol system to overcome the operating problems reported in the literature. Because of conflicting claims and results found in the literature, it was decided to concentrate a large part of this study on a kinetic analysis of the reactor systems, in order to provide a theoretical analysis of intrinsic strengths and weaknesses of the reactors unclouded by different catalysts, operating conditions and feed compositions. The remainder of the study considers the physical attributes of the four reactor systems and compares their respective investment costs, yields, catalyst requirements and thermal efficiencies from simplified conceptual designs.

  4. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  5. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  6. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  7. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  8. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

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

    NASA Technical Reports Server (NTRS)

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

    1991-01-01

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

  10. Future Research in Agile Systems Development: Applying Open Innovation Principles Within the Agile Organisation

    NASA Astrophysics Data System (ADS)

    Conboy, Kieran; Morgan, Lorraine

    A particular strength of agile approaches is that they move away from ‘introverted' development and intimately involve the customer in all areas of development, supposedly leading to the development of a more innovative and hence more valuable information system. However, we argue that a single customer representative is too narrow a focus to adopt and that involvement of stakeholders beyond the software development itself is still often quite weak and in some cases non-existent. In response, we argue that current thinking regarding innovation in agile development needs to be extended to include multiple stakeholders outside the business unit. This paper explores the intra-organisational applicability and implications of open innovation in agile systems development. Additionally, it argues for a different perspective of project management that includes collaboration and knowledge-sharing with other business units, customers, partners, and other relevant stakeholders pertinent to the business success of an organisation, thus embracing open innovation principles.

  11. Open Innovation and the Erosion of the Traditional Information Systems Project's Boundaries

    NASA Astrophysics Data System (ADS)

    Elbanna, Amany

    This paper examines the notion of open innovation and its implication on information systems management. It investigates a project of an enterprise resource planning system implementation in an international organization to unravel the resemblance with the open innovation model. The study shows that the conceptualization of ERP project as an open innovation could reveal the complex architecture of today's organization from which the ERP project cannot be isolated. It argues that the traditional boundaries around IS projects are dissolving and the relationship between what used to be outside and what used to be inside the project is increasingly blurred. The study calls for a different perspective of project management that goes beyond single and multiple project management to scan the open space of innovation and actively look for partners, competitors, and collaborators.

  12. Lunar electric power systems utilizing the SP-100 reactor coupled to dynamic conversion systems

    NASA Technical Reports Server (NTRS)

    Harty, Richard B.; Durand, Richard E.

    1993-01-01

    An integration study was performed by Rocketdyne under contract to NASA-LeRC. The study was concerned with coupling an SP-0100 reactor to either a Brayton or Stirling power conversion system. The application was for a surface power system to supply power requirements to a lunar base. A power level of 550 kWe was selected based on the NASA Space Exploration Initiative 90-day study. Reliability studies were initially performed to determine optimum power conversion redundancy. This study resulted in selecting three operating engines and one stand-by unit. Integration design studies indicated that either the Brayton or Stirling power conversion systems could be integrated with the PS-100 reactor. The Stirling system had an integration advantage because of smaller piping size and fewer components. The Stirling engine, however, is more complex and heavier than the Brayton rotating unit, which tends to off-set the Stirling integration advantage. From a performance consideration, the Brayton had a 9 percent mass advantage, and the Stirling had a 50 percent radiator advantage.

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

    NASA Astrophysics Data System (ADS)

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

    2015-02-01

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

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

    PubMed

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

    2008-08-01

    Molten salt oxidation (MSO) is one of the promising alternative destruction technologies for chlorinated organics, because it is capable of trapping chlorine during organic destruction. This study investigated the characteristics of a two-stage MSO reactor system for the destruction of CCl(4) and C(6)H(5)Cl. Investigated parameters were the MSO reactor temperature (from 1023 K to 1223 K) and the excess oxidizing air feed rate (50% and 100%). The destruction of chlorinated solvents is substantial in the Li(2)CO(3)-Na(2)CO(3) eutectic molten salt, irrespective of the tested condition. However, further oxidation of CO, which is found to be the major destruction product, is not substantial due to the limited temperature and gas residence time in the MSO reactor. Increases in the reactor temperature as well as those in the oxidizing air feed rate consistently lead to decreased emissions of carbon monoxide. No significant influence of the MSO reactor operating condition on the chlorine capturing efficiency was found. Over 99.95% and 99.997% of the chlorine was captured in the hot MSO reactors during the C(6)H(5)Cl and CCl(4) destructions, respectively. This result suggests a relatively low potential of the MSO system in the recombination of chlorinated organics, when compared to a conventional incineration system. PMID:18501405

