Sample records for scram

  1. The SCRAM tool-kit

    NASA Technical Reports Server (NTRS)

    Tamir, David; Flanigan, Lee A.; Weeks, Jack L.; Siewert, Thomas A.; Kimbrough, Andrew G.; Mcclure, Sidney R.

    1994-01-01

    This paper proposes a new series of on-orbit capabilities to support the near-term Hubble Space Telescope, Extended Duration Orbiter, Long Duration Orbiter, Space Station Freedom, other orbital platforms, and even the future manned Lunar/Mars missions. These proposed capabilities form a toolkit termed Space Construction, Repair, and Maintenance (SCRAM). SCRAM addresses both intra-Vehicular Activity (IVA) and Extra-Vehicular Activity (EVA) needs. SCRAM provides a variety of tools which enable welding, brazing, cutting, coating, heating, and cleaning, as well as corresponding nondestructive examination. Near-term IVA-SCRAM applications include repair and modification to fluid lines, structure, and laboratory equipment inside a shirt-sleeve environment (i.e. inside Spacelab or Space Station). Near-term EVA-SCRAM applications include construction of fluid lines and structural members, repair of punctures by orbital debris, refurbishment of surfaces eroded by contaminants. The SCRAM tool-kit also promises future EVA applications involving mass production tasks automated by robotics and artificial intelligence, for construction of large truss, aerobrake, and nuclear reactor shadow shields structures. The leading candidate tool processes for SCRAM, currently undergoing research and development, include Electron Beam, Gas Tungsten Arc, Plasma Arc, and Laser Beam. A series of strategic space flight experiments would make SCRAM available to help conquer the space frontier.

  2. Army gas-cooled reactor systems program. Preliminary design report off-normal scram system

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bushnell, W.H.; Malmstrom, S.A.

    1965-06-01

    The maximum allowable ML-1 fuel element cladding (hot spot) temperature is established by ANTS 201 at 1750/sup 0/F. The existing ML-1 design makes no provision for automatic scram when this limit is reached. Operating experience has indicated a requirement for such an automatic system during plant startup and a revised hot spot envelope (generated during conceptual design of the scram system) established the desirability of extending this protection to operation at full power conditions. It was also determined that the scram system should include circuitry to initiate an automatic scram if reactor ..delta..T exceeded 450/sup 0/F (the limit established inmore » ANTS 201) and if reactor power exceeded 6 kw(t) without coolant flow in the main loop. The preliminary design of the scram system (designated off-normal scram system) which will provide the required protection is described.« less

  3. Comparative study and evaluation of SCRAM use, recidivism rates, and characteristics.

    DOT National Transportation Integrated Search

    2015-04-01

    SCRAM (Secure Continuous Remote Alcohol Monitoring) is an ankle bracelet that conducts transdermal readings by sampling alcohol vapor just above the skin or insensible perspiration. It provides continuous monitoring of sobriety. The impact of SCRAM o...

  4. 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.

  5. 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.

  6. EVA-SCRAM operations

    NASA Technical Reports Server (NTRS)

    Flanigan, Lee A.; Tamir, David; Weeks, Jack L.; Mcclure, Sidney R.; Kimbrough, Andrew G.

    1994-01-01

    This paper wrestles with the on-orbit operational challenges introduced by the proposed Space Construction, Repair, and Maintenance (SCRAM) tool kit for Extra-Vehicular Activity (EVA). SCRAM undertakes a new challenging series of on-orbit tasks in support of the near-term Hubble Space Telescope, Extended Duration Orbiter, Long Duration Orbiter, Space Station Freedom, other orbital platforms, and even the future manned Lunar/Mars missions. These new EVA tasks involve welding, brazing, cutting, coating, heat-treating, and cleaning operations. Anticipated near-term EVA-SCRAM applications include construction of fluid lines and structural members, repair of punctures by orbital debris, refurbishment of surfaces eroded by atomic oxygen, and cleaning of optical, solar panel, and high emissivity radiator surfaces which have been degraded by contaminants. Future EVA-SCRAM applications are also examined, involving mass production tasks automated with robotics and artificial intelligence, for construction of large truss, aerobrake, and reactor shadow shield structures. Realistically achieving EVA-SCRAM is examined by addressing manual, teleoperated, semi-automated, and fully-automated operation modes. The operational challenges posed by EVA-SCRAM tasks are reviewed with respect to capabilities of existing and upcoming EVA systems, such as the Extravehicular Mobility Unit, the Shuttle Remote Manipulating System, the Dexterous End Effector, and the Servicing Aid Tool.

  7. Computational Model of Heat Transfer on the ISS

    NASA Technical Reports Server (NTRS)

    Torian, John G.; Rischar, Michael L.

    2008-01-01

    SCRAM Lite (SCRAM signifies Station Compact Radiator Analysis Model) is a computer program for analyzing convective and radiative heat-transfer and heat-rejection performance of coolant loops and radiators, respectively, in the active thermal-control systems of the International Space Station (ISS). SCRAM Lite is a derivative of prior versions of SCRAM but is more robust. SCRAM Lite computes thermal operating characteristics of active heat-transport and heat-rejection subsystems for the major ISS configurations from Flight 5A through completion of assembly. The program performs integrated analysis of both internal and external coolant loops of the various ISS modules and of an external active thermal control system, which includes radiators and the coolant loops that transfer heat to the radiators. The SCRAM Lite run time is of the order of one minute per day of mission time. The overall objective of the SCRAM Lite simulation is to process input profiles of equipment-rack, crew-metabolic, and other heat loads to determine flow rates, coolant supply temperatures, and available radiator heat-rejection capabilities. Analyses are performed for timelines of activities, orbital parameters, and attitudes for mission times ranging from a few hours to several months.

  8. OVERALL CONTROL SYSTEM FOR HIGH FLUX PILE

    DOEpatents

    Newson, H.W.; Durham, N.C.; Wigner, E.P.; Princeton, N.J.; Epler, E.P.

    1961-05-23

    A control system is given for a high fiux reactor incorporating an anti- scram control feature whereby a neutron absorbing control rod acts as a fine adjustment while a neutron absorbing shim rod, actuated upon a command received from reactor period and level signals, has substantially greater effect on the neutron level and is moved prior to scram conditions to alter the reactor activity before a scram condition is created. Thus the probability that a scram will have to be initiated is substantially decreased.

  9. SCRAM: a pipeline for fast index-free small RNA read alignment and visualization.

    PubMed

    Fletcher, Stephen J; Boden, Mikael; Mitter, Neena; Carroll, Bernard J

    2018-03-15

    Small RNAs play key roles in gene regulation, defense against viral pathogens and maintenance of genome stability, though many aspects of their biogenesis and function remain to be elucidated. SCRAM (Small Complementary RNA Mapper) is a novel, simple-to-use short read aligner and visualization suite that enhances exploration of small RNA datasets. The SCRAM pipeline is implemented in Go and Python, and is freely available under MIT license. Source code, multiplatform binaries and a Docker image can be accessed via https://sfletc.github.io/scram/. s.fletcher@uq.edu.au. Supplementary data are available at Bioinformatics online.

  10. 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.

  11. 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.

  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. Scram signal generator

    DOEpatents

    Johanson, Edward W.; Simms, Richard

    1981-01-01

    A scram signal generating circuit for nuclear reactor installations monitors a flow signal representing the flow rate of the liquid sodium coolant which is circulated through the reactor, and initiates reactor shutdown for a rapid variation in the flow signal, indicative of fuel motion. The scram signal generating circuit includes a long-term drift compensation circuit which processes the flow signal and generates an output signal representing the flow rate of the coolant. The output signal remains substantially unchanged for small variations in the flow signal, attributable to long term drift in the flow rate, but a rapid change in the flow signal, indicative of a fast flow variation, causes a corresponding change in the output signal. A comparator circuit compares the output signal with a reference signal, representing a given percentage of the steady state flow rate of the coolant, and generates a scram signal to initiate reactor shutdown when the output signal equals the reference signal.

  14. Scram signal generator

    DOEpatents

    Johanson, E.W.; Simms, R.

    A scram signal generating circuit for nuclear reactor installations monitors a flow signal representing the flow rate of the liquid sodium coolant which is circulated through the reactor, and initiates reactor shutdown for a rapid variation in the flow signal, indicative of fuel motion. The scram signal generating circuit includes a long-term drift compensation circuit which processes the flow signal and generates an output signal representing the flow rate of the coolant. The output signal remains substantially unchanged for small variations in the flow signal, attributable to long term drift in the flow rate, but a rapid change in the flow signal, indicative of a fast flow variation, causes a corresponding change in the output signal. A comparator circuit compares the output signal with a reference signal, representing a given percentage of the steady state flow rate of the coolant, and generates a scram signal to initiate reactor shutdown when the output signal equals the reference signal.

  15. Summary of Resource Conservation and Recovery Act (RCRA) State Authorization Rule Checklist 3006(f)

    EPA Pesticide Factsheets

    This checklist is an electronic version of the original document found in the 1986 State Consolidated RCRA Authorization Manual (SCRAM). The checklist has not undergone any formal legal review since publication in the SCRAM.

  16. Cerebellar contributions to biological motion perception in autism and typical development.

    PubMed

    Jack, Allison; Keifer, Cara M; Pelphrey, Kevin A

    2017-04-01

    Growing evidence suggests that posterior cerebellar lobe contributes to social perception in healthy adults. However, they know little about how this process varies across age and with development. Using cross-sectional fMRI data, they examined cerebellar response to biological (BIO) versus scrambled (SCRAM) motion within typically developing (TD) and autism spectrum disorder (ASD) samples (age 4-30 years old), characterizing cerebellar response and BIO > SCRAM-selective effective connectivity, as well as associations with age and social ability. TD individuals recruited regions throughout cerebellar posterior lobe during BIO > SCRAM, especially bilateral lobule VI, and demonstrated connectivity with right posterior superior temporal sulcus (RpSTS) in left VI, Crus I/II, and VIIIb. ASD individuals showed BIO > SCRAM activity in left VI and left Crus I/II, and bilateral connectivity with RpSTS in Crus I/II and VIIIb/IX. No between-group differences emerged in well-matched subsamples. Among TD individuals, older age predicted greater BIO > SCRAM response in left VIIb and left VIIIa/b, but reduced connectivity between RpSTS and widespread regions of the right cerebellum. In ASD, older age predicted greater response in left Crus I and bilateral Crus II, but decreased effective connectivity with RpSTS in bilateral Crus I/II. In ASD, increased BIO > SCRAM signal in left VI/Crus I and right Crus II, VIIb, and dentate predicted lower social symptomaticity; increased effective connectivity with RpSTS in right Crus I/II and bilateral VI and I-V predicted greater symptomaticity. These data suggest that posterior cerebellum contributes to the neurodevelopment of social perception in both basic and clinical populations. Hum Brain Mapp 38:1914-1932, 2017. © 2017 Wiley Periodicals, Inc. © 2017 Wiley Periodicals, Inc.

  17. Browns Ferry Nuclear Plant Unit 2: Control rod drive scram discharge headers decontamination effort

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Traynor, J.C.

    1983-08-01

    The control rod drive (CRD) scram discharge headers were decontaminated during the Browns Ferry unit 2, cycle 4 refueling outage (August 2-5, 1982). Hydrolasing (high-pressure water blasting) was used as the method of decontamination to remove fixed and loose radioactive contaminants from the headers. It was found that hydrolasing of the west scram discharge headers resulted in approximate maximum and average decontamination factors (DFs) on contact of 13 and 5, respectively. For the east scram discharge headers, hydrolasing resulted in a maximum and average DF on contact of approximately 3. The maximum and average DFs on contact for the individualmore » headers ranged from 1 to 33 and 1 to 10, respectively, while the walkway (head-level) DFs were in the range of 3 to 4. Higher DFs were impeded by inadequate drainage and backwashing of fluid. This led to increased radiation levels in some areas and recontamination of adjacent headers.« less

  18. Numerical Simulation of Measurements during the Reactor Physical Startup at Unit 3 of Rostov NPP

    NASA Astrophysics Data System (ADS)

    Tereshonok, V. A.; Kryakvin, L. V.; Pitilimov, V. A.; Karpov, S. A.; Kulikov, V. I.; Zhylmaganbetov, N. M.; Kavun, O. Yu.; Popykin, A. I.; Shevchenko, R. A.; Shevchenko, S. A.; Semenova, T. V.

    2017-12-01

    The results of numerical calculations and measurements of some reactor parameters during the physical startup tests at unit 3 of Rostov NPP are presented. The following parameters are considered: the critical boron acid concentration and the currents from ionization chambers (IC) during the scram system efficiency evaluation. The scram system efficiency was determined using the inverse point kinetics equation with the measured and simulated IC currents. The results of steady-state calculations of relative power distribution and efficiency of the scram system and separate groups of control rods of the control and protection system are also presented. The calculations are performed using several codes, including precision ones.

  19. On the interpretation of the inverted kinetics equation and space-time calculations of the effectiveness of the VVER-1000 reactor scram system

    NASA Astrophysics Data System (ADS)

    Zizin, M. N.; Ivanov, L. D.

    2013-12-01

    In the present paper, an attempt is made to analyze the accuracy of calculating the effectiveness of the VVER-1000 reactor scram system by means of the inverted solution of the kinetics equation (ISKE). In the numerical studies in the intellectual ShIPR software system, the actuation of the reactor scram system with the possible jamming of one of the two most effective rods is simulated. First, the connection of functionals calculated in the space-time computation in different approximations with the kinetics equation is considered on the theoretical level. The formulas are presented in a manner facilitating their coding. Then, the results of processing of several such functions by the ISKE are presented. For estimating the effectiveness of the VVER-1000 reactor scram system, it is proposed to use the measured currents of ionization chambers (IC) jointly with calculated readings of IC imitators. In addition, the integral of the delayed neutron (DN) generation rate multiplied by the adjoint DN source over the volume of the reactor, calculated for the instant of time when insertion of safety rods ends, is used. This integral is necessary for taking into account the spatial reactivity effects. Reasonable agreement was attained for the considered example between the effectiveness of the scram system evaluated by this method and the values obtained by steady-state calculations as the difference of the reciprocal effective multiplication factors with withdrawn and inserted control rods. This agreement was attained with the use of eight-group DN parameters.

  20. Conceptual study of space plane powered by hypersonic airbreathing propulsion system

    NASA Astrophysics Data System (ADS)

    Maita, Masataka; Ohkami, Yoshiaki; Yamanaka, Tatsuo; Mori, Takashige

    1990-10-01

    The paper describes the investigations of aerospace plane concept, conducted by the National Aerospace Laboratory (NAL) of Japan, with particular attention given to a concept which integrates a scram/liquid air cycle engine (LACE) hypersonic propulsion system fueling with slush hydrogen. The key requirements in achieving the space plane using scram/LACE propulsion system are described along with the mission requirements and the vehicle characteristics. Typical outputs of SSTO analysis are presented.

  1. Fluidic self-actuating control assembly

    DOEpatents

    Grantz, Alan L.

    1979-01-01

    A fluidic self-actuating control assembly for use in a reactor wherein no external control inputs are required to actuate (scram) the system. The assembly is constructed to scram upon sensing either a sudden depressurization of reactor inlet flow or a sudden increase in core neutron flux. A fluidic control system senses abnormal flow or neutron flux transients and actuates the system, whereupon assembly coolant flow reverses, forcing absorber balls into the reactor core region.

  2. Quick Fix for Managing Risks

    NASA Technical Reports Server (NTRS)

    2002-01-01

    Under a Phase II SBIR contract, Kennedy and Lumina Decision Systems, Inc., jointly developed the Schedule and Cost Risk Analysis Modeling (SCRAM) system, based on a version of Lumina's flagship software product, Analytica(R). Acclaimed as "the best single decision-analysis program yet produced" by MacWorld magazine, Analytica is a "visual" tool used in decision-making environments worldwide to build, revise, and present business models, minus the time-consuming difficulty commonly associated with spreadsheets. With Analytica as their platform, Kennedy and Lumina created the SCRAM system in response to NASA's need to identify the importance of major delays in Shuttle ground processing, a critical function in project management and process improvement. As part of the SCRAM development project, Lumina designed a version of Analytica called the Analytica Design Engine (ADE) that can be easily incorporated into larger software systems. ADE was commercialized and utilized in many other developments, including web-based decision support.

  3. CONTROL ROD DRIVE

    DOEpatents

    Chapellier, R.A.

    1960-05-24

    BS>A drive mechanism was invented for the control rod of a nuclear reactor. Power is provided by an electric motor and an outside source of fluid pressure is utilized in conjunction with the fluid pressure within the reactor to balance the loadings on the motor. The force exerted on the drive mechanism in the direction of scramming the rod is derived from the reactor fluid pressure so that failure of the outside pressure source will cause prompt scramming of the rod.

  4. Knockdown of Progesterone Receptor (PGR) in Macaque Granulosa Cells Disrupts Ovulation and Progesterone Production.

    PubMed

    Bishop, Cecily V; Hennebold, Jon D; Kahl, Christoph A; Stouffer, Richard L

    2016-05-01

    Adenoviral vectors (vectors) expressing short-hairpin RNAs complementary to macaque nuclear progesterone (P) receptor PGR mRNA (shPGR) or a nontargeting scrambled control (shScram) were used to determine the role PGR plays in ovulation/luteinization in rhesus monkeys. Nonluteinized granulosa cells collected from monkeys (n = 4) undergoing controlled ovarian stimulation protocols were exposed to either shPGR, shScram, or no virus for 24 h; human chorionic gonadotropin (hCG) was then added to half of the wells to induce luteinization (luteinized granulosa cells [LGCs]; n = 4-6 wells/treatment/monkey). Cells/media were collected 48, 72, and 120 h postvector for evaluation of PGR mRNA and P levels. Addition of hCG increased (P < 0.05) PGR mRNA and medium P levels in controls. However, a time-dependent decline (P < 0.05) in PGR mRNA and P occurred in shPGR vector groups. Injection of shPGR, but not shScram, vector into the preovulatory follicle 20 h before hCG administration during controlled ovulation protocols prevented follicle rupture in five of six monkeys as determined by laparoscopic evaluation, with a trapped oocyte confirmed in three of four follicles of excised ovaries. Injection of shPGR also prevented the rise in serum P levels following the hCG bolus compared to shScram (P < 0.05). Nuclear PGR immunostaining was undetectable in granulosa cells from shPGR-injected follicles, compared to intense staining in shScram controls. Thus, the nuclear PGR appears to mediate P action in the dominant follicle promoting ovulation in primates. In vitro and in vivo effects of PGR knockdown in LGCs also support the hypothesis that P enhances its own synthesis in the primate corpus luteum by promoting luteinization. © 2016 by the Society for the Study of Reproduction, Inc.

  5. Control rod drive hydraulic system

    DOEpatents

    Ose, Richard A.

    1992-01-01

    A hydraulic system for a control rod drive (CRD) includes a variable output-pressure CR pump operable in a charging mode for providing pressurized fluid at a charging pressure, and in a normal mode for providing the pressurized fluid at a purge pressure, less than the charging pressure. Charging and purge lines are disposed in parallel flow between the CRD pump and the CRD. A hydraulic control unit is disposed in flow communication in the charging line and includes a scram accumulator. An isolation valve is provided in the charging line between the CRD pump and the scram accumulator. A controller is operatively connected to the CRD pump and the isolation valve and is effective for opening the isolation valve and operating the CRD pump in a charging mode for charging the scram accumulator, and closing the isolation valve and operating the CRD pump in a normal mode for providing to the CRD through the purge line the pressurized fluid at a purge pressure lower than the charging pressure.

  6. ATWS at Browns Ferry Unit One - accident sequence analysis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Harrington, R.M.; Hodge, S.A.

    1984-07-01

    This study describes the predicted response of Unit One at the Browns Ferry Nuclear Plant to a postulated complete failure to scram following a transient occurrence that has caused closure of all Main Steam Isolation Valves (MSIVs). This hypothetical event constitutes the most severe example of the type of accident classified as Anticipated Transient Without Scram (ATWS). Without the automatic control rod insertion provided by scram, the void coefficient of reactivity and the mechanisms by which voids are formed in the moderator/coolant play a dominant role in the progression of the accident. Actions taken by the operator greatly influence themore » quantity of voids in the coolant and the effect is analyzed in this report. The progression of the accident sequence under existing and under recommended procedures is discussed. For the extremely unlikely cases in which equipment failure and wrongful operator actions might lead to severe core damage, the sequence of emergency action levels and the associated timing of events are presented.« less

  7. CONTROL ROD DRIVE

    DOEpatents

    Chapellier, R.A.; Rogers, I.

    1961-06-27

    Accurate and controlled drive for the control rod is from an electric motor. A hydraulic arrangement is provided to balance a piston against which a control rod is urged by the application of fluid pressure. The electric motor drive of the control rod for normal operation is made through the aforementioned piston. In the event scramming is required, the fluid pressure urging the control rod against the piston is relieved and an opposite fluid pressure is applied. The lack of mechanical connection between the electric motor and control rod facilitates the scramming operation.

  8. Impact of nonabsorbing control rod tips on kinetics feedback for BWR turbine trip without bypass RETRAN-03 analyses

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Feltus, M.A.; Knerr, R.; Shoop, U.

    1993-01-01

    RETRAN-03 studies were performed for the boiling water reactor (BWR) turbine trip without bypass (TTWOB) event to investigate how the non-neutron-absorbing material on control rod tips affect scram delay timing and reactivity feedback. Scram delay, Doppler temperature, and moderator void (density) feedback were varied to assess their relative impact on kinetics behavior. Although a generic point-kinetics RETRAN-03 TTWOB model 2 was employed, actual plant information was used to develop the basic and parametric cases.

  9. Control rod drive for reactor shutdown

    DOEpatents

    McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

    1976-01-20

    A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

  10. CONTEMPT/LT-028 Browns Ferry studies of an anticipated transient without scram

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Holcomb, E.E.

    1983-01-01

    The Browns Ferry Nuclear Plant containment response during the first 30 min of an anticipated transient without scram (ATWS) is the subject of this paper. Three cases, each initiated by a main steam isolation valve closure, are presented: the ATWS is mitigated by operator actions in the spirit of the General Electric Emergency Procedure Guidelines; the ATWS is managed by the plant automatic control systems; and the ATWS proceeds as in first case except that the drywell coolers are unavailable. Success of the standby liquid control system is assumed in the last two transients.

  11. DOE Office of Scientific and Technical Information (OSTI.GOV)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

    The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos supported the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. Baseline calculations of the PIUS design were performed for active and passive reactor scrams using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following active-system scrams.

  12. Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dallman, R J; Gottula, R C; Holcomb, E E

    1987-05-01

    An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented.

  13. Rocket-Based Combined Cycle Engine Concept Development

    NASA Technical Reports Server (NTRS)

    Ratekin, G.; Goldman, Allen; Ortwerth, P.; Weisberg, S.; McArthur, J. Craig (Technical Monitor)

    2001-01-01

    The development of rocket-based combined cycle (RBCC) propulsion systems is part of a 12 year effort under both company funding and contract work. The concept is a fixed geometry integrated rocket, ramjet, scramjet, which is hydrogen fueled and uses hydrogen regenerative cooling. The baseline engine structural configuration uses an integral structure that eliminates panel seals, seal purge gas, and closeout side attachments. Engine A5 is the current configuration for NASA Marshall Space Flight Center (MSFC) for the ART program. Engine A5 models the complete flight engine flowpath of inlet, isolator, airbreathing combustor, and nozzle. High-performance rocket thrusters are integrated into the engine enabling both low speed air-augmented rocket (AAR) and high speed pure rocket operation. Engine A5 was tested in GASL's new Flight Acceleration Simulation Test (FAST) facility in all four operating modes, AAR, RAM, SCRAM, and Rocket. Additionally, transition from AAR to RAM and RAM to SCRAM was also demonstrated. Measured performance demonstrated vision vehicle performance levels for Mach 3 AAR operation and ramjet operation from Mach 3 to 4. SCRAM and rocket mode performance was above predictions. For the first time, testing also demonstrated transition between operating modes.

  14. XCALIBUR: a Vertical Takeoff TSTO RLV Concept with a HEDM Upperstage and a Scram-Rocket Booster

    NASA Astrophysics Data System (ADS)

    Bradford, J.

    2002-01-01

    A new 3rd generation, two-stage-to-orbit (TSTO) reusable launch vehicle (RLV) has been designed. The Xcalibur concept represents a novel approach due to its integration method for the upperstage element of the system. The vertical-takeoff booster, which is powered by rocket-based combined-cycle (RBCC) engines, carries the upperstage internally in the aft section of the airframe to a Mach 15 staging condition. The upperstage is released from the booster and carries the 6,820 kg of payload to low earth orbit (LEO) using its high energy density matter (HEDM) propulsion system. The booster element is capable of returning to the original launch site in a ramjet-cruise propulsion mode. Both the booster and the upperstage utilize advanced technologies including: graphite-epoxy tanks, metal-matrix composites, UHTC TPS materials, electro- mechanical actuators (EMAs), and lightweight subsystems (avionics, power distribution, etc.). The booster system is enabled main propulsion system which utilizes four RBCC engines. These engines operate in four distinct modes: air- augmented rocket (AAR), ramjet, scram-rocket, and all-rocket. The booster operates in AAR mode from takeoff to Mach 3, with ramjet mode operation from Mach 3 to Mach 6. The rocket re-ignition for scram-rocket mode occurs at Mach 6, with all-rocket mode from Mach 14 to the staging condition. The extended utilization of the scram-rocket mode greatly improves vehicle performance by providing superior vehicle acceleration when compared to the scramjet mode performance over the same flight region. Results indicate that the specific impulse penalty due to the scram-rocket mode operation is outweighed by the reduced flight time, smaller vehicle size due to increased mixture ratio, and lower allowable maximum dynamic pressure. A complete vehicle system life-cycle analysis was performed in an automated, multi-disciplinary design environment. Automated disciplinary performance analysis tools include: trajectory (POST), propulsion (SCCREAM), aeroheating (TCAT II), and an Excel spreadsheet for component weight estimation. These tools were automated using `file wrappers' in Phoenix Integration's ModelCenter collaborative design environment. Performance tools utilized for the analysis, but not requiring automation included IDEAS for solid modeling and APAS for the aerodynamic analysis. The paper describes the vehicle concept and operation, discussing the types of technologies used and the nominal flight scenario. A brief discussion explaining the decision-making process for the vehicle configuration is included. For cost predictions, NAFCOM-derived cost estimating relationships were used. Economic predictions were developed using a number of codes, including CABAM (financials), AATe (operations), and GTSafetyII (safety and reliability).

  15. Quick release latch for reactor scram

    DOEpatents

    Johnson, Melvin L.; Shawver, Bruce M.

    1976-01-01

    A simple, reliable, and fast-acting means for releasing a control element and allowing it to be inserted rapidly into the core region of a nuclear reactor for scram purposes. A latch mechanism grips a coupling head on a nuclear control element to connect the control element to the control drive assembly. The latch mechanism is closed by tensioning a cable or rod with an actuator. The control element is released by de-energizing the actuator, providing fail-safe, rapid release of the control element to effect reactor shutdown. A sensing rod provides indication that the control element is properly positioned in the latch. Two embodiments are illustrated, one involving a collet-type latch mechanism, the other a pliers-type latch mechanism with the actuator located inside the reactor vessel.

  16. Quick release latch for reactor scram

    DOEpatents

    Johnson, M.L.; Shawver, B.M.

    1975-09-16

    A simple, reliable, and fast-acting means for releasing a control element and allowing it to be inserted rapidly into the core region of a nuclear reactor for scram purposes is described. A latch mechanism grips a coupling head on a nuclear control element to connect the control element to the control drive assembly. The latch mechanism is closed by tensioning a cable or rod with an actuator. The control element is released by de-energizing the actuator, providing fail-safe, rapid release of the control element to effect reactor shutdown. A sensing rod provides indication that the control element is properly positioned in the latch. Two embodiments are illustrated, one involving a collet- type latch mechanism, the other a pliers-type latch mechanism with the actuator located inside the reactor vessel. (auth)

  17. Support Center for Regulatory Atmospheric Modeling (SCRAM)

    EPA Pesticide Factsheets

    This technical site provides access to air quality models (including computer code, input data, and model processors) and other mathematical simulation techniques used in assessing air emissions control strategies and source impacts.

  18. 10 CFR 50.73 - Licensee event report system.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... scram or reactor trip. (2) General containment isolation signals affecting containment isolation valves... plant design; or (2) Normal and expected wear or degradation. (x) Any event that posed an actual threat...

  19. 10 CFR 50.73 - Licensee event report system.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... scram or reactor trip. (2) General containment isolation signals affecting containment isolation valves... plant design; or (2) Normal and expected wear or degradation. (x) Any event that posed an actual threat...

  20. 10 CFR 50.73 - Licensee event report system.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... scram or reactor trip. (2) General containment isolation signals affecting containment isolation valves... plant design; or (2) Normal and expected wear or degradation. (x) Any event that posed an actual threat...

  1. A size-composition resolved aerosol model for simulating the dynamics of externally mixed particles: SCRAM (v 1.0)

    NASA Astrophysics Data System (ADS)

    Zhu, S.; Sartelet, K. N.; Seigneur, C.

    2015-06-01

    The Size-Composition Resolved Aerosol Model (SCRAM) for simulating the dynamics of externally mixed atmospheric particles is presented. This new model classifies aerosols by both composition and size, based on a comprehensive combination of all chemical species and their mass-fraction sections. All three main processes involved in aerosol dynamics (coagulation, condensation/evaporation and nucleation) are included. The model is first validated by comparison with a reference solution and with results of simulations using internally mixed particles. The degree of mixing of particles is investigated in a box model simulation using data representative of air pollution in Greater Paris. The relative influence on the mixing state of the different aerosol processes (condensation/evaporation, coagulation) and of the algorithm used to model condensation/evaporation (bulk equilibrium, dynamic) is studied.

  2. Evaluating transdermal alcohol measuring devices

    DOT National Transportation Integrated Search

    2007-11-01

    This report is an evaluation study of two types of transdermal devices that detect alcohol at the skin surface representing two types of electrochemical sensing technology. The AMS SCRAM ankle device and the Giner WrisTAS wrist device were worn...

  3. 78 FR 41118 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-09

    ..., such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC... range monitoring (APRM) system ``Upscale'' and ``Inoperative'' scram and control rod withdrawal block...

  4. Computed potential energy surfaces for chemical reactions

    NASA Technical Reports Server (NTRS)

    Walch, Stephen P.

    1990-01-01

    The objective was to obtain accurate potential energy surfaces (PES's) for a number of reactions which are important in the H/N/O combustion process. The interest in this is centered around the design of the SCRAM jet engine for the National Aerospace Plane (NASP), which was envisioned as an air-breathing hydrogen-burning vehicle capable of reaching velocities as large as Mach 25. Preliminary studies indicated that the supersonic flow in the combustor region of the scram jet engine required accurate reaction rate data for reactions in the H/N/O system, some of which was not readily available from experiment. The most important class of combustion reactions from the standpoint of the NASP project are radical recombinaton reactions, since these reactions result in most of the heat release in the combustion process. Theoretical characterizations of the potential energy surfaces for these reactions are presented and discussed.

  5. Taylor Impact Tests and Simulations on PBX 9501

    NASA Astrophysics Data System (ADS)

    Clements, Brad; Thompson, Darla G.; Luscher, D. J.; Deluca, Racci

    2011-06-01

    Taylor impact tests have been conducted previously on plastic bonded explosives (PBXs) to characterize the stress state of these materials as they impact smooth and flat steel anvil surfaces at speeds of ~100m/s (i.e. Christopher, et al, 11th Detonation Symposium). In 2003, C. Liu and R. Ellis (unpublished, Los Alamos National Laboratory) performed Taylor tests on PBX 9501 up to speeds of 115 m/s, capturing impact images. In the work presented here, we have extended these tests to velocities of 200 m/s using a composite-lined gun barrel and no specimen sabot. Specimen images are used to validate the thermo-mechanical constitutive model ViscoSCRAM. ViscoSCRAM has been parameterized for PBX 9501 in uniaxial stress configurations. Simulating Taylor impact experiments tests the model in situations undergoing extreme damage. In addition, experimental variations to specimen confinement and friction are introduced in an attempt to establish ignition thresholds in this velocity regime.

  6. Method for automatically scramming a nuclear reactor

    DOEpatents

    Ougouag, Abderrafi M.; Schultz, Richard R.; Terry, William K.

    2005-12-27

    An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

  7. ATWS analysis for Browns Ferry Nuclear Plant Unit 1

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dallman, R.J.; Jouse, W.C.

    1985-01-01

    Analyses of postulated Anticipated Transients Without Scram (ATWS) were performed at the Idaho National Engineering Laboratory (INEL). The Browns Ferry Nuclear Plant Unit 1 (BFNP1) was selected as the subject of this work because of the cooperation of the Tennessee Valley Authority (TVA). The work is part of the Severe Accident Sequence Analysis (SASA) Program of the US Nuclear Regulatory Commission (NRC). A Main Steamline Isolation Valve (MSIV) closure served as the transient initiator for these analyses, which proceeded a complete failure to scram. Results from the analyses indicate that operator mitigative actions are required to prevent overpressurization of themore » primary containment. Uncertainties remain concerning the effectiveness of key mitigative actions. The effectiveness of level control as a power reduction procedure is limited. Power level resulting from level control only reduce the Pressure Suppression Pool (PSP) heatup rate from 6 to 4F/min.« less

  8. SBLOCA outside containment at Browns Ferry Unit One. Volume 2. Iodine, cesium, and noble gas distribution and release

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wichner, R.P.; Weber, C.F.; Wright, A.L.

    1983-09-01

    This is the second volume of a two-part study regarding the response of Browns Ferry Unit 1 to a postulated break in the scram discharge volume of the control rod drive hydraulic system immediately following a scram. The material in this second volume pertains to the second aspect of the study, the resultant transport of fission products from their original locations in the fuel to a series of repositories within the primary system, the primary and secondary containment structures, and ultimately the release of a small portion to the environment. Transport models are developed for the noble gases krypton andmore » xenon and for iodine and cesium to describe the release of these fission products from the overheated fuel and their subsequent movement under the conditions predicted to exist in the various repositories during the course of the accident.« less

  9. SBLOCA outside containment at Browns Ferry Unit One: accident sequence analysis. [Small break

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Condon, W.A.; Harrington, R.M.; Greene, S.R.

    1982-11-01

    This study describes the predicted response of Unit 1 at the Browns Ferry Nuclear Plant to a postulated small-break loss-of-coolant accident outside of the primary containment. The break has been assumed to occur in the scram discharge volume piping immediately following a reactor scram that cannot be reset. The events before core uncovering are discussed for both the worst-case accident sequence without operator action and for the more likely sequences with operator action. Without operator action, the events after core uncovering would include core meltdown and subsequent containment failure, and this event sequence has been determined through use of themore » MARCH code. An estimate of the magnitude and timing of the concomitant release of the noble gas, cesium, and iodine-based fission products to the environment is provided in Volume 2 of this report.« less

  10. DOE Office of Scientific and Technical Information (OSTI.GOV)

    Steiner, J.L.; Lime, J.F.; Elson, J.S.

    One dimensional TRAC transient calculations of the process inherent ultimate safety (PIUS) advanced reactor design were performed for a pump-trip SCRAM. The TRAC calculations showed that the reactor power response and shutdown were in qualitative agreement with the one-dimensional analyses presented in the PIUS Preliminary Safety Information Document (PSID) submitted by Asea Brown Boveri (ABB) to the US Nuclear Regulatory Commission for preapplication safety review. The PSID analyses were performed with the ABB-developed RIGEL code. The TRAC-calculated phenomena and trends were also similar to those calculated with another one-dimensional PIUS model, the Brookhaven National Laboratory developed PIPA code. A TRACmore » pump-trip SCRAM transient has also been calculated with a TRAC model containing a multi-dimensional representation of the PIUS intemal flow structures and core region. The results obtained using the TRAC fully one-dimensional PIUS model are compared to the RIGEL, PIPA, and TRAC multi-dimensional results.« less

  11. Predictors of detection of alcohol use episodes using a transdermal alcohol sensor.

    PubMed

    Barnett, Nancy P; Meade, E B; Glynn, Tiffany R

    2014-02-01

    The objective of this investigation was to establish the ability of the Secure Continuous Remote Alcohol Monitoring (SCRAM) alcohol sensor to detect different levels of self-reported alcohol consumption, and to determine whether gender and body mass index, alcohol dependence, bracelet version, and age of bracelet influenced detection of alcohol use. Heavy drinking adults (N = 66, 46% female) wore the SCRAM for 1-28 days and reported their alcohol use in daily Web-based surveys. Participant reports of alcohol use were matched with drinking episodes identified from bracelet readings. On days when bracelets were functional, 690 drinking episodes were reported and 502 of those episodes (72.8%) were detected using sensor data. Using generalized estimating equations, we found no gender differences in detection of reported drinking episodes (77% for women, 69% for men). In univariate analyses, at the level of fewer than 5 drinks, women's episodes were more likely to be detected, likely because of the significantly higher transdermal alcohol concentration levels of these episodes, whereas at the level of 5 or more drinks, there was no gender difference in detection (92.6% for women, 93.4% for men). In multivariable analyses, no variables other than number of drinks significantly predicted alcohol detection. In summary, the SCRAM sensor is very good at detecting 5 or more drinks; performance of the monitor below this level was better among women because of their higher transdermal alcohol concentration levels. Individual person characteristics and bracelet features were not related to detection after number of drinks was included. Minimal bracelet malfunctions were noted.

  12. Contingency Management for Alcohol Use Reduction: A Pilot Study using a Transdermal Alcohol Sensor*

    PubMed Central

    Barnett, Nancy P.; Tidey, Jennifer; Murphy, James G.; Swift, Robert; Colby, Suzanne M.

    2011-01-01

    Background Contingency management (CM) has not been thoroughly evaluated as a treatment for alcohol abuse or dependence, in part because verification of alcohol use reduction requires frequent in-person breath tests. Transdermal alcohol sensors detect alcohol regularly throughout the day, providing remote monitoring and allowing for rapid reinforcement of reductions in use. Methods The purpose of this study was to evaluate the efficacy of CM for reduction in alcohol use, using a transdermal alcohol sensor to provide a continuous measure of alcohol use. Participants were 13 heavy drinking adults who wore the Secure Continuous Remote Alcohol Monitoring (SCRAM) bracelet for three weeks and provided reports of alcohol and drug use using daily web-based surveys. In Week 1, participants were asked to drink as usual; in Weeks 2 and 3, they were reinforced on an escalating schedule with values ranging from $5-$17 per day on days when alcohol use was not reported or detected by the SCRAM. Results Self-reports of percent days abstinent and drinks per week, and transdermal measures of average and peak transdermal alcohol concentration and area under the curve declined significantly in Weeks 2-3. A nonsignificant but large effect size for reduction in days of tobacco use also was found. An adjustment to the SCRAM criteria for detecting alcohol use provided an accurate but less conservative method for use with non-mandated clients. Conclusion Results support the efficacy of CM for alcohol use reductions and the feasibility of using transdermal monitoring of alcohol use for clinical purposes. PMID:21665385

  13. Predictors of Detection of Alcohol Use Episodes Using a Transdermal Alcohol Sensor

    PubMed Central

    Barnett, Nancy P.; Meade, E.B.; Glynn, Tiffany R.

    2014-01-01

    The objective of this investigation was to establish the ability of the Secure Continuous Remote Alcohol Monitoring (SCRAM) alcohol sensor to detect different levels of self-reported alcohol consumption, and to determine whether gender and body mass index, alcohol dependence, bracelet version, and age of bracelet influenced detection of alcohol use. Method Heavy drinking adults (N = 66; 46% female) wore the SCRAM for 1–28 days and reported their alcohol use in daily web-based surveys. Participant reports of alcohol use were matched with drinking episodes identified from bracelet readings. Results On days when bracelets were functional, 690 drinking episodes were reported and 502 of those episodes (72.8%) were detected using sensor data. Using Generalized Estimating Equations, we found no gender differences in detection of reported drinking episodes (77% for women, 69% for men). In univariate analyses, at the level of fewer than five drinks, women’s episodes were more likely to be detected, likely due to the significantly higher TAC levels of these episodes, whereas at the level of five or more drinks, there was no gender difference in detection (92.6% for women, 93.4% of men’s). In multivariable analyses, no variables other than number of drinks significantly predicted alcohol detection. Conclusion The SCRAM sensor is very good at detecting five or more drinks; performance of the monitor below this level was better among women due to their higher TAC levels. Individual person characteristics and bracelet features were not related to detection after number of drinks was included. Minimal bracelet malfunctions were noted. PMID:24490713

  14. Contingency management for alcohol use reduction: a pilot study using a transdermal alcohol sensor.

    PubMed

    Barnett, Nancy P; Tidey, Jennifer; Murphy, James G; Swift, Robert; Colby, Suzanne M

    2011-11-01

    Contingency management (CM) has not been thoroughly evaluated as a treatment for alcohol abuse or dependence, in part because verification of alcohol use reduction requires frequent in-person breath tests. Transdermal alcohol sensors detect alcohol regularly throughout the day, providing remote monitoring and allowing for rapid reinforcement of reductions in use. The purpose of this study was to evaluate the efficacy of CM for reduction in alcohol use, using a transdermal alcohol sensor to provide a continuous measure of alcohol use. Participants were 13 heavy drinking adults who wore the Secure Continuous Remote Alcohol Monitoring (SCRAM) bracelet for three weeks and provided reports of alcohol and drug use using daily web-based surveys. In Week 1, participants were asked to drink as usual; in Weeks 2 and 3, they were reinforced on an escalating schedule with values ranging from $5 to $17 per day on days when alcohol use was not reported or detected by the SCRAM. Self-reports of percent days abstinent and drinks per week, and transdermal measures of average and peak transdermal alcohol concentration and area under the curve declined significantly in Weeks 2-3. A nonsignificant but large effect size for reduction in days of tobacco use also was found. An adjustment to the SCRAM criteria for detecting alcohol use provided an accurate but less conservative method for use with non-mandated clients. Results support the efficacy of CM for alcohol use reductions and the feasibility of using transdermal monitoring of alcohol use for clinical purposes. Copyright © 2011 Elsevier Ireland Ltd. All rights reserved.

  15. Comparative study and evaluation of scram use, recidivism rates, and characteristics : traffic tech.

    DOT National Transportation Integrated Search

    2015-04-01

    Impaired driving continues to cause hundreds of thousands of alcohol-related crashes each year, many resulting in serious injury or death. Many offenders are repeat offenders despite sanctions and court processes that attempt to dissuade offenders fr...

  16. Section 7 reactor incident file general information from 1945

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Not Available

    1969-01-10

    At 0308 on January 10, 1966, both B and C Reactors ``scrammed`` due to an electrical fault on Line C2-L8 caused by a raccoon coming in contact with the 13-8 KV line on top of transformer No. 2 at 182-B Building. Line C2-L8 relayed out at the 151-B Building. Details of the occurrence at 151-B are covered in the attachment. C-Reactor scrammed due to reduced voltage on the pressure monitor system. The reduction in voltage caused the auxiliary relays of the pressure monitor ground detector to open, de-energizing the end result relays PSR and PSRA. The safety circuit trip identificationmore » system displayed ``Pressure Monitor`` and ``Ground Detector.`` B-Reactor scrammed by a power failure signal from 190-B Building. The power failure relays for pump numbers 1 and 3 opened due to these pumps contributing power to the fault. The power failure relays at 190-B remained open long enough for the end result relays PF and PFA to open. Since these relays are timed delayed, 0.26 seconds, the power failure relays must have remained open at least that long. At the 190-B Building the steam turbines started due to the power failure relays for pump numbers 1 and 3 opening. The main process pumps remained stable and continued to supply normal flow to the reactor. Pumps were tripped from the line at 182-B and 183-B Buildings. The surge suppressors cycled normally and the turbine export pumps started as a result of low export line pressure. No power equipment was affected in C Area.« less

  17. Count-doubling time safety circuit

    DOEpatents

    Rusch, Gordon K.; Keefe, Donald J.; McDowell, William P.

    1981-01-01

    There is provided a nuclear reactor count-factor-increase time monitoring circuit which includes a pulse-type neutron detector, and means for counting the number of detected pulses during specific time periods. Counts are compared and the comparison is utilized to develop a reactor scram signal, if necessary.

  18. On formation of the asymptotic spectrum of delayed neutron emitters in measuring the VVER-1000 scram system effectiveness

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Shishkov, L. K., E-mail: slk@vver.kiae.ru; Zizin, M. N., E-mail: zizin_m@mail.ru

    The process of formation of an asymptotic distribution of the neutron flux density in the reactor systems after introducing different negative reactivities is considered. The impact of two factors after the reactivity introduction is evaluated: (1) nonuniformity of perturbation of core properties, on one hand, and (2) a sharp reduction in the density of prompt neutrons, which prevents the appearance of new delayed neutron emitters distributed in accordance with the “new” prompt neutron distribution, on the other hand. The results of calculations show that the errors of measuring the scram system effectiveness using the method of inverse solution of themore » kinetics equation are caused by the fact that, after the negative reactivity insertion, the sources of prompt and delayed neutrons have different spatial distributions. In the case of high negative reactivities, this difference remains while the system still has neutrons, which can be measured.« less

  19. Loss-of-flow-without-scram tests in Experimental Breeder Reactor-II and comparison with pretest predictions

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Chang, L.K.; Mohr, D.; Planchon, H.P.