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

    PubMed

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

    2008-08-01

    Molten salt oxidation (MSO) is one of the promising alternative destruction technologies for chlorinated organics, because it is capable of trapping chlorine during organic destruction. This study investigated the characteristics of a two-stage MSO reactor system for the destruction of CCl(4) and C(6)H(5)Cl. Investigated parameters were the MSO reactor temperature (from 1023 K to 1223 K) and the excess oxidizing air feed rate (50% and 100%). The destruction of chlorinated solvents is substantial in the Li(2)CO(3)-Na(2)CO(3) eutectic molten salt, irrespective of the tested condition. However, further oxidation of CO, which is found to be the major destruction product, is not substantial due to the limited temperature and gas residence time in the MSO reactor. Increases in the reactor temperature as well as those in the oxidizing air feed rate consistently lead to decreased emissions of carbon monoxide. No significant influence of the MSO reactor operating condition on the chlorine capturing efficiency was found. Over 99.95% and 99.997% of the chlorine was captured in the hot MSO reactors during the C(6)H(5)Cl and CCl(4) destructions, respectively. This result suggests a relatively low potential of the MSO system in the recombination of chlorinated organics, when compared to a conventional incineration system.

  16. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

    NASA Technical Reports Server (NTRS)

    George, Jeffrey A.

    1991-01-01

    A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

  17. Preliminary design of reactor power systems for the manned space base.

    NASA Technical Reports Server (NTRS)

    Mckhann, G. G.; Coggi, J. V.; Diamond, S. D.

    1972-01-01

    The results of design integration studies of uranium-zirconium hydride (UZr-Hx) reactor power systems for the NASA space base study program are presented. The power conversion systems investigated include the Brayton cycle, the organic Rankine cycle, the SNAP-8 mercury Rankine cycle, and thermoelectric (PbTe). The proposed space base has a 10-year life and requires 100 kWe of power. Two 50-kWe power systems with a nominal replacement life of 5 years are utilized. Parametric design data such as life, weight, radiator area, reactor outlet-temperature, reactor thermal power, and power conversion system efficiency are presented and used for the design and integration of the system with the space base.

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

    NASA Astrophysics Data System (ADS)

    Bragg-Sitton, Shannon M.

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

  19. Triple Helix Systems: An Analytical Framework for Innovation Policy and Practice in the Knowledge Society

    ERIC Educational Resources Information Center

    Ranga, Marina; Etzkowitz, Henry

    2013-01-01

    This paper introduces the concept of Triple Helix systems as an analytical construct that synthesizes the key features of university--industry--government (Triple Helix) interactions into an "innovation system" format, defined according to systems theory as a set of components, relationships and functions. Among the components of Triple…

  20. Do entrepreneurial food systems innovations impact rural economies and health? Evidence and gaps

    PubMed Central

    Sitaker, Marilyn; Kolodinsky, Jane; Jilcott Pitts, Stephanie B.; Seguin, Rebecca A.

    2015-01-01

    A potential solution for weakened rural economies is the development of local food systems, which include affordable foods sources for consumers and economically feasible structures for producers. Local food systems are purported to promote sustainability, improve local economies, increase access to healthy foods, and improve the local diets. Four entrepreneurial food systems innovations that support local economies include farmers’ markets, community supported agriculture, farm to institution programs and food hubs. We review current literature to determine whether innovations for aggregation, processing, distribution and marketing in local food systems: 1) enable producers to make a living; 2) improve local economies; 3) provide local residents with greater access to affordable, healthy food; and 4) contribute to greater consumption of healthy food among residents. While there is some evidence for each, more transdisciplinary research is needed to determine whether entrepreneurial food systems innovations provide economic and public health benefits. PMID:26613066