    This article discusses a series of successful loss-of-flow-without-scram tests conducted in Experimental Breeder Reactor-II (EBR-II), a metal-fueled, sodium-cooled fast reactor. These May 1985 tests demonstrated the capability of the EBR to reduce reactor power passively during a loss of flow and to maintain reactor temperatures within bounds without any reliance on an active safety system. The tests were run from reduced power to ensure that temperatures could be maintained well below the fuel-clad eutectic temperature. Good agreement was found between selected test data and pretest predictions made with the EBR-II system analysis code NATDEMO and the hot channel analysis codemore » HOTCHAN. The article also discusses safety assessments of the tests as well as modifications required on the EBR-II reactor safety system for conducting required on the EBR-II reactor safety system for the conducting the tests.« less

  20. TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cheng L. Y.; Baek J.; Cuadra,A.

    2013-11-10

    A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic tomore » initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.« less

  1. BWR Anticipated Transients Without Scram Leading to Instability

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cheng L. Y.; Baek J.; Cuadra, A.

    2013-11-10

    Anticipated transients without scram (ATWS) in aboiling water reactor (BWR) were simulated in order to understand reactor response and determine the effectiveness of automatic and operator actions to mitigate this beyond-design-basis accident. The events of interest herein are initiated by a turbine trip when the reactor is operating in the expanded operating domainMELLLA+ [maximum extended load line limit plus]. In these events the reactor may initially be at up to 120% of the original licensed thermal power (OLTP) and at flow rates as low as 80% of rated.For these (and similar) ATWS events the concern isthat when the reactor powermore » decreases in response to a dual recirculation pump trip, the core will become unstable and large amplitude oscillations will begin. The occurrence of these power oscillations, if left unmitigated, may result in fuel damage, and the amplitude of the poweroscillations may hamper the effectiveness of the injection of dissolved neutron absorber through the standby liquid control system (SLCS).« less

  2. CONTROL IN NUCLEAR REACTORS

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bell, F.R.

    1963-01-16

    An arrangement was described for scramming a reactor in an emergency. Control rods were position adjusted by an electric motor and transmission. A locking system kept the control rods in position but was arranged to be released in an emergency to allow the rods to drop into their shutdown position. (C.E.S.)

  3. 77 FR 43374 - Biweekly Notice, Applications and Amendments to Facility Operating Licenses and Combined Licenses...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-24

    ... degrees Fahrenheit ([deg]F) as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while... [Limited Conditions of Operation] 3.10.1, Inservice Leak and Hydrostatic Testing Operation Using...

  4. Multidisciplinary Analysis of a Hypersonic Engine

    NASA Technical Reports Server (NTRS)

    Suresh, Ambady; Stewart, Mark

    2003-01-01

    The objective is to develop high fidelity tools that can influence ISTAR design In particular, tools for coupling Fluid-Thermal-Structural simulations RBCC/TBCC designers carefully balance aerodynamic, thermal, weight, & structural considerations; consistent multidisciplinary solutions reveal details (at modest cost) At Scram mode design point, simulations give details of inlet & combustor performance, thermal loads, structural deflections.

  5. Dual-Pump CARS Measurements in the University of Virginia's Dual-Mode Scramjet: Configuration "C"

    NASA Technical Reports Server (NTRS)

    Cutler, Andrew D.; Magnotti, Gaetano; Cantu, Luca; Gallo, Emanuela; Danehy, Paul M.; Rockwell, Robert; Goyne, Christopher; McDaniel, James

    2013-01-01

    Measurements have been conducted at the University of Virginia Supersonic Combustion Facility in configuration C of the dual-mode scramjet. This is a continuation of previously published works on configuration A. The scramjet is hydrogen fueled and operated at two equivalence ratios, one representative of the scram mode and the other of the ram mode. Dual-pump CARS was used to acquire the mole fractions of the major species as well as the rotational and vibrational temperatures of N2. Developments in methods and uncertainties in fitting CARS spectra for vibrational temperature are discussed. Mean quantities and the standard deviation of the turbulent fluctuations at multiple planes in the flow path are presented. In the scram case the combustion of fuel is completed before the end of the measurement domain, while for the ram case the measurement domain extends into the region where the flow is accelerating and combustion is almost completed. Higher vibrational than rotational temperature is observed in those parts of the hot combustion plume where there is substantial H2 (and hence chemical reaction) present.

  6. Transpiration Cooling Experiment

    NASA Technical Reports Server (NTRS)

    Song, Kyo D.; Ries, Heidi R.; Scotti, Stephen J.; Choi, Sang H.

    1997-01-01

    The transpiration cooling method was considered for a scram-jet engine to accommodate thermally the situation where a very high heat flux (200 Btu/sq. ft sec) from hydrogen fuel combustion process is imposed to the engine walls. In a scram-jet engine, a small portion of hydrogen fuel passes through the porous walls of the engine combustor to cool the engine walls and at the same time the rest passes along combustion chamber walls and is preheated. Such a regenerative system promises simultaneously cooling of engine combustor and preheating the cryogenic fuel. In the experiment, an optical heating method was used to provide a heat flux of 200 Btu/sq. ft sec to the cylindrical surface of a porous stainless steel specimen which carried helium gas. The cooling efficiencies by transpiration were studied for specimens with various porosity. The experiments of various test specimens under high heat flux have revealed a phenomenon that chokes the medium flow when passing through a porous structure. This research includes the analysis of the system and a scaling conversion study that interprets the results from helium into the case when hydrogen medium is used.

  7. 10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

  8. 10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

  9. 10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

  10. 10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

  11. 10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

  12. Structural Equation Model Approach to the Use of Response Times for Improving Estimation in Item Response Models

    ERIC Educational Resources Information Center

    Sen, Rohini

    2012-01-01

    In the last five decades, research on the uses of response time has extended into the field of psychometrics (Schnikpe & Scrams, 1999; van der Linden, 2006; van der Linden, 2007), where interest has centered around the usefulness of response time information in item calibration and person measurement within an item response theory. framework.…

  13. CRDM with separate SCRAM latch engagement and locking

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dodd, Christopher D.; DeSantis, Paul K.; Stambaugh, Kevin J.

    A control rod drive mechanism (CRDM) configured to latch onto the lifting rod of a control rod assembly and including separate latch engagement and latch holding mechanisms. A CRDM configured to latch onto the lifting rod of a control rod assembly and including a four-bar linkage closing the latch, wherein the four-bar linkage biases the latch closed under force of gravity.

  14. CONTROL ROD DRIVE MECHANISM FOR A NUCLEAR REACTOR

    DOEpatents

    Hawke, B.C.; Liederbach, F.J.; Lones, W.

    1963-05-14

    A lead-screw-type control rod drive featuring an electric motor and a fluid motor arranged to provide a selectably alternative driving means is described. The electric motor serves to drive the control rod slowly during normal operation, while the fluid motor, assisted by an automatic declutching of the electric motor, affords high-speed rod insertion during a scram. (AEC)

  15. Multidisciplinary Analysis of a Hypersonic Engine

    NASA Technical Reports Server (NTRS)

    Stewart, M. E. M.; Suresh, A.; Liou, M. S.; Owen, A. K.; Messitt, D. G.

    2002-01-01

    This paper describes implementation of a technique used to obtain a high fidelity fluid-thermal-structural solution of a combined cycle engine at its scram design point. Single-discipline simulations are insufficient here since interactions from other disciplines are significant. Using off-the-shelf, validated solvers for the fluid, chemistry, thermal, and structural solutions, this approach couples together their results to obtain consistent solutions.

  16. System studies on space plane powered by scram/LACE propulsion system

    NASA Astrophysics Data System (ADS)

    Maita, Masataka; Miyajima, Hiroshi; Mori, Takashige

    1992-12-01

    Japan's NAL has undertaken concept-development studies for hypersonic technologies-integrating SSTO spaceplane configurations. Attention is presently given to the scramjet/liquefied air cycle engine (LACE). While the scramjet powers the vehicle begining at Mach 5, the LACE is used above Mach 12 on the basis of excess hydrogen fuel consumption; the Mach 20 orbital speed is thereby gained.

  17. 75 FR 13786 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-23

    .../petitioner shall provide a brief explanation of the bases for the contention and a concise statement of the...? Response: No. The Bases of TS 3.6.2.2 state that the operability of the Spray Additive System ensures that... ``Scram Discharge Volume (SDV) Vent and Drain Valves'' and associated Bases of NUREG-1433, Revision 3...

  18. Damper mechanism for nuclear reactor control elements

    DOEpatents

    Taft, William Elwood

    1976-01-01

    A damper mechanism which provides a nuclear reactor control element decelerating function at the end of the scram stroke. The total damping function is produced by the combination of two assemblies, which operate in sequence. First, a tapered dashram assembly decelerates the control element to a lower velocity, after which a spring hydraulic damper assembly takes over to complete the final damping.

  19. EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS

    DOEpatents

    Paget, J.A.; Koutz, S.L.; Stone, R.S.; Stewart, H.B.

    1963-12-24

    An emergency shutdown or scram apparatus for use in a nuclear reactor that includes a neutron absorber suspended from a temperature responsive substance that is selected to fail at a preselected temperature in excess of the normal reactor operating temperature, whereby the neutron absorber is released and allowed to fall under gravity to a preselected position within the reactor core is presented. (AEC)

  20. Interactions between Flight Dynamics and Propulsion Systems of Air-Breathing Hypersonic Vehicles

    NASA Astrophysics Data System (ADS)

    Dalle, Derek J.

    The development and application of a first-principles-derived reduced-order model called MASIV (Michigan/AFRL Scramjet In Vehicle) for an air-breathing hypersonic vehicle is discussed. Several significant and previously unreported aspects of hypersonic flight are investigated. A fortunate coupling between increasing Mach number and decreasing angle of attack is shown to extend the range of operating conditions for a class of supersonic inlets. Detailed maps of isolator unstart and ram-to-scram transition are shown on the flight corridor map for the first time. In scram mode the airflow remains supersonic throughout the engine, while in ram mode there is a region of subsonic flow. Accurately predicting the transition between these two modes requires models for complex shock interactions, finite-rate chemistry, fuel-air mixing, pre-combustion shock trains, and thermal choking, which are incorporated into a unified framework here. Isolator unstart occurs when the pre-combustion shock train is longer than the isolator, which blocks airflow from entering the engine. Finally, cooptimization of the vehicle design and trajectory is discussed. An optimal control technique is introduced that greatly reduces the number of computations required to optimize the simulated trajectory.

  1. Scram recoveries---C Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Constable, D.W.; Pierce, J.R.; Wood, S.A.

    1962-04-26

    The purpose of this report is to discuss the observations made on two equilibrium scram recovery startups (April 5 and April 16). Normally, the two startups would have little significance but unusual ruptures were experienced in the top near section of the reactor shortly after both startups, which indicates that some similarity could exist between the two. The ruptures were unusual in that the two tubes involved both had multiple ruptures. One tube contained two E{sup 2} ruptures and the other tube contained three overbore metal ruptures. The overbore tube also contained three incipient ruptures (uranium split under the can).more » The initial rise to power on both startups appeared to be normal with the flux peaking on the near side as expected. On the April 16 startup the maximum level reached was 1050 at which time a rupture in overbore tube 3062 caused on increase in pressure resulting in a high trip on the Panellit gauge. A level of 1600 was reached on the April 5 startup which was held for approximately 14 hours at which time the reactor was shut down due to rupture indications on row 29.« less

  2. Center for Hypersonic Combined Cycle Flow Physics

    DTIC Science & Technology

    2015-03-24

    team of expert experimentalists and numerical and chemical kinetic modelers. Flowfields were examined in the turbine /ramjet dual inlet mode transition...using data from the NASA Glenn IMX facility and RANS calculations. In the ramjet/scramjet mode regime a dual-mode combustion wind tunnel was developed...SUBJECT TERMS Hypersonic combined cycle propulsion, turbine /ram dual-inlet transition, ram/scram dual-mode transition, hypervelocity regime, RANS, Hybrid

  3. Double-clad nuclear fuel safety rod

    DOEpatents

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  4. Double-clad nuclear-fuel safety rod

    DOEpatents

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  5. A Layman's Guide to Thrust Engine Development for Super/Hyper Sonic Flight.

    ERIC Educational Resources Information Center

    Thiesse, James L.

    The intention of this paper is to discuss the advances in thrust engines from the initial development of the J58/SR-71 (JT11D-20) of the U.S. Air Force's SR-71 Blackbird to the development of the RAM and SCRAM engines necessary to propel the new generations of high-flying super-speed aircraft. Engineering complexities suggest that the engines and…

  6. A Priori Analysis of Flamelet-Based Modeling for a Dual-Mode Scramjet Combustor

    NASA Technical Reports Server (NTRS)

    Quinlan, Jesse R.; McDaniel, James C.; Drozda, Tomasz G.; Lacaze, Guilhem; Oefelein, Joseph

    2014-01-01

    An a priori investigation of the applicability of flamelet-based combustion models to dual-mode scramjet combustion was performed utilizing Reynolds-averaged simulations (RAS). For this purpose, the HIFiRE Direct Connect Rig (HDCR) flowpath, fueled with a JP-7 fuel surrogate and operating in dual- and scram-mode was considered. The chemistry of the JP-7 fuel surrogate was modeled using a 22 species, 18-step chemical reaction mechanism. Simulation results were compared to experimentally-obtained, time-averaged, wall pressure measurements to validate the RAS solutions. The analysis of the dual-mode operation of this flowpath showed regions of predominately non-premixed, high-Damkohler number, combustion. Regions of premixed combustion were also present but associated with only a small fraction of the total heat-release in the flow. This is in contrast to the scram-mode operation, where a comparable amount of heat is released from non-premixed and premixed combustion modes. Representative flamelet boundary conditions were estimated by analyzing probability density functions for temperature and pressure for pure fuel and oxidizer conditions. The results of the present study reveal the potential for a flamelet model to accurately model the combustion processes in the HDCR and likely other high-speed flowpaths of engineering interest.

  7. Inverse design of a proper number, shapes, sizes, and locations of coolant flow passages

    NASA Technical Reports Server (NTRS)

    Dulikravich, George S.

    1992-01-01

    During the past several years we have developed an inverse method that allows a thermal cooling system designer to determine proper sizes, shapes, and locations of coolant passages (holes) in, say, an internally cooled turbine blade, a scram jet strut, a rocket chamber wall, etc. Using this method the designer can enforce a desired heat flux distribution on the hot outer surface of the object, while simultaneously enforcing desired temperature distributions on the same hot outer surface as well as on the cooled interior surfaces of each of the coolant passages. This constitutes an over-specified problem which is solved by allowing the number, sizes, locations and shapes of the holes to adjust iteratively until the final internally cooled configuration satisfies the over-specified surface thermal conditions and the governing equation for the steady temperature field. The problem is solved by minimizing an error function expressing the difference between the specified and the computed hot surface heat fluxes. The temperature field analysis was performed using our highly accurate boundary integral element code with linearly varying temperature along straight surface panels. Examples of the inverse design applied to internally cooled turbine blades and scram jet struts (coated and non-coated) having circular and non-circular coolant flow passages will be shown.

  8. Nonlinear process in the mode transition in typical strut-based and cavity-strut based scramjet combustors

    NASA Astrophysics Data System (ADS)

    Yan, Li; Liao, Lei; Huang, Wei; Li, Lang-quan

    2018-04-01

    The analysis of nonlinear characteristics and control of mode transition process is the crucial issue to enhance the stability and reliability of the dual-mode scramjet engine. In the current study, the mode transition processes in both strut-based combustor and cavity-strut based combustor are numerically studied, and the influence of the cavity on the transition process is analyzed in detail. The simulations are conducted by means of the Reynolds averaged Navier-Stokes (RANS) equations coupled with the renormalization group (RNG) k-ε turbulence model and the single-step chemical reaction mechanism, and this numerical approach is proved to be valid by comparing the predicted results with the available experimental shadowgraphs in the open literature. During the mode transition process, an obvious nonlinear property is observed, namely the unevenly variations of pressure along the combustor. The hysteresis phenomenon is more obvious upstream of the flow field. For the cavity-strut configuration, the whole flow field is more inclined to the supersonic state during the transition process, and it is uneasy to convert to the ramjet mode. In the scram-to-ram transition process, the process would be more stable, and the hysteresis effect would be reduced in the ram-to-scram transition process.

  9. LOGIC CIRCUIT

    DOEpatents

    Strong, G.H.; Faught, M.L.

    1963-12-24

    A device for safety rod counting in a nuclear reactor is described. A Wheatstone bridge circuit is adapted to prevent de-energizing the hopper coils of a ball backup system if safety rods, sufficient in total control effect, properly enter the reactor core to effect shut down. A plurality of resistances form one arm of the bridge, each resistance being associated with a particular safety rod and weighted in value according to the control effect of the particular safety rod. Switching means are used to switch each of the resistances in and out of the bridge circuit responsive to the presence of a particular safety rod in its effective position in the reactor core and responsive to the attainment of a predetermined velocity by a particular safety rod enroute to its effective position. The bridge is unbalanced in one direction during normal reactor operation prior to the generation of a scram signal and the switching means and resistances are adapted to unbalance the bridge in the opposite direction if the safety rods produce a predetermined amount of control effect in response to the scram signal. The bridge unbalance reversal is then utilized to prevent the actuation of the ball backup system, or, conversely, a failure of the safety rods to produce the predetermined effect produces no unbalance reversal and the ball backup system is actuated. (AEC)

  10. Nuclear Energy in Southeast Asia: Pull Rods or Scram

    DTIC Science & Technology

    2009-06-01

    December 29, 2008);  Seth Mydans, “Tens of thousands join Myanmar protest,” International Harold Tribune, September 24, 2007, http://www.iht.com...articles/2007/09/24/news/myanmar.php (accessed December 29, 2008); Seth Mydans; “Myanmar monk protest contained by Junta forces,” The New York Times...Nuclear Plant for Electricity.” Associated Press, September 26, 2008. http://www.ap.org (accessed October 20, 2008). Mydans, Seth . “Myanmar monk

  11. Technologies for Propelled Hypersonic Flight Volume 2 - Subgroup 2: Scram Propulsion

    DTIC Science & Technology

    2006-01-01

    effort is focused on the MSD code, initially developed by ONERA to simulate internal aerodynamic flows, which has been upgraded in cooperation...inlets were studied: a mixed, external/ internal , compression inlet studied at DLR with testing in the H2K and TMK wind-tunnels, and an internal ...movable panels during operation along the trajectory, modification of the internal geometry by a control-command computer connected with sensors on the

  12. Period meter for reactors

    DOEpatents

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  13. Numerical Modeling and Combustion Studies of Scram Jet Simulation

    DTIC Science & Technology

    2014-12-01

    and this work is dedicated to them. xiii Chapter 1 1 Introduction 1.1 Background and Overview Scramjet ( Supersonic Combustion Ramjet) is a type of...engine that op- erates under supersonic airflow conditions. The efficiency in its propulsion system over ramjet has made it a very active research...from the boundary layer of the wall [41]. Moreover, when the crossflow is supersonic , as is the case in the Scramjet configuration, some additional

  14. Review of Findings for Human Performance Contribution to Risk in Operating Events

    DTIC Science & Technology

    2002-03-01

    and loss of DC power. Key to this event was failure to control setpoints on safety-related equipment and failure to maintain the load tap changer...34 Therefore, "to optimize task execution at the job site, it is important to align organizational processes and values." Effective team skills are an...reactor was blocked and the water level rapidly dropped to the automatic low-level scram setpoint . Human Performance Issues Control rods were fully

  15. Effect of twice quenching and tempering on the mechanical properties and microstructures of SCRAM steel for fusion application

    NASA Astrophysics Data System (ADS)

    Xiong, Xuesong; Yang, Feng; Zou, Xingrong; Suo, Jinping

    2012-11-01

    The effect of twice quenching and tempering on the mechanical properties and microstructures of SCRAM steel was investigated. The results from tensile tests showed that whether twice quenching and tempering processes(1253 K/0.5 h/W.C(water cool) + 1033 K/2 h/A.C(air cool) + 1233 K/0.5 h/W.C + 1033 K/2 h/A.C named after 2Q&2TI, and 1253 K/0.5 h/W.C + 1033 K/2 h/A.C + 1233 K/0.5 h/W.C + 1013 K/2 h/A.C named after 2Q&2TII)increased strength of steel or not depended largely on the second tempering temperature compared to quenching and tempering process(1253 K/0.5 h/W.C + 1033 K/2 h/A.C named after 1Q&1T). Charpy V-notch impact tests indicated that twice quenching and tempering processes reduced the ductile brittle transition temperature (DBTT). Microstructure inspection revealed that the prior austenitic grain size and martensite lath width were refined after twice quenching and tempering treatments. Precipitate growth was inhibited by a slight decrease of the second tempering temperature from 1033 to 1013 K. The finer average size of precipitates is considered to be the main possible reason for the higher strength and lower DBTT of 2Q&2TII compared with 2Q&2TI.

  16. Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

    NASA Astrophysics Data System (ADS)

    Susyadi; Endiah, P. H.; Sukmanto, D.; Andi, S. E.; Syaiful, B.; Hendro, T.; Geni, R. S.

    2018-02-01

    RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

  17. Hypersonic Combined Cycle Propulsion Panel Symposium (75th) Held in Madrid, Spain on 28 May - 1 June 1990 (La Propulsion Hypersonique a Cycles Combines)

    DTIC Science & Technology

    1990-12-01

    mutually exclusive. That is, they may be utilized simultaneously to compound the additive refrigerative enhancement effect. The Recycled ScramLACE (Figure...small positive reaction (say 10%) in order to obviate diffusion. Impulse stages can be velocity compounded , an arrangement in which a large pressure...with more effective seals. Conceptually, it is possible to design a series of velocity compounded stages to run in tandem to give the correct overall

  18. NUMBER AND TYPE OF OPERATING CYCLES FOR THE FFTF

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Boyd, D. C.

    1969-05-15

    The choice of materials and other vessel design decisions necessary to provide the desired life expectancy for the FTR vessel are partially dependent upon estimates of the number and type of reactor shutdowns and startups which may be anticipated. Current estimates of these so-called "cycles" are given, including scram frequency, experimental outage frequency, standard shutdowns and startups, and rapid controlled shutdowns. Also discussed are abnormal heatup or cooldown, and tentative goals for temperature controls. MTR, ETR, and typical PRTR operating histories are tabulated.

  19. SRGULL - AN ADVANCED ENGINEERING MODEL FOR THE PREDICTION OF AIRFRAME INTEGRATED SCRAMJET CYCLE PERFORMANCE

    NASA Technical Reports Server (NTRS)

    Walton, J. T.

    1994-01-01

    The development of a single-stage-to-orbit aerospace vehicle intended to be launched horizontally into low Earth orbit, such as the National Aero-Space Plane (NASP), has concentrated on the use of the supersonic combustion ramjet (scramjet) propulsion cycle. SRGULL, a scramjet cycle analysis code, is an engineer's tool capable of nose-to-tail, hydrogen-fueled, airframe-integrated scramjet simulation in a real gas flow with equilibrium thermodynamic properties. This program facilitates initial estimates of scramjet cycle performance by linking a two-dimensional forebody, inlet and nozzle code with a one-dimensional combustor code. Five computer codes (SCRAM, SEAGUL, INLET, Progam HUD, and GASH) originally developed at NASA Langley Research Center in support of hypersonic technology are integrated in this program to analyze changing flow conditions. The one-dimensional combustor code is based on the combustor subroutine from SCRAM and the two-dimensional coding is based on an inviscid Euler program (SEAGUL). Kinetic energy efficiency input for sidewall area variation modeling can be calculated by the INLET program code. At the completion of inviscid component analysis, Program HUD, an integral boundary layer code based on the Spaulding-Chi method, is applied to determine the friction coefficient which is then used in a modified Reynolds Analogy to calculate heat transfer. Real gas flow properties such as flow composition, enthalpy, entropy, and density are calculated by the subroutine GASH. Combustor input conditions are taken from one-dimensionalizing the two-dimensional inlet exit flow. The SEAGUL portions of this program are limited to supersonic flows, but the combustor (SCRAM) section can handle supersonic and dual-mode operation. SRGULL has been compared to scramjet engine tests with excellent results. SRGULL was written in FORTRAN 77 on an IBM PC compatible using IBM's FORTRAN/2 or Microway's NDP386 F77 compiler. The program is fully user interactive, but can also run in batch mode. It operates under the UNIX, VMS, NOS, and DOS operating systems. The source code is not directly compatible with all PC compilers (e.g., Lahey or Microsoft FORTRAN) due to block and segment size requirements. SRGULL executable code requires about 490K RAM and a math coprocessor on PC's. The SRGULL program was developed in 1989, although the component programs originated in the 1960's and 1970's. IBM, IBM PC, and DOS are registered trademarks of International Business Machines. VMS is a registered trademark of Digital Equipment Corporation. UNIX is a registered trademark of Bell Laboratories. NOS is a registered trademark of Control Data Corporation.

  20. Computations in turbulent flows and off-design performance predictions for airframe-integrated scramjets

    NASA Technical Reports Server (NTRS)

    Goglia, G. L.; Spiegler, E.

    1977-01-01

    The research activity focused on two main tasks: (1) the further development of the SCRAM program and, in particular, the addition of a procedure for modeling the mechanism of the internal adjustment process of the flow, in response to the imposed thermal load across the combustor and (2) the development of a numerical code for the computation of the variation of concentrations throughout a turbulent field, where finite-rate reactions occur. The code also includes an estimation of the effect of the phenomenon called 'unmixedness'.

  1. High Energy Absorption Top Nozzle For A Nuclaer Fuel Assembly

    DOEpatents

    Sparrow, James A.; Aleshin, Yuriy; Slyeptsov, Aleksey

    2004-05-18

    A high energy absorption top nozzle for a nuclear fuel assembly that employs an elongated upper tubular housing and an elongated lower tubular housing slidable within the upper tubular housing. The upper and lower housings are biased away from each other by a plurality of longitudinally extending springs that are restrained by a longitudinally moveable piston whose upward travel is limited within the upper housing. The energy imparted to the nozzle by a control rod scram is mostly absorbed by the springs and the hydraulic affect of the piston within the nozzle.

  2. Assured crew return vehicle post landing configuration design and test

    NASA Technical Reports Server (NTRS)

    Anderson, Loren A.; Armitage, Pamela Kay

    1992-01-01

    The 1991-1992 senior Mechanical and Aerospace Engineering Design class continued work on the post landing configurations for the Assured Crew Return Vehicle (ACRV) and the Emergency Egress Couch (EEC). The ACRV will be permanently docked to Space Station Freedom, fulfilling NASA's commitment of Assured Crew Return Capability in the event of an accident or illness aboard Space Station Freedom. The EEC provides medical support and a transportation surface for an incapacitated crew member. The objective of the projects was to give the ACRV Project Office data to feed into their feasibility studies. Four design teams were given the task of developing models with dynamically and geometrically scaled characteristics. Groups one and two combined effort to design a one-fifth scale model of the Apollo Command Module derivative, an on-board flotation system, and a lift attachment point system. This model was designed to test the feasibility of a rigid flotation and stabilization system and to determine the dynamics associated with lifting the vehicle during retrieval. However, due to priorities, it was not built. Group three designed a one-fifth scale model of the Johnson Space Center (JSC) benchmark configuration, the Station Crew Return Alternative Module (SCRAM) with a lift attachment point system. This model helped to determine the flotation and lifting characteristics of the SCRAM configuration. Group four designed a full scale EEC with changeable geometric and dynamic characteristics. This model provided data on the geometric characteristics of the EEC and on the placement of the CG and moment of inertia. It also gave the helicopter rescue personnel direct input to the feasibility study.

  3. Selective Blockade of Cytoskeletal Actin Remodeling Reduces Experimental Choroidal Neovascularization

    PubMed Central

    Caballero, Sergio; Yang, Ru; Chaqour, Brahim

    2011-01-01

    Purpose. The efficacy of the peptide Ac-EEED on reducing cell adhesion and proliferation in vitro and choroidal neovascularization (CNV) in vivo was examined. Methods. The peptide chimera containing the Ac-EEED sequence was chemically linked to the N terminus of the XMTM delivery peptide from the Erns viral surface protein. Ac-EEED or scrambled control peptide (SCRAM) was added to cultures of vascular smooth muscle cells, pericytes, endothelial cells, and fibroblasts, and adhesion, growth, and matrix production was assessed. Ac-EEED or SCRAM was injected into the vitreous of mice undergoing laser rupture of Bruch's membrane to induce CNV and lesion volume, neovascularization and lesion fibrosis were assessed. Results. Ac-EEED–induced changes in the morphology of the actin cytoskeleton by inhibiting polymerization of G-actin and disrupting the formation of stress fibers. Pretreatment with Ac-EEED resulted in endothelial cells becoming less responsive to the mitogenic and pro-adhesive effects of VEGF. Ac-EEED treatment in fibroblasts reduced TGF-β–induced fibrosis as assessed by decreased levels of connective tissue growth factor, cysteine-rich 61, collagen I (COL1A2), and collagen III (COL3A1). CNV lesion size and fibrosis were reduced in a concentration-dependent manner by up to 60%. Conclusions. In vitro studies showed that Ac-EEED affects a broad range of mechanical properties associated with cytoskeletal actin to reduce growth factor effects. The utilization of Ac-EEED in vivo may offer a novel therapeutic strategy by both suppressed neovessel growth and curtailing fibrosis typically associated with the involutional stage of CNV. PMID:21178140

  4. Rectifier cabinet static breaker

    DOEpatents

    Costantino, Jr, Roger A.; Gliebe, Ronald J.

    1992-09-01

    A rectifier cabinet static breaker replaces a blocking diode pair with an SCR and the installation of a power transistor in parallel with the latch contactor to commutate the SCR to the off state. The SCR serves as a static breaker with fast turnoff capability providing an alternative way of achieving reactor scram in addition to performing the function of the replaced blocking diodes. The control circuitry for the rectifier cabinet static breaker includes on-line test capability and an LED indicator light to denote successful test completion. Current limit circuitry provides high-speed protection in the event of overload.

  5. Analysis of the OPERA 15-pin experiment with SABRE-2P. [LMFBR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Rose, S.D.; Carbajo, J.J.

    The OPERA (Out-of-Pile Expulsion and Reentry Apparatus) experiment simulates the initial phase of a pump coastdown without scram of a liquid-metal fast breeder reactor, specifically the Fast Flux Test Facility. The test section is a 15-pin 60/sup 0/ triangular sector designed to simulate a full-size 61-pin hexagonal bundle. A previous study indicates this to be an adequate simulation. In this paper, experimental results from the OPERA 15-pin experiment performed at ANL in 1982 are compared to analytical calculations obtained with the SABRE-2P code at ORNL.

  6. Nuclear reactor control apparatus

    DOEpatents

    Sridhar, Bettadapur N.

    1983-11-01

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  7. G T-Mohr Start-up Reactivity Insertion Transient Analysis Using Simulink

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Fard, Mehdi Reisi; Blue, Thomas E.; Miller, Don W.

    2006-07-01

    As a part of a Department of Energy-Nuclear Engineering Research Initiative (NERI) Project, we at OSU are investigating SiC semiconductor detectors as neutron power monitors for Generation IV power reactors. As a part of this project, we are investigating the power monitoring requirements for a specific type of Generation IV reactor, namely the GT-MHR. To evaluate the power monitoring requirements for the GT-MHR that are most demanding for a SiC diode power monitor, we have developed a Simulink model to study the transient behavior of the GT-MHR. In this paper, we describe the application of the Simulink code to themore » analysis of a series of Start-up Reactivity Insertion Transients (SURITs). The SURIT is considered to be a limiting protectable accident in terms of establishing the dynamic range of a SiC power monitor because of the low count rate of the detector during the start-up and absence of the reactivity feedback mechanism at the beginning of transient. The SURIT is studied with the ultimate goal of identifying combinations of 1) reactor power scram setpoints and 2) cram initiation times (the time in which a scram must be initiated once the setpoint is exceeded) for which the GT-MHR core is protected in the event of a continuous withdrawal of a control rod bank from the core from low powers. The SURIT is initiated by withdrawing a rod bank when the reactor is cold (300 K) and sub-critical at the BOEC (Beginning of Equilibrium Cycle) condition. Various initial power levels have been considered corresponding to various degrees of sub-criticality and various source strengths. An envelope of response is determined to establish which initial powers correspond to the worst case SURIT. (authors)« less

  8. Numerical investigations in three-dimensional internal flows

    NASA Technical Reports Server (NTRS)

    Rose, William C.

    1991-01-01

    In previous efforts, a two-dimensional full Navier-Stokes (FNS) code (SCRAM2D) was used in a design process that involved parametric modifications of the inlet geometry to arrive at what appeared to be an optimum inlet flowfield that produced a uniform flow at the exit in a very short distance. In these previous studies, the technologies for determining the contours with a 'man-in-the-loop' approach for both the ramp and cowl of the inlet were demonstrated, and nearly shock-free exiting flowfields were shown to be obtainable. The resulting two-dimensional compression contours were then used with swept sidewalls to form a three-dimensional inlet. Then the three-dimensional Navier-Stokes code (SCRAM3D) was used to investigate the inlet's three-dimensional flow. One of the major difficulties encountered in the previous studies was that associated with the relatively long time required to obtain a solution using even the 2D FNS code in the design process. Since one of the goals of high-speed inlet design is to produce inputs to the overall aircraft design in a timely manner, it was proposed for this year's research to examine 2D and 3D viscous flow solver techniques alternative to the NFS codes used to date. Areas of the inlet particularly identified for code speed up are those associated with the forebody and external flow ramp systems of the inlet. In these areas, parabolized, or space-marched, Navier-Stokes codes were proposed to be investigated for their applicability in the design process developed previously. This report describes the results of an investigation into the use of two other codes for analyzing the forebody and inlet ramp systems of high-speed inlets.

  9. Sensitivity to VSR failure: K pipe break accident

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Meichle, R.H.

    1969-09-12

    Reactor effects of failure of a safety rod to scram can be considered in two major respects: The reduction in total safety system strength which will affect the amount of ``prompt drop`` and subsequent flux decay rate of the average neutron flux-level; and the change in local flux distribution due to the absence of the particular rod which fails to enter the reactor. The purpose of this memorandum is to describe the physical effects involved and to indicate the approximate magnitude of both reactor-wide and localized changes in event of failure of a VSR simultaneous with a K Reactor risermore » accident.« less

  10. Fast Flux Test Facility thermal and pressure transient events during Cycle 11

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ahrens, D. M.

    1992-03-01

    This report documents the thermal and pressure transients experienced by the Reactor Heat Transport System (RHTS) during Cycle 11 which included Cycles 11A, 11B-1, 11B-2 and 11C (i.e. 4 startups and 4 shutdowns). Cycle 11 consisted of a refueling period that began on March 14, 1989 and power operation which began on May 3, 1989 and ended on October 27, 1990. Transients resulted from secondary pump starts/stops while at refueling conditions. The major causes of transients at power were five unplanned reactor scrams from 100% power and problems with Loop 2 DHX Fan Controls During 11A.

  11. 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)

  12. JEN-1 Reactor Control System; SISTEMA DE CONTROL DEL REACTOR JEN-1

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cantillo, M.F.; Nuno, C.M.; Andreu, J.L.M.

    1963-01-01

    ABS>The JEN-1 3Mw power swimming pool reactor electrical control circuits are described. Start-up, power generation in the core, and shutdown are controlled by the reactor control system. This control system guarantees in each moment the safety conditions during reactor operation. Each circuit was represented by a scheme, complemented with a description of its function, components, and operation theory. Components described include: scram circuit; fission counter control circuit; servo control circuit; control circuit of safety sheets; control circuits of primary, secondary, and clean-up pump motors and tower fan motor; primary valve motor circuit; center cubicle alarm circuit; and process alarm circuit.more » (auth)« less

  13. Fission product transport analysis in a loss of decay heat removal accident at Browns Ferry

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wichner, R.P.; Weber, C.F.; Hodge, S.A.

    1984-01-01

    This paper summarizes an analysis of the movement of noble gases, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal (DHR) capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris ontomore » the drywell floor.« less

  14. Reactor transient control in support of PFR/TREAT TUCOP experiments

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Burrows, D.R.; Larsen, G.R.; Harrison, L.J.

    1984-01-01

    Unique energy deposition and experiment control requirements posed bythe PFR/TREAT series of transient undercooling/overpower (TUCOP) experiments resulted in equally unique TREAT reactor operations. New reactor control computer algorithms were written and used with the TREAT reactor control computer system to perform such functions as early power burst generation (based on test train flow conditions), burst generation produced by a step insertion of reactivity following a controlled power ramp, and shutdown (SCRAM) initiators based on both test train conditions and energy deposition. Specialized hardware was constructed to simulate test train inputs to the control computer system so that computer algorithms couldmore » be tested in real time without irradiating the experiment.« less

  15. NASA thrusts in high-speed aeropropulsion research and development: An overview

    NASA Technical Reports Server (NTRS)

    Ziemianski, Joseph A.

    1990-01-01

    NASA is conducting aeronautical research over a broad range of Mach numbers. In addition to the advanced conventional takeoff or landing (CTOL) propulsion research described elsewhere, NASA Lewis has intensified its efforts towards propulsion technology for selected high speed flight applications. In a companion program, NASA Langley has also accomplished significant research in supersonic combustion ramjet (SCRAM) propulsion. An unclassified review is presented of the propulsion research results that are applicable for supersonic to hypersonic vehicles. This overview not only provides a preview of the more detailed presentations which follow, it also presents a viewpoint on future research directions by calling attention to the unique cycles, components, and facilities involved in this expanding area of work.

  16. ``Sleeping reactor`` irradiations: Shutdown reactor determination of short-lived activation products

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jerde, E.A.; Glasgow, D.C.

    1998-09-01

    At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux ({phi}) of {approximately} 4 {times} 10{sup 14} n/cm{sup 2} {center_dot} s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of {approximately} 6 s, but the requirement of immediate countingmore » leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about {+-} 0.5 s) make irradiations of < 6 s less reliable. Therefore, the determination of these ultra-short-lived species in mixed matrices has not generally been made at HFIR. The authors have found that very short lived activation products can be produced easily during the period after reactor shutdown (SCRAM), but prior to the removal of spent fuel elements. During this 24- to 36-h period (dubbed the ``sleeping reactor``), neutrons are produced in the beryllium reflector by the reaction {sup 9}Be({gamma},n){sup 8}Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to {approximately} 1 {times} 10{sup 10} n/cm{sup 2} {center_dot} s within 1 h. By the time the fuel elements are removed, the flux has dropped to {approximately} 6 {times} 10{sup 8}. Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant.« less

  17. Transpiring Cooling of a Scram-Jet Engine Combustion Chamber

    NASA Technical Reports Server (NTRS)

    Choi, Sang H.; Scotti, Stephen J.; Song, Kyo D.; Ries,Heidi

    1997-01-01

    The peak cold-wall heating rate generated in a combustion chamber of a scram-jet engine can exceed 2000 Btu/sq ft sec (approx. 2344 W/sq cm). Therefore, a very effective heat dissipation mechanism is required to sustain such a high heating load. This research focused on the transpiration cooling mechanism that appears to be a promising approach to remove a large amount of heat from the engine wall. The transpiration cooling mechanism has two aspects. First, initial computations suggest that there is a reduction, as much as 75%, in the heat flux incident on the combustion chamber wall due to the transpirant modifying the combustor boundary layer. Secondly, the heat reaching the combustor wall is removed from the structure in a very effective manner by the transpirant. It is the second of these two mechanisms that is investigated experimentally in the subject paper. A transpiration cooling experiment using a radiant heating method, that provided a heat flux as high as 200 Btu/sq ft sec ( approx. 234 W/sq cm) on the surface of a specimen, was performed. The experiment utilized an arc-lamp facility (60-kW radiant power output) to provide a uniform heat flux to a test specimen. For safety reasons, helium gas was used as the transpirant in the experiments. The specimens were 1.9-cm diameter sintered, powdered-stainless-steel tubes of various porosities and a 2.54cm square tube with perforated multi-layered walls. A 15-cm portion of each specimen was heated. The cooling effectivenes and efficiencies by transpiration for each specimen were obtained using the experimental results. During the testing, various test specimens displayed a choking phenomenon in which the transpirant flow was limited as the heat flux was increased. The paper includes a preliminary analysis of the transpiration cooling mechanism and a scaling conversion study that translates the results from helium tests into the case when a hydrogen medium is used.

  18. 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.

  19. 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.

  20. Aeropropulsion 1987. Session 6: High-Speed Propulsion Technology

    NASA Technical Reports Server (NTRS)

    1987-01-01

    NASA is conducting aeronautical research over a broad range of Mach numbers. In addition to the advanced CTOL propulsion research described in a separate session, the Lewis Research Center has intensified its efforts towards propulsion technology for selected high-speed flight applications. In a companion program, the Langley Research Center has also accomplished excellent research in Supersonic Combustion Ramjet (SCRAM) propulsion. What is presented in this session is an unclassified review of some of the propulsion research results that are applicable for supersonic to hypersonic vehicles. Not only is a review provided for several key work areas, it also presents a viewpoint on future research directions by calling attention to cycles, components, and facilities involved in this rapidly expanding field of work.

  1. Simulation of sodium pumps for nuclear power plants. Technical report 1 Oct 80-1 May 81

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Boadu, H.O.

    1981-05-01

    A single-phase pump model for analysis of transients in sodium cooled fast breeder nuclear power plants has been presented, where homologous characteristic curves are used to predict the behavior of the pump during operating transients. The pump model has been incorporated into BRENDA and FFTF; two system cases to simulate Clinch River Breeder Reactor Plant (CRBRP) and the Fast Flux Test Facility (FFTF) respectively. Two simulation test results for BRENDA which is one loop representation of a three loop plant have been presented. They are: (1) Primary pump coastdown to natural circulation coupled with scram failure, and (2) 10 percentmore » deviation of primary speed with plant controllers incorporated.« less

  2. Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Feltus, M.A.

    1989-11-01

    The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

  3. Assured Crew Return Vehicle

    NASA Technical Reports Server (NTRS)

    Stone, D. A.; Craig, J. W.; Drone, B.; Gerlach, R. H.; Williams, R. J.

    1991-01-01

    The developmental status is discussed regarding the 'lifeboat' vehicle to enhance the safety of the crew on the Space Station Freedom (SSF). NASA's Assured Crew Return Vehicle (ACRV) is intended to provide a means for returning the SSF crew to earth at all times. The 'lifeboat' philosophy is the key to managing the development of the ACRV which further depends on matrixed support and total quality management for implementation. The risk of SSF mission scenarios are related to selected ACRV mission requirements, and the system and vehicle designs are related to these precepts. Four possible ACRV configurations are mentioned including the lifting-body, Apollo shape, Discoverer shape, and a new lift-to-drag concept. The SCRAM design concept is discussed in detail with attention to the 'lifeboat' philosophy and requirements for implementation.

  4. Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Gorzel, A.

    2006-07-01

    Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a Germanmore » BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)« less

  5. QUICK RELEASABLE DRIVE

    DOEpatents

    Dickson, J.J.

    1958-07-01

    A quick releasable mechanical drive system suitable for use in a nuclear reactor is described. A small reversible motor positions a control rod by means of a worm and gear speed reducer, a magnetic torque clutch, and a bell crank. As the control rod is raised to the operating position, a heavy coil spring is compressed. In the event of an emergency indicated by either a''scram'' signal or a power failure, the current to the magnetic clutch is cut off, thereby freeing the coil spring and the bell crank positioner from the motor and speed reduction gearing. The coil spring will immediately act upon the bell crank to cause the insertion of the control rod. This arrangement will allow the slow, accurate positioning of the control rod during reactor operation, while providing an independent force to rapidly insert the rod in the event of an emergency.

  6. A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wheeler, Timothy A.; Liao, Huafei

    2014-12-01

    United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constitutedmore » a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.« less

  7. HWCTR CONTROL ROD AND SAFETY ROD DRIVE SYSTEMS

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kale, S.H.

    1963-07-01

    The Heavy Water Components Test Reactor (HWCTR) is a pressurized, D/sub 2/O reactor designed for operation up to 70 Mw at 1500 psig and 3l5 deg C. It has 18 control rods and six safety rods, each driven by an electric motor through a rack and pinion gear train. Racks, pinions, and bearings are located inside individual pressure housings that are penetrated by means of floating ring labyrinth seals. The drives are mounted on the reactor vessel top head. Safety rods have electromagnetic clutches and fall into the reactor when scrammed. The reliability and performance of the rod drives aremore » very good. Seal leakage is well within design limits. Recent inspections of seals and control rod plants showed no evidence of crud buildup or stress corrosion cracking of type 17- 4PH'' stainless steel components. (auth)« less

  8. DOE Office of Scientific and Technical Information (OSTI.GOV)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

    The PIUS Advanced Reactor is a 640-MW(e) pressurized-water reactor developed by Asea Brown Boveri. A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity normally is controlled by the boron concentration in the coolant and the temperature of the moderator coolant. Analyses of five initiating events have been completed on the basis of calculations performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. The initiating events analyzed are (1) reactor scram, (2) loss of off-site power (3) main steam-line break, (4) small-break loss of coolant, and (5) large-break loss of coolant. Inmore » addition to the baseline calculation for each sequence, sensitivity studies were performed to explore the response of the PIUS reactor to severe off-normal conditions having a very low probability of occurrence. The sensitivity studies provide insights into the robustness of the design.« less

  9. Computational Analysis of the Combustion Processes in an Axisymmetric, RBCC Flowpath

    NASA Technical Reports Server (NTRS)

    Steffen, Christopher J., Jr.; Yungster, Shaye

    2001-01-01

    Computational fluid dynamic simulations have been used to study the combustion processes within an axisymmetric, RBCC flowpath. Two distinct operating modes have been analyzed to date, including the independent ramjet stream (IRS) cycle and the supersonic combustion ramjet (scramJet) cycle. The IRS cycle investigation examined the influence of fuel-air ratio, fuel distribution, and rocket chamber pressure upon the combustion physics and thermal choke characteristics. Results indicate that adjustment of the amount and radial distribution of fuel can control the thermal choke point. The secondary massflow rate was very sensitive to the fuel-air ratio and the rocket chamber pressure. The scramjet investigation examined the influence of fuel-air ratio and fuel injection schedule upon combustion performance estimates. An analysis of the mesh-dependence of these calculations was presented. Jet penetration data was extracted from the three-dimensional simulations and compared favorably with experimental correlations of similar flows. Results indicate that combustion efficiency was very sensitive to the fuel schedule.

  10. DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hagen, E.W.

    This report reviews and evaluates the performance of the compressed-air and pressurized-nitrogen gas systems in commercial nuclear power units. The information was collected from readily available operating experiences, licensee event reports, system designs in safety analysis reports, and regulatory documents. The results are collated and analyzed for significance and impact on power plant safety performance. Under certain circumstances, the fail-safe philosophy for a piece of equipment or subsystem of the compressed-air systems initiated a series of actions culminating in reactor transient or unit scram. However, based on this study of prevailing operating experiences, reclassifying the compressed-gas systems to a highermore » safety level will neither prevent (nor mitigate) the reoccurrences of such happenings nor alleviate nuclear power plant problems caused by inadequate maintenance, operating procedures, and/or practices. Conversely, because most of the problems were derived from the sources listed previously, upgrading of both maintenance and operating procedures will not only result in substantial improvement in the performance and availability of the compressed-air (and backup nitrogen) systems but in improved overall plant performance.« less

  11. Recent improvements of reactor physics codes in MHI

    NASA Astrophysics Data System (ADS)

    Kosaka, Shinya; Yamaji, Kazuya; Kirimura, Kazuki; Kamiyama, Yohei; Matsumoto, Hideki

    2015-12-01

    This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO's Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.

  12. Inlet and Propulsion Integration of Scram Propelled Vehicles

    NASA Technical Reports Server (NTRS)

    Povinelli, Louis A.

    1996-01-01

    The material to be presented in these two lectures begins with cycle considerations of the turbojet engine combined with a ramjet engine to provide thrust over the range of Mach 0 to 5. We will then examine in some detail the aerodynamic behavior that occurs in the inlet operating near the peak speed. Following that, we shall view a numerical simulation through a baseline scramjet engine, starting at the entrance to the inlet, proceeding into the combustor and through the nozzle. In the next segment, we examine a combined rocket and ramjet propulsion system. Analysis and test results will be examined with a view toward evaluation of the concept as a practical device. Two other inlets will then be reviewed: a Mach 12 inlet and a Mach 18 configuration. Finally, we close our lectures with a discussion of the Detonation Wave engine, and inspect the physical and chemical behavior obtained from numerical simulation. A few final remarks will be made regarding the application of CFD for hypersonic propulsion components.

  13. 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.

  14. Final report, PT IP-535-C: Test of smaller VSR`s in DR reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Vaughn, A.D.

    1963-04-17

    Because of rod-sticking problems at DR Reactor, a knuckle rod of B Reactor design was installed in vertical safety channel number 28. The substitute VSR, which has a smaller diameter than the original DR rod, has been tested for its operational feasibility including both drop time and reactivity effect. The reactivity effect of the rod was estimated by comparison of the reactivity transient caused by insertion of the specific B-type rod after scramming into the pile, with similar transients caused by normal vertical safety rod in an adjacent channel. This document lists the indicated relative control strength of the rodmore » as an empirical basis for future safety calculations. Results indicate that the B-type knuckel rod in DR reactor is about 80% as strong as a normal DR vertical safety rod if used in equivalent flux distribution and location; this makes it reasonable to assume that the local control strength of the B-type knuckel rod is 98 {mu}b.« less

  15. Electron Temperature and Plasma Flow Measurements of NIF Hohlraum Plasmas

    NASA Astrophysics Data System (ADS)

    Barrios, M. A.; Liedahl, D. A.; Schneider, M. B.; Jones, O.; Brow, G. V.; Regan, S. P.; Fournier, K. B.; Moore, A. S.; Ross, J. S.; Eder, D.; Landen, O.; Kauffman, R. L.; Nikroo, A.; Kroll, J.; Jaquez, J.; Huang, H.; Hansen, S. B.; Callahan, D. A.; Hinkel, D. E.; Bradley, D.; Moody, J. D.; LLNL Collaboration; LLE Collaboration; GA Collaboration; SNL Collaboration

    2016-10-01

    Characterizing the plasma conditions inside NIF hohlraums, in particular mapping the plasma Te, is critical to gaining insight into mechanisms that affect energy coupling and transport in the hohlraum. The dot spectroscopy platform provides a temporal history of the localized Te and plasma flow inside a NIF hohlraum, by introducing a Mn-Co tracer dot, at strategic locations inside the hohlraum, that comes to equilibrium with the local plasma. K-shell X-ray spectroscopy of the tracer dot is recorded onto an absolutely calibrated X-ray streak spectrometer. Isoelectronic and interstage line ratios are used to infer localized Te through comparison with atomic physics calculations using SCRAM. Time resolved X-ray images are simultaneously taken of the expanding dot, providing plasma (ion) flow information. We present recent results provided by this platform and compare with simulations using HYDRA. This work was performed under the auspices of the U.S. Department of Energy by LLNL under Contract DE-AC52-07NA27344.

  16. RELAP5 Application to Accident Analysis of the NIST Research Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.

    Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accidentmore » and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.« less

  17. Trace Assessment for BWR ATWS Analysis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cheng, L.Y.; Diamond, D.; Arantxa Cuadra, Gilad Raitses, Arnold Aronson

    2010-04-22

    A TRACE/PARCS input model has been developed in order to be able to analyze anticipated transients without scram (ATWS) in a boiling water reactor. The model is based on one developed previously for the Browns Ferry reactor for doing loss-of-coolant accident analysis. This model was updated by adding the control systems needed for ATWS and a core model using PARCS. The control systems were based on models previously developed for the TRAC-B code. The PARCS model is based on information (e.g., exposure and moderator density (void) history distributions) obtained from General Electric Hitachi and cross sections for GE14 fuel obtainedmore » from an independent source. The model is able to calculate an ATWS, initiated by the closure of main steam isolation valves, with recirculation pump trip, water level control, injection of borated water from the standby liquid control system and actuation of the automatic depres-surization system. The model is not considered complete and recommendations are made on how it should be improved.« less

  18. Loss of DHR sequences at Browns Ferry Unit One - accident-sequence analysis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cook, D.H.; Grene, S.R.; Harrington, R.M.

    1983-05-01

    This study describes the predicted response of Unit One at the Browns Ferry Nuclear Plant to a postulated loss of decay heat removal (DHR) capability following scram from full power with the power conversion system unavailable. In accident sequences without DHR capability, the residual heat removal (RHR) system functions of pressure suppression pool cooling and reactor vessel shutdown cooling are unavailable. Consequently, all decay heat energy is stored in the pressure suppression pool with a concomitant increase in pool temperature and primary containment pressure. With the assumption that DHR capability is not regained during the lengthy course of this accidentmore » sequence, the containment ultimately fails by overpressurization. Although unlikely, this catastrophic failure might lead to loss of the ability to inject cooling water into the reactor vessel, causing subsequent core uncovery and meltdown. The timing of these events and the effective mitigating actions that might be taken by the operator are discussed in this report.« less

  19. RAMONA-3B application to Browns Ferry ATWS

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Slovik, G.C.; Neymotin, L.Y.; Saha, P.

    1985-01-01

    The Anticipated Transient Without Scram (ATWS) is known to be a dominant accident sequence for possible core melt in a Boiling Water Reactor (BWR). A recent Probabilistic Risk Assessment (PRA) analysis for the Browns Ferry nuclear power plant indicates that ATWS is the second most dominant transient for core melt in BWR/4 with Mark I containment. The most dominant sequence being the failure of long term decay heat removal function of the Residual Heat Removal (RHR) system. Of all the various ATWS scenarios, the Main Steam Isolation Valve (MSIV) closure ATWS sequence was chosen for present analysis because of itsmore » relatively high frequency of occurrence and its challenge to the residual heat removal system and containment integrity. The objective of this paper is to discuss four MSIV closure ATWS calculations using the RAMONA-3B code. The paper is a summary of a report being prepared for the USNRC Severe Accident Sequence Analysis (SASA) program which should be referred to for details. 10 refs., 20 figs., 3 tabs.« less

  20. Recent improvements of reactor physics codes in MHI

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kosaka, Shinya, E-mail: shinya-kosaka@mhi.co.jp; Yamaji, Kazuya; Kirimura, Kazuki

    2015-12-31

    This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO’s Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipatedmore » transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.« less

  1. Assured crew return vehicle post landing configuration design and test

    NASA Technical Reports Server (NTRS)

    1992-01-01

    The 1991-1992 senior Mechanical and Aerospace Engineering Design class continued work on the post landing configurations for the Assured Crew Return Vehicle (ACRV) and the Emergency Egress Couch (EEC). The ACRV will be permanently docked to Space Station Freedom fulfilling NASA's commitment of Assured Crew Return Capability in the event of an accident or illness aboard Space Station Freedom. The EEC provides medical support and a transportation surface for an incapacitated crew member. The objective of the projects was to give the ACRV Project Office data to feed into their feasibility studies. Four design teams were given the task of developing models with dynamically and geometrically scaled characteristics. Groups one and two combined efforts to design a one-fifth scale model for the Apollo Command Module derivative, an on-board flotation system, and a lift attachment point system. This model was designed to test the feasibility of a rigid flotation and stabilization system and to determine the dynamics associated with lifting the vehicle during retrieval. However, due to priorities, it was not built. Group three designed a one-fifth scale model of the Johnson Space Center (JSC) benchmark configuration, the Station Crew Return Alternative Module (SCRAM) with a lift attachment point system. This model helped to determine the flotation and lifting characteristics of the SCRAM configuration. Group four designed a full scale EEC with changeable geometric and geometric and dynamic characteristics. This model provided data on the geometric characteristics of the EEC and on the placement of the CG and moment of inertia. It also gave the helicopter rescue personnel direct input to the feasibility study. Section 1 describes in detail the design of a one-fifth scale model of the Apollo Command Module Derivative (ACMD) ACRV. The objective of the ACMD Configuration Model Team was to use geometric and dynamic constraints to design a one-fifth scale working model of the Apollo Command Module Derivative (ACMD) configuration with a Lift Attachment Point (LAP) System. This model was required to incorporate a rigidly mounted flotation system and the egress system designed the previous academic year. The LAP system was to be used to determine the dynamic effects of locating the lifting points at different locations on the vehicle. The team was then to build and test the model; however, due to priorities, this did not occur. To better simulate the ACMD after a water landing, the nose cone section was removed and the deck area exposed. The areas researched during the design process were construction, center of gravity and moment of inertia, and lift attachment points.

  2. Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    De Rosa, Felice

    2006-07-01

    In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to themore » accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when {delta}Tsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its specific models (candling, corium pool behaviour, etc.) they were less good. A future work will be the preparation of an input deck for the new MELCOR 1.8.6. and to perform a code-to-code comparison with ASTEC v1.2 rev. 1. (author)« less

  3. Improvements in Modeling Au Sphere Non-LTE X-ray Emission

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Rosen, M D; Scott, H A; Suter, L J

    2008-10-30

    We've previously reported on experiments at the Omega laser at URLLE, in which 1.0 mm in diameter, Au coated, spheres, were illuminated at either 10{sup 14} W/cm{sup 2} (10 kJ/3 ns) or at 10{sup 15} W/cm{sup 2} (30 kJ/1 ns). Spectral information on the 1 keV thermal x-rays, as well as the multi-keV M-band were obtained. We compared a variety of non-LTE atomic physics packages to this data with varying degrees of success. In this paper we broaden the scope of the investigation, and compare the data to newer models: (1) An improved Detailed Configuration Accounting (DCA) method; and (2)more » This model involves adjustments to the standard XSN non-LTE model which lead to a better match of coronal emission as calculated by XSN to that calculated by SCRAM, a more sophisticated stand-alone model. We show some improvements in the agreement with Omega data when using either of these new approaches.« less

  4. Impact of aeroelasticity on propulsion and longitudinal flight dynamics of an air-breathing hypersonic vehicle

    NASA Technical Reports Server (NTRS)

    Raney, David L.; Mcminn, John D.; Pototzky, Anthony S.; Wooley, Christine L.

    1993-01-01

    Many air-breathing hypersonic aerospacecraft design concepts incorporate an elongated fuselage forebody acting as the aerodynamic compression surface for a hypersonic combustion module, or scram jet. This highly integrated design approach creates the potential for an unprecedented form of aero-propulsive-elastic interaction in which deflections of the vehicle fuselage give rise to propulsion transients, producing force and moment variations that may adversely impact the rigid body flight dynamics and/or further excite the fuselage bending modes. To investigate the potential for such interactions, a math model was developed which included the longitudinal flight dynamics, propulsion system, and first seven elastic modes of a hypersonic air-breathing vehicle. Perturbation time histories from a simulation incorporating this math model are presented that quantify the propulsive force and moment variations resulting from aeroelastic vehicle deflections. Root locus plots are presented to illustrate the effect of feeding the propulsive perturbations back into the aeroelastic model. A concluding section summarizes the implications of the observed effects for highly integrated hypersonic air-breathing vehicle concepts.

  5. Impact of aeroelasticity on propulsion and longitudinal flight dynamics of an air-breathing hypersonic vehicle

    NASA Astrophysics Data System (ADS)

    Raney, David L.; McMinn, John D.; Pototzky, Anthony S.; Wooley, Christine L.

    1993-04-01

    Many air-breathing hypersonic aerospacecraft design concepts incorporate an elongated fuselage forebody acting as the aerodynamic compression surface for a hypersonic combustion module, or scram jet. This highly integrated design approach creates the potential for an unprecedented form of aero-propulsive-elastic interaction in which deflections of the vehicle fuselage give rise to propulsion transients, producing force and moment variations that may adversely impact the rigid body flight dynamics and/or further excite the fuselage bending modes. To investigate the potential for such interactions, a math model was developed which included the longitudinal flight dynamics, propulsion system, and first seven elastic modes of a hypersonic air-breathing vehicle. Perturbation time histories from a simulation incorporating this math model are presented that quantify the propulsive force and moment variations resulting from aeroelastic vehicle deflections. Root locus plots are presented to illustrate the effect of feeding the propulsive perturbations back into the aeroelastic model. A concluding section summarizes the implications of the observed effects for highly integrated hypersonic air-breathing vehicle concepts.

  6. Reliability enhancement of APR + diverse protection system regarding common cause failures

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Oh, Y. G.; Kim, Y. M.; Yim, H. S.

    2012-07-01

    The Advanced Power Reactor Plus (APR +) nuclear power plant design has been developed on the basis of the APR1400 (Advanced Power Reactor 1400 MWe) to further enhance safety and economics. For the mitigation of Anticipated Transients Without Scram (ATWS) as well as Common Cause Failures (CCF) within the Plant Protection System (PPS) and the Emergency Safety Feature - Component Control System (ESF-CCS), several design improvement features have been implemented for the Diverse Protection System (DPS) of the APR + plant. As compared to the APR1400 DPS design, the APR + DPS has been designed to provide the Safety Injectionmore » Actuation Signal (SIAS) considering a large break LOCA accident concurrent with the CCF. Additionally several design improvement features, such as channel structure with redundant processing modules, and changes of system communication methods and auto-system test methods, are introduced to enhance the functional reliability of the DPS. Therefore, it is expected that the APR + DPS can provide an enhanced safety and reliability regarding possible CCF in the safety-grade I and C systems as well as the DPS itself. (authors)« less

  7. Self-actuated shutdown system for a commercial size LMFBR. Final report

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dupen, C.F.G.

    1978-08-01

    A Self-Actuated Shutdown System (SASS) is defined as a reactor shutdown system in which sensors, release mechanisms and neutron absorbers are contained entirely within the reactor core structure, where they respond inherently to abnormal local process conditions, by shutting down the reactor, independently of the plant protection system (PPS). It is argued that a SASS, having a response time similar to that of the PPS, would so reduce the already very low probability of a failure-to-scram event that costly design features, derived from core disruptive accident analysis, could be eliminated. However, the thrust of the report is the feasibility andmore » reliability of the in-core SASS hardware to achieve sufficiently rapid shutdown. A number of transient overpower and transient undercooling-responsive systems were investigated leading to the selection of a primary candidate and a backup concept. During a transient undercooling event, the recommended device is triggered by the associated rate of change of pressure, whereas the alternate concept responds to the reduction in core pressure drop and requires calibration and adjustment by the operators to accommodate changes in reactor power.« less

  8. Electron temperature measurements inside the ablating plasma of gas-filled hohlraums at the National Ignition Facility

    NASA Astrophysics Data System (ADS)

    Barrios, M. A.; Liedahl, D. A.; Schneider, M. B.; Jones, O.; Brown, G. V.; Regan, S. P.; Fournier, K. B.; Moore, A. S.; Ross, J. S.; Landen, O.; Kauffman, R. L.; Nikroo, A.; Kroll, J.; Jaquez, J.; Huang, H.; Hansen, S. B.; Callahan, D. A.; Hinkel, D. E.; Bradley, D.; Moody, J. D.

    2016-05-01

    The first measurement of the electron temperature (Te) inside a National Ignition Facility hohlraum is obtained using temporally resolved K-shell X-ray spectroscopy of a mid-Z tracer dot. Both isoelectronic- and interstage-line ratios are used to calculate the local Te via the collisional-radiative atomic physics code SCRAM [Hansen et al., High Energy Density Phys 3, 109 (2007)]. The trajectory of the mid-Z dot as it is ablated from the capsule surface and moves toward the laser entrance hole (LEH) is measured using side-on x-ray imaging, characterizing the plasma flow of the ablating capsule. Data show that the measured dot location is farther away from the LEH in comparison to the radiation-hydrodynamics simulation prediction using HYDRA [Marinak et al., Phys. Plasmas 3, 2070 (1996)]. To account for this discrepancy, the predicted simulation Te is evaluated at the measured dot trajectory. The peak Te, measured to be 4.2 keV ± 0.2 keV, is ˜0.5 keV hotter than the simulation prediction.

  9. A Coding Variant of ANO10, Affecting Volume Regulation of Macrophages, Is Associated with Borrelia Seropositivity

    PubMed Central

    Hammer, Christian; Wanitchakool, Podchanart; Sirianant, Lalida; Papiol, Sergi; Monnheimer, Mathieu; Faria, Diana; Ousingsawat, Jiraporn; Schramek, Natalie; Schmitt, Corinna; Margos, Gabriele; Michel, Angelika; Kraiczy, Peter; Pawlita, Michael; Schreiber, Rainer; Schulz, Thomas F; Fingerle, Volker; Tumani, Hayrettin; Ehrenreich, Hannelore; Kunzelmann, Karl

    2015-01-01

    In a first genome-wide association study (GWAS) approach to anti-Borrelia seropositivity, we identified two significant single nucleotide polymorphisms (SNPs) (rs17850869, P = 4.17E-09; rs41289586, P = 7.18E-08). Both markers, located on chromosomes 16 and 3, respectively, are within or close to genes previously connected to spinocerebellar ataxia. The risk SNP rs41289586 represents a missense variant (R263H) of anoctamin 10 (ANO10), a member of a protein family encoding Cl− channels and phospholipid scram-blases. ANO10 augments volume-regulated Cl− currents (IHypo) in Xenopus oocytes, HEK293 cells, lymphocytes and macrophages and controls volume regulation by enhancing regulatory volume decrease (RVD). ANO10 supports migration of macrophages and phagocytosis of spirochetes. The R263H variant is inhibitory on IHypo, RVD and intracellular Ca2+ signals, which may delay spirochete clearance, thereby sensitizing adaptive immunity. Our data demonstrate for the first time that ANO10 has a central role in innate immune defense against Borrelia infection. PMID:25730773

  10. Characterization of laser-cut copper foil X-pinches

    NASA Astrophysics Data System (ADS)

    Collins, G. W.; Valenzuela, J. C.; Hansen, S. B.; Wei, M. S.; Reed, C. T.; Forsman, A. C.; Beg, F. N.

    2016-10-01

    Quantitative data analyses of laser-cut Cu foil X-pinch experiments on the 150 ns quarter-period, ˜250 kA GenASIS driver are presented. Three different foil designs are tested to determine the effects of initial structure on pinch outcome. Foil X-pinch data are also presented alongside the results from wire X-pinches with comparable mass. The X-ray flux and temporal profile of the emission from foil X-pinches differed significantly from that of wire X-pinches, with all emission from the foil X-pinches confined to a ˜3 ns period as opposed to the delayed, long-lasting electron beam emission common in wire X-pinches. Spectroscopic data show K-shell as well as significant L-shell emission from both foil and wire X-pinches. Fits to synthetic spectra using the SCRAM code suggest that pinching foil X's produced a ˜1 keV, ne ≥ 1023 cm-3 plasma. The spectral data combined with the improved reliability of the source timing, flux, and location indicate that foil X-pinches generate a reproducible, K-shell point-projection radiography source that can be easily modified and tailored to suit backlighting needs across a variety of applications.

  11. Posttest data analysis of FIST experimental TRAC-BD1/MOD1 power transient experiment

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wheatley, P.D.; Wagner, K.C.

    The FIST power transient test 6PMC2 was analyzed to further the understanding of the FIST facility and provide an assessment of TRAC-BD1/MOD1. FIST power transient 6PMC2 investigated the thermal-hydraulic response following inadvertent closure of the main steam isolation valve and the subsequent failure of the reactor to scram. Failure of the high pressure core spray system was also assumed, resulting in only the reactor core isolation cooling flow for inventory makeup during the transient. The experiment was a sensitivity study with relatively high core power and low makeup rates. This study provides one of the first opportunities to assess TRAC-BD1/MOD1more » under power transient and natural circulation conditions with data from a facility with prototypical BWR geometry. The power transient test was analyzed with emphasis on the following phenomena: (a) the system pressure response, (b) the natural circulation flows and rates, and (c) the heater rod cladding temperature response. Based on the results of this study, TRAC-BD1/MOD1 can be expected to calculate the thermal-hydraulic behavior of a BWR during a power transient.« less

  12. TRACE/PARCS Analysis of ATWS with Instability for a MELLLA+BWR/5

    DOE PAGES

    L. Y. Cheng; Baek, J. S.; Cuadra, A.; ...

    2016-06-06

    A TRACE/PARCS model has been developed to analyze anticipated transient without SCRAM (ATWS) events for a boiling water reactor (BWR) operating in the maximum extended load line limit analysis-plus (MELLLA+) expanded operating domain. The MELLLA+ domain expands allowable operation in the power/flow map of a BWR to low flow rates at high power conditions. Such operation exacerbates the likelihood of large amplitude power/flow oscillations during certain ATWS scenarios. The analysis shows that large amplitude power/flow oscillations, both core-wide and out-of-phase, arise following the establishment of natural circulation flow in the reactor pressure vessel (RPV) after the trip of the recirculationmore » pumps and an increase in core inlet subcooling. The analysis also indicates a mechanism by which the fuel may experience heat-up that could result in localized fuel damage. TRACE predicts the heat-up to occur when the cladding surface temperature exceeds the minimum stable film boiling temperature after periodic cycles of dryout and rewet; and the fuel becomes “locked” into a film boiling regime. Further, the analysis demonstrates the effectiveness of the simulated manual operator actions to suppress the instability.« less

  13. DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kolaczkowski, A.M.; Lambright, J.A.; Ferrell, W.L.

    This document contains the internal event initiated accident sequence analyses for Peach Bottom, Unit 2; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Peach Bottom, Unit 2, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provided additional insights regarding the dominant contributors to the Peach Bottom core damage frequency estimate. The mean core damage frequency at Peach Bottom wasmore » calculated to be 8.2E-6. Station blackout type accidents (loss of all ac power) were found to dominate the overall results. Anticipated Transient Without Scram accidents were also found to be non-negligible contributors. The numerical results are largely driven by common mode failure probability estimates and to some extent, human error. Because of significant data and analysis uncertainties in these two areas (important, for instance, to the most dominant scenario in this study), it is recommended that the results of the uncertainty and sensitivity analyses be considered before any actions are taken based on this analysis.« less

  14. Operator Support System Design forthe Operation of RSG-GAS Research Reactor

    NASA Astrophysics Data System (ADS)

    Santoso, S.; Situmorang, J.; Bakhri, S.; Subekti, M.; Sunaryo, G. R.

    2018-02-01

    The components of RSG-GAS main control room are facing the problem of material ageing and technology obsolescence as well, and therefore the need for modernization and refurbishment are essential. The modernization in control room can be applied on the operator support system which bears the function in providing information for assisting the operator in conducting diagnosis and actions. The research purpose is to design an operator support system for RSG-GAS control room. The design was developed based on the operator requirement in conducting task operation scenarios and the reactor operation characteristics. These scenarios include power operation, low power operation and shutdown/scram reactor. The operator support system design is presented in a single computer display which contains structure and support system elements e.g. operation procedure, status of safety related components and operational requirements, operation limit condition of parameters, alarm information, and prognosis function. The prototype was developed using LabView software and consisted of components structure and features of the operator support system. Information of each component in the operator support system need to be completed before it can be applied and integrated in the RSG-GAS main control room.

  15. Spectral and Atomic Physics Analysis of Xenon L-Shell Emission From High Energy Laser Produced Plasmas

    NASA Astrophysics Data System (ADS)

    Thorn, Daniel; Kemp, G. E.; Widmann, K.; Benjamin, R. D.; May, M. J.; Colvin, J. D.; Barrios, M. A.; Fournier, K. B.; Liedahl, D.; Moore, A. S.; Blue, B. E.

    2016-10-01

    The spectrum of the L-shell (n =2) radiation in mid to high-Z ions is useful for probing plasma conditions in the multi-keV temperature range. Xenon in particular with its L-shell radiation centered around 4.5 keV is copiously produced from plasmas with electron temperatures in the 5-10 keV range. We report on a series of time-resolved L-shell Xe spectra measured with the NIF X-ray Spectrometer (NXS) in high-energy long-pulse (>10 ns) laser produced plasmas at the National Ignition Facility. The resolving power of the NXS is sufficiently high (E/ ∂E >100) in the 4-5 keV spectral band that the emission from different charge states is observed. An analysis of the time resolved L-shell spectrum of Xe is presented along with spectral modeling by detailed radiation transport and atomic physics from the SCRAM code and comparison with predictions from HYDRA a radiation-hydrodynamics code with inline atomic-physics from CRETIN. This work was performed under the auspices of the U.S. Department of Energy by LLNL under Contract DE-AC52-07NA27344.

  16. Electron temperature measurements inside the ablating plasma of gas-filled hohlraums at the National Ignition Facility

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Barrios, M. A.; Liedahl, D. A.; Schneider, M. B.

    The first measurement of the electron temperature (T{sub e}) inside a National Ignition Facility hohlraum is obtained using temporally resolved K-shell X-ray spectroscopy of a mid-Z tracer dot. Both isoelectronic- and interstage-line ratios are used to calculate the local T{sub e} via the collisional–radiative atomic physics code SCRAM [Hansen et al., High Energy Density Phys 3, 109 (2007)]. The trajectory of the mid-Z dot as it is ablated from the capsule surface and moves toward the laser entrance hole (LEH) is measured using side-on x-ray imaging, characterizing the plasma flow of the ablating capsule. Data show that the measured dotmore » location is farther away from the LEH in comparison to the radiation-hydrodynamics simulation prediction using HYDRA [Marinak et al., Phys. Plasmas 3, 2070 (1996)]. To account for this discrepancy, the predicted simulation T{sub e} is evaluated at the measured dot trajectory. The peak T{sub e}, measured to be 4.2 keV ± 0.2 keV, is ∼0.5 keV hotter than the simulation prediction.« less

  17. RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hongbin Zhang; Haihua Zhao; Cliff Davis

    2008-06-01

    An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heatmore » from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet temperature. The peak cladding, hot pool, cold pool and reactor inlet temperatures were calculated during LOFC. The results indicate that there are two phases during LOFC transient – the initial thermal equilibration phase and the long term decay heat removal phase. The initial thermal equilibration phase occurs over a few hundred seconds, as the system adjusts from forced circulation to natural circulation flow. Subsequently, during long-term heat removal phase all temperatures evolve very slowly due to the large thermal inertia of the primary and buffer pool systems. The results clearly show that passive safety PRACS can effectively transfer decay heat from the primary system to the buffer pool by natural circulation. The DRACS system in turn can effectively transfer the decay heat to the environment.« less

  18. Posttest data analysis and assessment of TRAC-BD1/MOD1 with data from a Full Integral Simulation Test (FIST) power transient experiment

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wheatley, P.D.; Wagner, K.C.

    The FIST power transient test 6PMC2 was analyzed to further the understanding of the FIST facility and provide an assessment of TRAC-BD1/MOD1. FIST power transient 6PMC2 investigated the thermal-hydraulic response following inadvertent closure of the main steam isolation valve and the subsequent failure of the reactor to scram. Failure of the high pressure core spray system was also assumed, resulting on only the reactor core isolation cooling flow for inventory makeup during the transient. The experiment was a sensitivity study with relatively high core power and low makeup rates. This study provides one of the first opportunities to assess TRAC-BD1/MOD1more » under power transient and natural circulation conditions with data from a facility with prototypical BWR geometry. The power transient test was analyzed with emphasis on the following phenomena; (a) the system pressure response, (b) the natural circulation flows and rates, and (c) the heater rod cladding temperature response. Based on the results of this study, TRAC-BD1/MOD1 can be expected to calculate the thermal-hydraulic behavior of a BWR during a power transient.« less

  19. Shock Position Control for Mode Transition in a Turbine Based Combined Cycle Engine Inlet Model

    NASA Technical Reports Server (NTRS)

    Csank, Jeffrey T.; Stueber, Thomas J.

    2013-01-01

    A dual flow-path inlet for a turbine based combined cycle (TBCC) propulsion system is to be tested in order to evaluate methodologies for performing a controlled inlet mode transition. Prior to experimental testing, simulation models are used to test, debug, and validate potential control algorithms which are designed to maintain shock position during inlet disturbances. One simulation package being used for testing is the High Mach Transient Engine Cycle Code simulation, known as HiTECC. This paper discusses the development of a mode transition schedule for the HiTECC simulation that is analogous to the development of inlet performance maps. Inlet performance maps, derived through experimental means, describe the performance and operability of the inlet as the splitter closes, switching power production from the turbine engine to the Dual Mode Scram Jet. With knowledge of the operability and performance tradeoffs, a closed loop system can be designed to optimize the performance of the inlet. This paper demonstrates the design of the closed loop control system and benefit with the implementation of a Proportional-Integral controller, an H-Infinity based controller, and a disturbance observer based controller; all of which avoid inlet unstart during a mode transition with a simulated disturbance that would lead to inlet unstart without closed loop control.

  20. Three Dimensional CFD Analysis of the GTX Combustor

    NASA Technical Reports Server (NTRS)

    Steffen, C. J., Jr.; Bond, R. B.; Edwards, J. R.

    2002-01-01

    The annular combustor geometry of a combined-cycle engine has been analyzed with three-dimensional computational fluid dynamics. Both subsonic combustion and supersonic combustion flowfields have been simulated. The subsonic combustion analysis was executed in conjunction with a direct-connect test rig. Two cold-flow and one hot-flow results are presented. The simulations compare favorably with the test data for the two cold flow calculations; the hot-flow data was not yet available. The hot-flow simulation indicates that the conventional ejector-ramjet cycle would not provide adequate mixing at the conditions tested. The supersonic combustion ramjet flowfield was simulated with frozen chemistry model. A five-parameter test matrix was specified, according to statistical design-of-experiments theory. Twenty-seven separate simulations were used to assemble surrogate models for combustor mixing efficiency and total pressure recovery. ScramJet injector design parameters (injector angle, location, and fuel split) as well as mission variables (total fuel massflow and freestream Mach number) were included in the analysis. A promising injector design has been identified that provides good mixing characteristics with low total pressure losses. The surrogate models can be used to develop performance maps of different injector designs. Several complex three-way variable interactions appear within the dataset that are not adequately resolved with the current statistical analysis.

  1. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bohachek, Randolph Charles

    2015-09-01

    The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactorsmore » is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.« less

  2. A numerical study of mixing in supersonic combustors with hypermixing injectors

    NASA Technical Reports Server (NTRS)

    Lee, J.

    1993-01-01

    A numerical study was conducted to evaluate the performance of wall mounted fuel-injectors designed for potential Supersonic Combustion Ramjet (SCRAM-jet) engine applications. The focus of this investigation was to numerically simulate existing combustor designs for the purpose of validating the numerical technique and the physical models developed. Three different injector designs of varying complexity were studied to fully understand the computational implications involved in accurate predictions. A dual transverse injection system and two streamwise injector designs were studied. The streamwise injectors were designed with swept ramps to enhance fuel-air mixing and combustion characteristics at supersonic speeds without the large flow blockage and drag contribution of the transverse injection system. For this study, the Mass-Average Navier-Stokes equations and the chemical species continuity equations were solved. The computations were performed using a finite-volume implicit numerical technique and multiple block structured grid system. The interfaces of the multiple block structured grid systems were numerically resolved using the flux-conservative technique. Detailed comparisons between the computations and existing experimental data are presented. These comparisons show that numerical predictions are in agreement with the experimental data. These comparisons also show that a number of turbulence model improvements are needed for accurate combustor flowfield predictions.

  3. A numerical study of mixing in supersonic combustors with hypermixing injectors

    NASA Technical Reports Server (NTRS)

    Lee, J.

    1992-01-01

    A numerical study was conducted to evaluate the performance of wall mounted fuel-injectors designed for potential Supersonic Combustion Ramjet (SCRAM-jet) engine applications. The focus of this investigation was to numerically simulate existing combustor designs for the purpose of validating the numerical technique and the physical models developed. Three different injector designs of varying complexity were studied to fully understand the computational implications involved in accurate predictions. A dual transverse injection system and two streamwise injector designs were studied. The streamwise injectors were designed with swept ramps to enhance fuel-air mixing and combustion characteristics at supersonic speeds without the large flow blockage and drag contribution of the transverse injection system. For this study, the Mass-Averaged Navier-Stokes equations and the chemical species continuity equations were solved. The computations were performed using a finite-volume implicit numerical technique and multiple block structured grid system. The interfaces of the multiple block structured grid systems were numerically resolved using the flux-conservative technique. Detailed comparisons between the computations and existing experimental data are presented. These comparisons show that numerical predictions are in agreement with the experimental data. These comparisons also show that a number of turbulence model improvements are needed for accurate combustor flowfield predictions.

  4. Technical activities report: Heat, water, and mechanical studies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Alexander, W.K.

    1951-10-04

    Topics in the heat studies section include: front and rear face reflector shields at the C-pile; process tube channel thermocouples; water temperature limits for horizontal rods; slug temperature and thermal conductivity calculations; maximum slug-end cap temperature; boiling consideration studies; scram time limit for Panellit alarm; heat transfer test; slug stresses; thermal insulation of bottom tube row at C-pile; flow tests; present pile enrichment; electric analog; and measurement of thermal contact resistance. Topics in the water studies section include: 100-D flow laboratory; process water studies; fundamental studies on film formation; coatings on tip-offs; can difference tests; slug jacket abrasion at highmore » flow rates; corrosion studies; front tube dummy slugs; metallographic examination of tubes from H-pile; fifty-tube mock-up; induction heating facility; operational procedures and standards; vertical safety rod dropping time tests; recirculation; and power recovery. Mechanical development studies include: effect of Sphincter seal and lubricant VSR drop time; slug damage; slug bubble tester; P-13 removal; chemical slug stripper; effect of process tube rib spacing and width; ink facility installation; charging and discharging machines; process tube creep; flapper nozzle assembly test; test of single gun barrel assembly; pigtail fixture test; horizontal rod gland seal test; function test of C-pile; and intermediate test of Ball 3-X and VSR systems.« less

  5. Quantitative determination of caffeine and alcohol in energy drinks and the potential to produce positive transdermal alcohol concentrations in human subjects.

    PubMed

    Ayala, Jessica; Simons, Kelsie; Kerrigan, Sarah

    2009-01-01

    The purpose of this study was to determine whether non-alcoholic energy drinks could result in positive "alcohol alerts" based on transdermal alcohol concentration (TAC) using a commercially available electrochemical monitoring device. Eleven energy drinks were quantitatively assayed for both ethanol and caffeine. Ethanol concentrations for all of the non-alcoholic energy drinks ranged in concentration from 0.03 to 0.230% (w/v) and caffeine content per 8-oz serving ranged from 65 to 126 mg. A total of 15 human subjects participated in the study. Subjects consumed between 6 and 8 energy drinks over an 8-h period. The SCRAM II monitoring device was used to determine TACs every 30 min before, during, and after the study. None of the subjects produced TAC readings that resulted in positive "alcohol alerts". TAC measurements for all subjects before, during and after the energy drink study period (16 h total) were <0.02% (w/v). Subjects in the study consumed a quantity of non-alcoholic energy drink that greatly exceeds what would be considered typical. Based on these results, it appears that energy drink consumption is an unlikely explanation for elevated TACs that might be identified as potential drinking episodes or "alcohol alerts" using this device.

  6. NOAA Atmospheric Sciences Modeling Division support to the US Environmental Protection Agency

    NASA Astrophysics Data System (ADS)

    Poole-Kober, Evelyn M.; Viebrock, Herbert J.

    1991-07-01

    During FY-1990, the Atmospheric Sciences Modeling Division provided meteorological research and operational support to the U.S. Environmental Protection Agency. Basic meteorological operational support consisted of applying dispersion models and conducting dispersion studies and model evaluations. The primary research effort was the development and evaluation of air quality simulation models using numerical and physical techniques supported by field studies. Modeling emphasis was on the dispersion of photochemical oxidants and particulate matter on urban and regional scales, dispersion in complex terrain, and the transport, transformation, and deposition of acidic materials. Highlights included expansion of the Regional Acid Deposition Model/Engineering Model family to consist of the Tagged Species Engineering Model, the Non-Depleting Model, and the Sulfate Tracking Model; completion of the Acid-MODES field study; completion of the RADM2.1 evaluation; completion of the atmospheric processes section of the National Acid Precipitation Assessment Program 1990 Integrated Assessment; conduct of the first field study to examine the transport and entrainment processes of convective clouds; development of a Regional Oxidant Model-Urban Airshed Model interface program; conduct of an international sodar intercomparison experiment; incorporation of building wake dispersion in numerical models; conduct of wind-tunnel simulations of stack-tip downwash; and initiation of the publication of SCRAM NEWS.

  7. Numerical simulations of microcrack-related damage and ignition behavior of mild-impacted polymer bonded explosives.

    PubMed

    Yang, Kun; Wu, Yanqing; Huang, Fenglei

    2018-08-15

    A physical model is developed to describe the viscoelastic-plastic deformation, cracking damage, and ignition behavior of polymer-bonded explosives (PBXs) under mild impact. This model improves on the viscoelastic-statistical crack mechanical model (Visco-SCRAM) in several respects. (i) The proposed model introduces rate-dependent plasticity into the framework which is more suitable for explosives with relatively high binder content. (ii) Damage evolution is calculated by the generalized Griffith instability criterion with the dominant (most unstable) crack size rather than the averaged crack size over all crack orientations. (iii) The fast burning of cracks following ignition and the effects of gaseous products on crack opening are considered. The predicted uniaxial and triaxial stress-strain responses of PBX9501 sample under dynamic compression loading are presented to illustrate the main features of the materials. For an uncovered cylindrical PBX charge impacted by a flat-nosed rod, the simulated results show that a triangular-shaped dead zone is formed beneath the front of the rod. The cracks in the dead zone are stable due to friction-locked stress state, whereas the cracks near the front edges of dead zone become unstable and turn into hotspots due to high-shear effects. Copyright © 2018 Elsevier B.V. All rights reserved.

  8. HIFIRE Flight 2 Overview and Status Update 2011

    NASA Technical Reports Server (NTRS)

    Jackson, Kevin R.; Gruber, Mark R.; Buccellato, Salvatore

    2011-01-01

    A collaborative international effort, the Hypersonic International Flight Research Experimentation (HIFiRE) Program aims to study basic hypersonic phenomena through flight experimentation. HIFiRE Flight 2 teams the United States Air Force Research Lab (AFRL), NASA, and the Australian Defence Science and Technology Organisation (DSTO). Flight 2 will develop an alternative test technique for acquiring high enthalpy scramjet flight test data, allowing exploration of accelerating hydrocarbon-fueled scramjet performance and dual-to-scram mode transition up to and beyond Mach 8 flight. The generic scramjet flowpath is research quality and the test fuel is a simple surrogate for an endothermically cracked liquid hydrocarbon fuel. HIFiRE Flight 2 will be a first of its kind in contribution to scramjets. The HIFiRE program builds upon the HyShot and HYCAUSE programs and aims to leverage the low-cost flight test technique developed in those programs. It will explore suppressed trajectories of a sounding rocket propelled test article and their utility in studying ramjet-scramjet mode transition and flame extinction limits research. This paper describes the overall scramjet flight test experiment mission goals and objectives, flight test approach and strategy, ground test and analysis summary, development status and project schedule. A successful launch and operation will present to the scramjet community valuable flight test data in addition to a new tool, and vehicle, with which to explore high enthalpy scramjet technologies.

  9. New fixed-point mini-cell to investigate thermocouple drift in a high-temperature environment under neutron irradiation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Laurie, M.; Vlahovic, L.; Rondinella, V.V.

    Temperature measurements in the nuclear field require a high degree of reliability and accuracy. Despite their sheathed form, thermocouples subjected to nuclear radiations undergo changes due to radiation damage and transmutation that lead to significant EMF drift during long-term fuel irradiation experiment. For the purpose of a High Temperature Reactor fuel irradiation to take place in the High Flux Reactor Petten, a dedicated fixed-point cell was jointly developed by LNE-Cnam and JRC-IET. The developed cell to be housed in the irradiation rig was tailor made to quantify the thermocouple drift during the irradiation (about two year duration) and withstand highmore » temperature (in the range 950 deg. C - 1100 deg. C) in the presence of contaminated helium in a graphite environment. Considering the different levels of temperature achieved in the irradiation facility and the large palette of thermocouple types aimed at surveying the HTR fuel pebble during the qualification test both copper (1084.62 deg. C) and gold (1064.18 deg. C) fixed-point materials were considered. The aim of this paper is to first describe the fixed-point mini-cell designed to be embedded in the reactor rig and to discuss the preliminary results achieved during some out of pile tests as much as some robustness tests representative of the reactor scram scenarios. (authors)« less

  10. Independent Qualification of the CIAU Tool Based on the Uncertainty Estimate in the Prediction of Angra 1 NPP Inadvertent Load Rejection Transient

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Borges, Ronaldo C.; D'Auria, Francesco; Alvim, Antonio Carlos M.

    2002-07-01

    The Code with - the capability of - Internal Assessment of Uncertainty (CIAU) is a tool proposed by the 'Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione (DIMNP)' of the University of Pisa. Other Institutions including the nuclear regulatory body from Brazil, 'Comissao Nacional de Energia Nuclear', contributed to the development of the tool. The CIAU aims at providing the currently available Relap5/Mod3.2 system code with the integrated capability of performing not only relevant transient calculations but also the related estimates of uncertainty bands. The Uncertainty Methodology based on Accuracy Extrapolation (UMAE) is used to characterize the uncertainty in themore » prediction of system code calculations for light water reactors and is internally coupled with the above system code. Following an overview of the CIAU development, the present paper deals with the independent qualification of the tool. The qualification test is performed by estimating the uncertainty bands that should envelope the prediction of the Angra 1 NPP transient RES-11. 99 originated by an inadvertent complete load rejection that caused the reactor scram when the unit was operating at 99% of nominal power. The current limitation of the 'error' database, implemented into the CIAU prevented a final demonstration of the qualification. However, all the steps for the qualification process are demonstrated. (authors)« less

  11. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    NASA Astrophysics Data System (ADS)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  12. DOE Office of Scientific and Technical Information (OSTI.GOV)

    Harmony, S.C.; Steiner, J.L.; Stumpf, H.J.

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. As part of the preapplication and eventual design certification process, advanced reactor applicants are required to submit neutronic and thermal-hydraulic safety analyses over a sufficient range of normal operation, transient conditions, and specified accident sequences. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. A fully one-dimensional modelmore » of the PIUS reactor has been developed for the Transient Reactor Analysis Code, TRACPF1/MOD2. Early in 1992, ABB submitted a Supplemental Information Package describing recent design modifications. An important feature of the PIUS Supplement design was the addition of an active scram system that will function for most transient and accident conditions. A one-dimensional Transient Reactor Analysis Code baseline calculation of the PIUS Supplement design were performed for a break in the main steam line at the outlet nozzle of the loop 3 steam generator. Sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions following a main steam line break. The sensitivity study results provide insights into the robustness of the design.« less

  13. Use of continuous transdermal alcohol monitoring during a contingency management procedure to reduce excessive alcohol use.

    PubMed

    Dougherty, Donald M; Hill-Kapturczak, Nathalie; Liang, Yuanyuan; Karns, Tara E; Cates, Sharon E; Lake, Sarah L; Mullen, Jillian; Roache, John D

    2014-09-01

    Research on contingency management to treat excessive alcohol use is limited due to feasibility issues with monitoring adherence. This study examined the effectiveness of using transdermal alcohol monitoring as a continuous measure of alcohol use to implement financial contingencies to reduce heavy drinking. Twenty-six male and female drinkers (from 21 to 39 years old) were recruited from the community. Participants were randomly assigned to one of the two treatment sequences. Sequence 1 received 4 weeks of no financial contingency (i.e., $0) drinking followed by 4 weeks each of $25 and then $50 contingency management; Sequence 2 received 4 weeks of $25 contingency management followed by 4 weeks each of no contingency (i.e., $0) and then $50 contingency management. During the $25 and $50 contingency management conditions, participants were paid each week when the Secure Continuous Remote Alcohol Monitor (SCRAM-II™) identified no heavy drinking days. Participants in both contingency management conditions had fewer drinking episodes and reduced frequencies of heavy drinking compared to the $0 condition. Participants randomized to Sequence 2 (receiving $25 contingency before the $0 condition) exhibited less frequent drinking and less heavy drinking in the $0 condition compared to participants from Sequence 1. Transdermal alcohol monitoring can be used to implement contingency management programs to reduce excessive alcohol consumption. Copyright © 2014 Elsevier Ireland Ltd. All rights reserved.

  14. Use of Continuous Transdermal Alcohol Monitoring during a Contingency Management Procedure to Reduce Excessive Alcohol Use

    PubMed Central

    Dougherty, Donald M.; Hill-Kapturczak, Nathalie; Liang, Yuanyuan; Karns, Tara E.; Cates, Sharon E.; Lake, Sarah L.; Mullen, Jillian; Roache, John D.

    2014-01-01

    Background Research on contingency management to treat excessive alcohol use is limited due to feasibility issues with monitoring adherence. This study examined the effectiveness of using transdermal alcohol monitoring as a continuous measure of alcohol use to implement financial contingencies to reduce heavy drinking. Methods Twenty-six male and female drinkers (from 21–39 years old) were recruited from the community. Participants were randomly assigned to one of two treatment sequences. Sequence 1 received 4 weeks of no financial contingency (i.e., $0) drinking followed by 4 weeks each of $25 and then $50 contingency management; Sequence 2 received 4 weeks of $25 contingency management followed by 4 weeks each of no contingency (i.e., $0) and then $50 contingency management. During the $25 and $50 contingency management conditions, participants were paid each week when the Secure Continuous Remote Alcohol Monitor (SCRAM-II™) identified no heavy drinking days. Results Participants in both contingency management conditions had fewer drinking episodes and reduced frequencies of heavy drinking compared to the $0 condition. Participants randomized to Sequence 2 (receiving $25 contingency before the $0 condition) exhibited less frequent drinking and less heavy drinking in the $0 condition compared to participants from Sequence 1. Conclusions Transdermal alcohol monitoring can be used to implement contingency management programs to reduce excessive alcohol consumption. PMID:25064019

  15. Final Report for the “WSU Neutron Capture Therapy Facility Support”

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Gerald E. Tripard; Keith G. Fox

    2006-08-24

    The objective for the cooperative research program for which this report has been written was to provide separate NCT facility user support for the students, faculty and scientists who would be doing the U.S. Department of Energy Office (DOE) of Science supported advanced radiotargeted research at the WSU 1 megawatt TRIGA reactor. The participants were the Idaho National laboratory (INL, P.I., Dave Nigg), the Veterinary Medical Research Center of Washington State University (WSU, Janean Fidel and Patrick Gavin), and the Washington State University Nuclear Radiation Center (WSU, P.I., Gerald Tripard). A significant number of DOE supported modifications were made tomore » the WSU reactor in order to create an epithermal neutron beam while at the same time maintaining the other activities of the 1 MW reactor. These modifications were: (1) Removal of the old thermal column. (2) Construction and insertion of a new epithermal filter, collimator and shield. (3) Construction of a shielded room that could accommodate the very high radiation field created by an intense neutron beam. (4) Removal of the previous reactor core fuel cluster arrangement. (5) Design and loading of the new reactor core fuel cluster arrangement in order to optimize the neutron flux entering the epithermal neutron filter. (6) The integration of the shielded rooms interlocks and radiological controls into the SCRAM chain and operating electronics of the reactor. (7) Construction of a motorized mechanism for moving and remotely controlling the position of the entire reactor bridge. (8) The integration of the reactor bridge control electronics into the SCRAM chain and operating electronics of the reactor. (9) The design, construction and attachment to the support structure of the reactor of an irradiation box that could be inserted into position next to the face of the reactor. (Necessitated by the previously mentioned core rearrangement). All of the above modifications were successfully completed and tested. The resulting epithermal beam of 1 x 10{sup 9} n/sec-cm{sup 2} was measured by Idaho National Laboratory with assistance from WSU's Neutron Activation Analysis Group. The beam is as good as our initial proposals for the project had predicted. In addition to all of the design, construction and insertion of the hardware, shielding, electronics and radiation monitoring systems there was considerable manpower and effort put into changes in the Technical Specifications of the reactor and implementing procedures for use of the new facility. This staff involvement is one of the reasons we requested special facility support from the DOE. Once the facility was competed and all of the recalibrations and measurements made to characterize the differences between this reactor core and the previous core we began to assist INL in making their beam measurements with foils and phantoms. Although we proposed support for only one additional staff position to support this new NCT facility the staff support provided by the WSU Nuclear Radiation Center was greater than had been anticipated by our initial proposal. INL was also assisted in the testing of a heavy water (deuterated water) bladder that can be inserted into the collimator in order to produce an intense, external thermal neutron beam. The external epithermal and/or thermal neutron beam capability remains available for use, if funding becomes available for future research projects.« less

  16. Affordable Flight Demonstration of the GTX Air-Breathing SSTO Vehicle Concept

    NASA Technical Reports Server (NTRS)

    Krivanek, Thomas M.; Roche, Joseph M.; Riehl, John P.; Kosareo, Daniel N.

    2002-01-01

    The rocket based combined cycle (RBCC) powered single-stage-to-orbit (SSTO) reusable launch vehicle has the potential to significantly reduce the total cost per pound for orbital payload missions. To validate overall system performance, a flight demonstration must be performed. This paper presents an overview of the first phase of a flight demonstration program for the GTX SSTO vehicle concept. Phase 1 will validate the propulsion performance of the vehicle configuration over the supersonic and hypersonic airbreathing portions of the trajectory. The focus and goal of Phase 1 is to demonstrate the integration and performance of the propulsion system flowpath with the vehicle aerodynamics over the air-breathing trajectory. This demonstrator vehicle will have dual mode ramjet/scramjets, which include the inlet, combustor, and nozzle with geometrically scaled aerodynamic surface outer mold lines (OML) defining the forebody, boundary layer diverter, wings, and tail. The primary objective of this study is to demonstrate propulsion system performance and operability including the ram to scram transition, as well as to validate vehicle aerodynamics and propulsion airframe integration. To minimize overall risk and development cost the effort will incorporate proven materials, use existing turbomachinery in the propellant delivery systems, launch from an existing unmanned remote launch facility, and use basic vehicle recovery techniques to minimize control and landing requirements. A second phase would demonstrate propulsion performance across all critical portions of a space launch trajectory (lift off through transition to all-rocket) integrated with flight-like vehicle systems.

  17. Affordable Flight Demonstration of the GTX Air-Breathing SSTO Vehicle Concept

    NASA Technical Reports Server (NTRS)

    Krivanek, Thomas M.; Roche, Joseph M.; Riehl, John P.; Kosareo, Daniel N.

    2003-01-01

    The rocket based combined cycle (RBCC) powered single-stage-to-orbit (SSTO) reusable launch vehicle has the potential to significantly reduce the total cost per pound for orbital payload missions. To validate overall system performance, a flight demonstration must be performed. This paper presents an overview of the first phase of a flight demonstration program for the GTX SSTO vehicle concept. Phase 1 will validate the propulsion performance of the vehicle configuration over the supersonic and hypersonic air- breathing portions of the trajectory. The focus and goal of Phase 1 is to demonstrate the integration and performance of the propulsion system flowpath with the vehicle aerodynamics over the air-breathing trajectory. This demonstrator vehicle will have dual mode ramjetkcramjets, which include the inlet, combustor, and nozzle with geometrically scaled aerodynamic surface outer mold lines (OML) defining the forebody, boundary layer diverter, wings, and tail. The primary objective of this study is to demon- strate propulsion system performance and operability including the ram to scram transition, as well as to validate vehicle aerodynamics and propulsion airframe integration. To minimize overall risk and develop ment cost the effort will incorporate proven materials, use existing turbomachinery in the propellant delivery systems, launch from an existing unmanned remote launch facility, and use basic vehicle recovery techniques to minimize control and landing requirements. A second phase would demonstrate propulsion performance across all critical portions of a space launch trajectory (lift off through transition to all-rocket) integrated with flight-like vehicle systems.

  18. Reactor Operations Monitoring System

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hart, M.M.

    1989-01-01

    The Reactor Operations Monitoring System (ROMS) is a VME based, parallel processor data acquisition and safety action system designed by the Equipment Engineering Section and Reactor Engineering Department of the Savannah River Site. The ROMS will be analyzing over 8 million signal samples per minute. Sixty-eight microprocessors are used in the ROMS in order to achieve a real-time data analysis. The ROMS is composed of multiple computer subsystems. Four redundant computer subsystems monitor 600 temperatures with 2400 thermocouples. Two computer subsystems share the monitoring of 600 reactor coolant flows. Additional computer subsystems are dedicated to monitoring 400 signals from assortedmore » process sensors. Data from these computer subsystems are transferred to two redundant process display computer subsystems which present process information to reactor operators and to reactor control computers. The ROMS is also designed to carry out safety functions based on its analysis of process data. The safety functions include initiating a reactor scram (shutdown), the injection of neutron poison, and the loadshed of selected equipment. A complete development Reactor Operations Monitoring System has been built. It is located in the Program Development Center at the Savannah River Site and is currently being used by the Reactor Engineering Department in software development. The Equipment Engineering Section is designing and fabricating the process interface hardware. Upon proof of hardware and design concept, orders will be placed for the final five systems located in the three reactor areas, the reactor training simulator, and the hardware maintenance center.« less

  19. Application Of The Iberdrola Licensing Methodology To The Cofrentes BWR-6 110% Extended Power Up-rate

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mata, Pedro; Fuente, Rafael de la; Iglesias, Javier

    Iberdrola (spanish utility) and Iberdrola Ingenieria (engineering branch) have been developing during the last two years the 110% Extended Power Up-rate Project (EPU 110%) for Cofrentes BWR-6. IBERDROLA has available an in-house design and licensing reload methodology that has been approved by the Spanish Nuclear Regulatory Authority. This methodology has been already used to perform the nuclear design and the reload licensing analysis for Cofrentes cycles 12 to 14. The methodology has been also applied to develop a significant number of safety analysis of the Cofrentes Extended Power Up-rate including: Reactor Heat Balance, Core and Fuel performance, Thermal Hydraulic Stability,more » ECCS LOCA Evaluation, Transient Analysis, Anticipated Transient Without Scram (ATWS) and Station Blackout (SBO) Since the scope of the licensing process of the Cofrentes Extended Power Up-rate exceeds the range of analysis included in the Cofrentes generic reload licensing process, it has been required to extend the applicability of the Cofrentes licensing methodology to the analysis of new transients. This is the case of the TLFW transient. The content of this paper shows the benefits of having an in-house design and licensing methodology, and describes the process to extend the applicability of the methodology to the analysis of new transients. The case of analysis of Total Loss of Feedwater with the Cofrentes Retran Model is included as an example of this process. (authors)« less

  20. Plum Brook Reactor Facility Control Room during Facility Startup

    NASA Image and Video Library

    1961-02-21

    Operators test the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility systems in the months leading up to its actual operation. The “Reactor On” signs are illuminated but the reactor core was not yet ready for chain reactions. Just a couple weeks after this photograph, Plum Brook Station held a media open house to unveil the 60-megawatt test reactor near Sandusky, Ohio. More than 60 members of the print media and radio and television news services met at the site to talk with community leaders and representatives from NASA and Atomic Energy Commission. The Plum Brook reactor went critical for the first time on the evening of June 14, 1961. It was not until April 1963 that the reactor reached its full potential of 60 megawatts. The reactor control room, located on the second floor of the facility, was run by licensed operators. The operators manually operated the shim rods which adjusted the chain reaction in the reactor core. The regulating rods could partially or completely shut down the reactor. The control room also housed remote area monitoring panels and other monitoring equipment that allowed operators to monitor radiation sensors located throughout the facility and to scram the reactor instantly if necessary. The color of the indicator lights corresponded with the elevation of the detectors in the various buildings. The reactor could also shut itself down automatically if the monitors detected any sudden irregularities.

  1. Rapid depressurization event analysis in BWR/6 using RELAP5 and contain

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mueftueoglu, A.K.; Feltus, M.A.

    1995-09-01

    Noncondensable gases may become dissolved in Boiling Water Reactor (BWR) water level instrumentation during normal operations. Any dissolved noncondensable gases inside these water columns may come out of solution during rapid depressurization events, and displace water from the reference leg piping resulting in a false high level. These water level errors may cause a delay or failure in actuation, or premature shutdown of the Emergency Core Cooling System. (ECCS). If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response and othermore » signals for automatic actuation such as high drywell pressure. It is also important to determine the effect of the level signal on ECCS operation after it is being actuated. The objective of this study is to determine the detailed coupled containment/NSSS response during this rapid depressurization events in BWR/6. The selected scenarios involve: (a) inadvertent opening of all ADS valves, (b) design basis (DB) large break loss of coolant accident (LOCA), and (c) main steam line break (MSLB). The transient behaviors are evaluated in terms of: (a) vessel pressure and collapsed water level response, (b) specific transient boundary conditions, (e.g., scram, MSIV closure timing, feedwater flow, and break blowdown rates), (c) ECCS initiation timing, (d) impact of operator actions, (e) whether indications besides low-low water level were available. The results of the analysis had shown that there would be signals to actuate ECCS other than low reactor level, such as high drywell pressure, low vessel pressure, high suppression pool temperature, and that the plant operators would have significant indications to actuate ECCS.« less

  2. Benchmark Simulation of Natural Circulation Cooling System with Salt Working Fluid Using SAM

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ahmed, K. K.; Scarlat, R. O.; Hu, R.

    Liquid salt-cooled reactors, such as the Fluoride Salt-Cooled High-Temperature Reactor (FHR), offer passive decay heat removal through natural circulation using Direct Reactor Auxiliary Cooling System (DRACS) loops. The behavior of such systems should be well-understood through performance analysis. The advanced system thermal-hydraulics tool System Analysis Module (SAM) from Argonne National Laboratory has been selected for this purpose. The work presented here is part of a larger study in which SAM modeling capabilities are being enhanced for the system analyses of FHR or Molten Salt Reactors (MSR). Liquid salt thermophysical properties have been implemented in SAM, as well as properties ofmore » Dowtherm A, which is used as a simulant fluid for scaled experiments, for future code validation studies. Additional physics modules to represent phenomena specific to salt-cooled reactors, such as freezing of coolant, are being implemented in SAM. This study presents a useful first benchmark for the applicability of SAM to liquid salt-cooled reactors: it provides steady-state and transient comparisons for a salt reactor system. A RELAP5-3D model of the Mark-1 Pebble-Bed FHR (Mk1 PB-FHR), and in particular its DRACS loop for emergency heat removal, provides steady state and transient results for flow rates and temperatures in the system that are used here for code-to-code comparison with SAM. The transient studied is a loss of forced circulation with SCRAM event. To the knowledge of the authors, this is the first application of SAM to FHR or any other molten salt reactors. While building these models in SAM, any gaps in the code’s capability to simulate such systems are identified and addressed immediately, or listed as future improvements to the code.« less

  3. Modeling Transients and Designing a Passive Safety System for a Nuclear Thermal Rocket Using Relap5

    NASA Astrophysics Data System (ADS)

    Khatry, Jivan

    Long-term high payload missions necessitate the need for nuclear space propulsion. Several nuclear reactor types were investigated by the Nuclear Engine for Rocket Vehicle Application (NERVA) program of National Aeronautics and Space Administration (NASA). Study of planned/unplanned transients on nuclear thermal rockets is important due to the need for long-term missions. A NERVA design known as the Pewee I was selected for this purpose. The following transients were run: (i) modeling of corrosion-induced blockages on the peripheral fuel element coolant channels and their impact on radiation heat transfer in the core, and (ii) modeling of loss-of-flow-accidents (LOFAs) and their impact on radiation heat transfer in the core. For part (i), the radiation heat transfer rate of blocked channels increases while their neighbors' decreases. For part (ii), the core radiation heat transfer rate increases while the flow rate through the rocket system is decreased. However, the radiation heat transfer decreased while there was a complete LOFA. In this situation, the peripheral fuel element coolant channels handle the majority of the radiation heat transfer. Recognizing the LOFA as the most severe design basis accident, a passive safety system was designed in order to respond to such a transient. This design utilizes the already existing tie rod tubes and connects them to a radiator in a closed loop. Hence, this is basically a secondary loop. The size of the core is unchanged. During normal steady-state operation, this secondary loop keeps the moderator cool. Results show that the safety system is able to remove the decay heat and prevent the fuel elements from melting, in response to a LOFA and subsequent SCRAM.

  4. Noble gas, iodine, and cesium transport in a postulated loss of decay heat removal accident at Browns Ferry

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wichner, R.P.; Hodge, S.A.; Weber, C.F.

    1984-08-01

    This report presents an analysis of the movement of noble gas, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto themore » drywell floor. The analysis of fission product transport presented in this report is based on the no-operator-action sequence and provides an estimate of fission product inventories, as a function of time, within 14 control volumes outside the core, with the atmosphere considered as the final control volume in the transport sequence. As in the case of accident sequences previously studied, we find small barrier for noble gas ejection to air, these gases being effectively purged from the drywell and reactor building by steam and concrete degradation gases. However, significant decay of krypton isotopes occurs during the long delay times involved in this sequence. In contrast, large degrees of holdup for iodine and cesium are projected due to the chemical reactivity of these elements. Only about 2 x 10/sup -4/% of the initial iodine and cesium activity are predicted to be released to the atmosphere. Principal barriers for release are deposition on reactor vessel and containment walls. A significant amount of iodine is captured in the water pool formed in the reactor building basement after actuation of the fire protection system.« less

  5. A Collaborative Analysis Tool for Integrating Hypersonic Aerodynamics, Thermal Protection Systems, and RBCC Engine Performance for Single Stage to Orbit Vehicles

    NASA Technical Reports Server (NTRS)

    Stanley, Thomas Troy; Alexander, Reginald

    1999-01-01

    Presented is a computer-based tool that connects several disciplines that are needed in the complex and integrated design of high performance reusable single stage to orbit (SSTO) vehicles. Every system is linked to every other system, as is the case of SSTO vehicles with air breathing propulsion, which is currently being studied by NASA. The deficiencies in the scramjet powered concept led to a revival of interest in Rocket-Based Combined-Cycle (RBCC) propulsion systems. An RBCC propulsion system integrates airbreathing and rocket propulsion into a single engine assembly enclosed within a cowl or duct. A typical RBCC propulsion system operates as a ducted rocket up to approximately Mach 3. At this point the transitions to a ramjet mode for supersonic-to-hypersonic acceleration. Around Mach 8 the engine transitions to a scram4jet mode. During the ramjet and scramjet modes, the integral rockets operate as fuel injectors. Around Mach 10-12 (the actual value depends on vehicle and mission requirements), the inlet is physically closed and the engine transitions to an integral rocket mode for orbit insertion. A common feature of RBCC propelled vehicles is the high degree of integration between the propulsion system and airframe. At high speeds the vehicle forebody is fundamentally part of the engine inlet, providing a compression surface for air flowing into the engine. The compressed air is mixed with fuel and burned. The combusted mixture must be expanded to an area larger than the incoming stream to provide thrust. Since a conventional nozzle would be too large, the entire lower after body of the vehicle is used as an expansion surface. Because of the high external temperatures seen during atmospheric flight, the design of an airbreathing SSTO vehicle requires delicate tradeoffs between engine design, vehicle shape, and thermal protection system (TPS) sizing in order to produce an optimum system in terms of weight (and cost) and maximum performance.

  6. Assessment of the SFC database for analysis and modeling

    NASA Technical Reports Server (NTRS)

    Centeno, Martha A.

    1994-01-01

    SFC is one of the four clusters that make up the Integrated Work Control System (IWCS), which will integrate the shuttle processing databases at Kennedy Space Center (KSC). The IWCS framework will enable communication among the four clusters and add new data collection protocols. The Shop Floor Control (SFC) module has been operational for two and a half years; however, at this stage, automatic links to the other 3 modules have not been implemented yet, except for a partial link to IOS (CASPR). SFC revolves around a DB/2 database with PFORMS acting as the database management system (DBMS). PFORMS is an off-the-shelf DB/2 application that provides a set of data entry screens and query forms. The main dynamic entity in the SFC and IOS database is a task; thus, the physical storage location and update privileges are driven by the status of the WAD. As we explored the SFC values, we realized that there was much to do before actually engaging in continuous analysis of the SFC data. Half way into this effort, it was realized that full scale analysis would have to be a future third phase of this effort. So, we concentrated on getting to know the contents of the database, and in establishing an initial set of tools to start the continuous analysis process. Specifically, we set out to: (1) provide specific procedures for statistical models, so as to enhance the TP-OAO office analysis and modeling capabilities; (2) design a data exchange interface; (3) prototype the interface to provide inputs to SCRAM; and (4) design a modeling database. These objectives were set with the expectation that, if met, they would provide former TP-OAO engineers with tools that would help them demonstrate the importance of process-based analyses. The latter, in return, will help them obtain the cooperation of various organizations in charting out their individual processes.

  7. Nuclear reactor I

    DOEpatents

    Ference, Edward W.; Houtman, John L.; Waldby, Robert N.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor whose upper internals include provision for channeling the liquid metal flowing from the core-component assemblies to the outlet plenum in vertical paths in direction generally along the direction of the respective assemblies. The metal is channeled by chimneys, each secured to, and extending from, a grid through whose openings the metal emitted by a plurality of core-component assemblies encompassed by the grid flows. To reduce the stresses resulting from structural interaction, or the transmissive of thermal strains due to large temperature differences in the liquid metal emitted from neighboring core-component assemblies, throughout the chimneys and the other components of the upper internals, the grids and the chimneys are supported from the heat plate and the core barrel by support columns (double portal support) which are secured to the head plate at the top and to a member, which supports the grids and is keyed to the core barrel, at the bottom. In addition to being restrained from lateral flow by the chimneys, the liquid metal is also restrained from flowing laterally by a peripheral seal around the top of the core. This seal limits the flow rate of liquid metal, which may be sharply cooled during a scram, to the outlet nozzles. The chimneys and the grids are formed of a highly-refractory, high corrosion-resistant nickel-chromium-iron alloy which can withstand the stresses produced by temperature differences in the liquid metal. The chimneys are supported by pairs of plates, each pair held together by hollow stubs coaxial with, and encircling, the chimneys. The plates and stubs are a welded structure but, in the interest of economy, are composed of stainless steel which is not weld compatible with the refractory metal. The chimneys and stubs are secured together by shells of another nickel-chromium-iron alloy which is weld compatible with, and is welded to, the stubs and has about the same coefficient of expansion as the highly-refractory, high corrosion-resistant alloy.

  8. Control Activity in Support of NASA Turbine Based Combined Cycle (TBCC) Research

    NASA Technical Reports Server (NTRS)

    Stueber, Thomas J.; Vrnak, Daniel R.; Le, Dzu K.; Ouzts, Peter J.

    2010-01-01

    Control research for a Turbine Based Combined Cycle (TBCC) propulsion system is the current focus of the Hypersonic Guidance, Navigation, and Control (GN&C) discipline team. The ongoing work at the NASA Glenn Research Center (GRC) supports the Hypersonic GN&C effort in developing tools to aid the design of control algorithms to manage a TBCC airbreathing propulsion system during a critical operating period. The critical operating period being addressed in this paper is the span when the propulsion system transitions from one cycle to another, referred to as mode transition. One such tool, that is a basic need for control system design activities, is computational models (hereto forth referred to as models) of the propulsion system. The models of interest for designing and testing controllers are Control Development Models (CDMs) and Control Validation Models (CVMs). CDMs and CVMs are needed for each of the following propulsion system elements: inlet, turbine engine, ram/scram dual-mode combustor, and nozzle. This paper presents an overall architecture for a TBCC propulsion system model that includes all of the propulsion system elements. Efforts are under way, focusing on one of the propulsion system elements, to develop CDMs and CVMs for a TBCC propulsion system inlet. The TBCC inlet aerodynamic design being modeled is that of the Combined-Cycle Engine (CCE) Testbed. The CCE Testbed is a large-scale model of an aerodynamic design that was verified in a small-scale screening experiment. The modeling approach includes employing existing state-of-the-art simulation codes, developing new dynamic simulations, and performing system identification experiments on the hardware in the NASA GRC 10 by10-Foot Supersonic Wind Tunnel. The developed CDMs and CVMs will be available for control studies prior to hardware buildup. The system identification experiments on the CCE Testbed will characterize the necessary dynamics to be represented in CDMs for control design. These system identification models will also be the reference models to validate the CDM and CVM models. Validated models will give value to the tools used to develop the models.

  9. PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Murray, J.L.

    1961-02-01

    BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

  10. Bayesian network representing system dynamics in risk analysis of nuclear systems

    NASA Astrophysics Data System (ADS)

    Varuttamaseni, Athi

    2011-12-01

    A dynamic Bayesian network (DBN) model is used in conjunction with the alternating conditional expectation (ACE) regression method to analyze the risk associated with the loss of feedwater accident coupled with a subsequent initiation of the feed and bleed operation in the Zion-1 nuclear power plant. The use of the DBN allows the joint probability distribution to be factorized, enabling the analysis to be done on many simpler network structures rather than on one complicated structure. The construction of the DBN model assumes conditional independence relations among certain key reactor parameters. The choice of parameter to model is based on considerations of the macroscopic balance statements governing the behavior of the reactor under a quasi-static assumption. The DBN is used to relate the peak clad temperature to a set of independent variables that are known to be important in determining the success of the feed and bleed operation. A simple linear relationship is then used to relate the clad temperature to the core damage probability. To obtain a quantitative relationship among different nodes in the DBN, surrogates of the RELAP5 reactor transient analysis code are used. These surrogates are generated by applying the ACE algorithm to output data obtained from about 50 RELAP5 cases covering a wide range of the selected independent variables. These surrogates allow important safety parameters such as the fuel clad temperature to be expressed as a function of key reactor parameters such as the coolant temperature and pressure together with important independent variables such as the scram delay time. The time-dependent core damage probability is calculated by sampling the independent variables from their probability distributions and propagate the information up through the Bayesian network to give the clad temperature. With the knowledge of the clad temperature and the assumption that the core damage probability has a one-to-one relationship to it, we have calculated the core damage probably as a function of transient time. The use of the DBN model in combination with ACE allows risk analysis to be performed with much less effort than if the analysis were done using the standard techniques.

  11. Characterizing Hohlraum Plasma Conditions at the National Ignition Facility (NIF) Using X-ray Spectroscopy

    NASA Astrophysics Data System (ADS)

    Barrios, Maria Alejandra

    2015-11-01

    Improved hohlraums will have a significant impact on increasing the likelihood of indirect drive ignition at the NIF. In indirect-drive Inertial Confinement Fusion (ICF), a high-Z hohlraum converts laser power into a tailored x-ray flux that drives the implosion of a spherical capsule filled with D-T fuel. The x-radiation drive to capsule coupling sets the velocity, adiabat, and symmetry of the implosion. Previous experiments in gas-filled hohlraums determined that the laser-hohlraum energy coupling is 20-25% less than modeled, therefore identifying energy loss mechanisms that reduce the efficacy of the hohlraum drive is central to improving implosion performance. Characterizing the plasma conditions, particularly the plasma electron temperature (Te) , is critical to understanding mechanism that affect the energy coupling such as the laser plasma interactions (LPI), hohlraum x-ray conversion efficiency, and dynamic drive symmetry. The first Te measurements inside a NIF hohlraum, presented here, were achieved using K-shell X-ray spectroscopy of an Mn-Co tracer dot. The dot is deposited on a thin-walled CH capsule, centered on the hohlraum symmetry axis below the laser entrance hole (LEH) of a bottom-truncated hohlraum. The hohlraum x-ray drive ablates the dot and causes it to flow upward, towards the LEH, entering the hot laser deposition region. An absolutely calibrated streaked spectrometer with a line of sight into the LEH records the temporal history of the Mn and Co X-ray emission. The measured (interstage) Lyα/ Heα line ratios for Co and Mn and the Mn-Heα/Co-Heα isoelectronic line ratio are used to infer the local plasma Te from the atomic physics code SCRAM. Time resovled x-ray images perpendicular to the hohlraum axis record the dot expansion and trajectory into the LEH region. The temporal evolution of the measured Te and dot trajectory are compared with simulations from radiation-hydrodynamic codes. This work was performed under the auspices of the U.S. Department of Energy by LLNL under Contract DE-AC52-07NA27344.

  12. Development of an inconel self powered neutron detector for in-core reactor monitoring

    NASA Astrophysics Data System (ADS)

    Alex, M.; Ghodgaonkar, M. D.

    2007-04-01

    The paper describes the development and testing of an Inconel600 (2 mm diameter×21 cm long) self-powered neutron detector for in-core neutron monitoring. The detector has 3.5 mm overall diameter and 22 cm length and is integrally coupled to a 12 m long mineral insulated cable. The performance of the detector was compared with cobalt and platinum detectors of similar dimensions. Gamma sensitivity measurements performed at the 60Co irradiation facility in 14 MR/h gamma field showed values of -4.4×10 -18 A/R/h/cm (-9.3×10 -24 A/ γ/cm 2-s/cm), -5.2×10 -18 A/R/h/cm (-1.133×10 -23 A/ γ/cm 2-s/cm) and 34×10 -18 A/R/h/cm (7.14×10 -23 A/ γ/cm 2-s/cm) for the Inconel, Co and Pt detectors, respectively. The detectors together with a miniature gamma ion chamber and fission chamber were tested in the in-core Apsara Swimming Pool type reactor. The ion chambers were used to estimate the neutron and gamma fields. With an effective neutron cross-section of 4b, the Inconel detector has a total sensitivity of 6×10 -23 A/nv/cm while the corresponding sensitivities for the platinum and cobalt detectors were 1.69×10 -22 and 2.64×10 -22 A/nv/cm. The linearity of the detector responses at power levels ranging from 100 to 200 kW was within ±5%. The response of the detectors to reactor scram showed that the prompt response of the Inconel detector was 0.95 while it was 0.7 and 0.95 for the platinum and cobalt self-powered detectors, respectively. The detector was also installed in the horizontal flux unit of 540 MW Pressurised Heavy Water Reactor (PHWR). The neutron flux at the detector location was calculated by Triveni code. The detector response was measured from 0.02% to 0.07% of full power and showed good correlation between power level and detector signals. Long-term tests and the dynamic response of the detector to shut down in PHWR are in progress.

  13. Simulation of particle diversity and mixing state over Greater Paris: a model-measurement inter-comparison.

    PubMed

    Zhu, Shupeng; Sartelet, Karine N; Healy, Robert M; Wenger, John C

    2016-07-18

    Air quality models are used to simulate and forecast pollutant concentrations, from continental scales to regional and urban scales. These models usually assume that particles are internally mixed, i.e. particles of the same size have the same chemical composition, which may vary in space and time. Although this assumption may be realistic for continental-scale simulations, where particles originating from different sources have undergone sufficient mixing to achieve a common chemical composition for a given model grid cell and time, it may not be valid for urban-scale simulations, where particles from different sources interact on shorter time scales. To investigate the role of the mixing state assumption on the formation of particles, a size-composition resolved aerosol model (SCRAM) was developed and coupled to the Polyphemus air quality platform. Two simulations, one with the internal mixing hypothesis and another with the external mixing hypothesis, have been carried out for the period 15 January to 11 February 2010, when the MEGAPOLI winter field measurement campaign took place in Paris. The simulated bulk concentrations of chemical species and the concentrations of individual particle classes are compared with the observations of Healy et al. (Atmos. Chem. Phys., 2013, 13, 9479-9496) for the same period. The single particle diversity and the mixing-state index are computed based on the approach developed by Riemer et al. (Atmos. Chem. Phys., 2013, 13, 11423-11439), and they are compared to the measurement-based analyses of Healy et al. (Atmos. Chem. Phys., 2014, 14, 6289-6299). The average value of the single particle diversity, which represents the average number of species within each particle, is consistent between simulation and measurement (2.91 and 2.79 respectively). Furthermore, the average value of the mixing-state index is also well represented in the simulation (69% against 59% from the measurements). The spatial distribution of the mixing-state index shows that the particles are not mixed in urban areas, while they are well mixed in rural areas. This indicates that the assumption of internal mixing traditionally used in transport chemistry models is well suited to rural areas, but this assumption is less realistic for urban areas close to emission sources.

  14. CVTR PROJECT. CAROLINAS VIRGINIA NUCLEAR POWER ASSOCIATES, INC. MONTHLY PROGRESS REPORT, MAY 1961

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    1961-10-31

    The capsule A-2 was removed from the WTR reflector hole at the end of the WTR Cycle 13, and was stored in the WTR canal. The in-pile loop has operated for eight months and the test thimble was irradiated a total of 108 days. Tensile tests were completed on the extruded and annealed Zircaloy-4 Phase-II pressure tubes. The tensile properties varied with location in the pressure tube. The lowest values were obtained in the top flange where the material was fully annealed for ten hours at 800 deg C. Increased properties were achieved from working the material during extrusion operations.more » A shielding ring is provided to prevent streaming through a void generated by the rotating shield volley supports. It was determined that an additional thickness of iron or steel is required to compensate for the loss of shielding from the removal of one foot of concrete at the bottom of the trench. Various portions of the U-tube and fuel assemblies were homogenized in various axial regions for computer studies. The studies indicated a decrease of 500 hours in core life from non-uniform axial burnup. Pressure tube specimens are being tested under the impulsive test burst program. A test specimen experienced a 51% increase in O.D. under 20 impact blows before it failed. Observations of the tested specimens indicated ductilities far in excess of those predicted from the material's behavior in uniaxial tension. Teste on a Zircaloy-stainless steel joint were concluded after an extensive program of testing under various pressure, temperature and bending moment conditions. No sign of leakage was noted throughout the program. Subsequent inspection of the joint showed cracks in the sleeve portion of the joint. Analysis of the test water indicated a chloride content of approx 88 ppm. A test fuel assembly was dismantled and converted to a four baffle design. Modifications were made to the prototype control-rod-drive system. The alignment between ths vertical and horizontal miter gears was improved by charging the mounting of horizontal shaft and bearings. Scram tests were resumed; these tests indicated that the dashpot was acting too soon. The dashpot is being modified. (auth)« less

  15. Delayed Gamma Measurements in Different Nuclear Research Reactors Bringing Out the Importance of the Delayed Contribution in Gamma Flux Calculations

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Fourmentel, D.; Radulovic, V.; Barbot, L.

    Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating ; a recent development atmore » the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC). The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR - 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA - 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR - 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is due to the delayed gamma rays. In this paper we describe experiments in each of the three reactors and how we estimate delayed gamma rays with MIC measurements. The results and perspectives are discussed. (authors)« less

  16. Activity ratios in soil contaminated by the source of different reactor condition in the FDNPP accident

    NASA Astrophysics Data System (ADS)

    Satou, Yukihiko; Sueki, Keisuke; Sasa, Kimikazu; Matsunaka, Tetsuya; Shibayama, Nao; Takahashi, Tsutomu; Kinoshita, Norikazu

    2014-05-01

    The Fukushima Dai-ichi Nuclear power plant (FDNPP) accident caused radioactive contamination on the surface soil at Fukushima and its adjacent prefectures. Substantial contamination has been found in the northwestern area from the FDNPP, according to the airborne monitoring and ground base survey by the Japanese government. Activity ratios would have characteristic information on emission sources because each relevant reactor had different amount of radionuclide and different activity ratio. The ratios can be used to clarify more detailed source and process in the contamination. We have addressed to consider them in Namie town, northwestern region from the FDNPP. This study focused on the gamma-ray emitting radionuclides of 134Cs, 137Cs, and 110mAg. The activities were decay-corrected as of 11th March, 2011 when all nuclear reactors scrammed. Data of activity ratios by our results and the Japanese official report classified the investigated northwestern region into 3 groups. Ratios of 0.02 for 110mAg/137Cs and 0.90 for 134Cs/137Cs were observed in the northern region of 15 km inside from the FDNPP. On the other hand, two kinds of 110mAg/137Cs ratios of 0.005 and 0.002 were distributed broadly in the region 60 km away from the plant. The 134Cs/137Cs ratio was 0.98 there. The activity ratios of 110mAg/137Cs and 134Cs/137Cs in the northern region from the FDNPP correspond to those of nuclear fuel in Unit 1 according to estimation using the ORIGEN code. The 134Cs/137Cs in the northwestern area from FDNPP agrees with that of Unit 2 and 3. The 110mAg/137Cs ratios of 0.005 and0.002 are 1/5 - 1/10 of the Unit 2 and 3. Official report has announced that discharges of the radionuclides from Unit 2 and 3 occurred on 14th March, 2011. It is known that contamination in the northwestern region from the FDNPP took place on 15th March, 2011. Plausible species for silver in reactor core, metal, and halide etc. have higher boiling point than those species for cesium. The core would be cooled down to lower temperature of the boiling point of silver at the timing contamination occurred. Thus, silver with higher boiling point was not much released than cesium with lower boiling point. The 110mAg/137Cs ratio has served to identify the specific sources of contamination in the northwestern area from the FDNPP.

  17. Measurement of a surface heat flux and temperature

    NASA Astrophysics Data System (ADS)

    Davis, R. M.; Antoine, G. J.; Diller, T. E.; Wicks, A. L.

    1994-04-01

    The Heat Flux Microsensor is a new sensor which was recently patented by Virginia Tech and is just starting to be marketed by Vatell Corp. The sensor is made using the thin-film microfabrication techniques directly on the material that is to be measured. It consists of several thin-film layers forming a differential thermopile across a thermal resistance layer. The measured heat flux q is proportional to the temperature difference across the resistance layer q= k(sub g)/delta(sub g) x (t(sub 1) - T(sub 2)), where k(sub g) is the thermal conductivity and delta (sub g) is the thickness of the thermal resistance layer. Because the gages are sputter coated directly onto the surface, their total thickness is less than 2 micrometers, which is two orders of magnitude thinner than previous gages. The resulting temperature difference across the thermal resistance layer (delta is less than 1 micrometer) is very small even at high heat fluxes. To generate a measurable signal many thermocouple pairs are put in series to form a differential thermopile. The combination of series thermocouple junctions and thin-film design creates a gage with very attractive characteristics. It is not only physically non-intrusive to the flow, but also causes minimal disruption of the surface temperature. Because it is so thin, the response time is less than 20 microsec. Consequently, the frequency response is flat from 0 to over 50 kHz. Moreover, the signal of the Heat Flux Microsensor is directly proportional to the heat flux. Therefore, it can easily be used in both steady and transient flows, and it measures both the steady and unsteady components of the surface heat flux. A version of the Heat Flux Microsensor has been developed to meet the harsh demands of combustion environments. These gages use platinum and platinum-10 percent rhodium as the thermoelectric materials. The thermal resistance layer is silicon monoxide and a protective coating of Al2O3 is deposited on top of the sensor. The superimposed thin-film pattern of all six layers is presented. The large pads are for connection with pins used to bring the signal out the back of the ceramic.

  18. Measurement of a surface heat flux and temperature

    NASA Technical Reports Server (NTRS)

    Davis, R. M.; Antoine, G. J.; Diller, T. E.; Wicks, A. L.

    1994-01-01

    The Heat Flux Microsensor is a new sensor which was recently patented by Virginia Tech and is just starting to be marketed by Vatell Corp. The sensor is made using the thin-film microfabrication techniques directly on the material that is to be measured. It consists of several thin-film layers forming a differential thermopile across a thermal resistance layer. The measured heat flux q is proportional to the temperature difference across the resistance layer q= k(sub g)/delta(sub g) x (t(sub 1) - T(sub 2)), where k(sub g) is the thermal conductivity and delta (sub g) is the thickness of the thermal resistance layer. Because the gages are sputter coated directly onto the surface, their total thickness is less than 2 micrometers, which is two orders of magnitude thinner than previous gages. The resulting temperature difference across the thermal resistance layer (delta is less than 1 micrometer) is very small even at high heat fluxes. To generate a measurable signal many thermocouple pairs are put in series to form a differential thermopile. The combination of series thermocouple junctions and thin-film design creates a gage with very attractive characteristics. It is not only physically non-intrusive to the flow, but also causes minimal disruption of the surface temperature. Because it is so thin, the response time is less than 20 microsec. Consequently, the frequency response is flat from 0 to over 50 kHz. Moreover, the signal of the Heat Flux Microsensor is directly proportional to the heat flux. Therefore, it can easily be used in both steady and transient flows, and it measures both the steady and unsteady components of the surface heat flux. A version of the Heat Flux Microsensor has been developed to meet the harsh demands of combustion environments. These gages use platinum and platinum-10 percent rhodium as the thermoelectric materials. The thermal resistance layer is silicon monoxide and a protective coating of Al2O3 is deposited on top of the sensor. The superimposed thin-film pattern of all six layers is presented. The large pads are for connection with pins used to bring the signal out the back of the ceramic. In addition to the heat flux measurement, the surface temperature is measured with a platinum resistance layer (RTS). The resistance of this layer increases with increasing temperature. Therefore, these gages simultaneously measure the surface temperature and heat flux. The demonstrated applications include rocket nozzles, SCRAM jet engines, gas turbine engines, boiling heat transfer, flame experiments, basic fluid heat transfer, hypersonic flight, and shock tube testing. The laboratory involves using one of these sensors in a small combustion flame. The sensor is made on a 2.5 cm diameter piece of aluminum nitride ceramic.

  19. NASA Hypersonic Propulsion: Overview of Progress from 1995 to 2005

    NASA Technical Reports Server (NTRS)

    Cikanek, Harry A., III; Bartolotta, Paul A.; Klem, Mark D.; Rausch, Vince L.

    2007-01-01

    Hypersonic propulsion work supported by the United States National Aeronautics and Space Administration had a primary focus on Space Transportation during the period from 1995 to 2005. The framework for these advances was established by policy and pursued with substantial funding. Many noteworthy advances were made, highlighted by the pinnacle flights of the X-43. This paper reviews and summarizes the programs and accomplishments of this era. The accomplishments are compared to the goals and objectives to lend an overarching perspective to what was achieved. At least dating back to the early days of the Space Shuttle program, NASA has had the objective of reducing the cost of access to space and concurrently improving safety and reliability. National Space Transportation Policy in 1994 coupled with a base of prior programs such as the National Aerospace Plane and the need to look beyond the Space Shuttle program set the stage for NASA to pursue Space Transportation Advances. Programs defined to pursue the advances represented a broad approach addressing classical rocket propulsion as well as airbreathing propulsion in various combinations and forms. The resulting portfolio of activities included systems analysis and design studies, discipline research and technology, component technology development, propulsion system ground test demonstration and flight demonstration. The types of propulsion systems that were pursued by these programs included classical rocket engines, "aerospike" rocket engines, high performance rocket engines, scram jets, rocket based combined cycles, and turbine based combined cycles. Vehicle architectures included single and two stage vehicles. Either single types of propulsion systems or combinations of the basic propulsion types were applied to both single and two stage vehicle design concepts. Some of the propulsion system design concepts were built and tested at full scale, large scale and small scale. Many flight demonstrators were conceptually defined, fewer designed and some built and one flown to demonstrate several technical advancements including propulsion. The X-43 flights were a culmination of these efforts for airbreathing propulsion. During the course of that period, there was a balance of funding and emphasis toward rocket propulsion but still very substantial airbreathing propulsion effort. The broad objectives of these programs were to both advance and test the state of the art so as to provide a basis for options to be pursued for broad space transportation needs, most importantly focused on crew carrying capability. NASA cooperated with the Department of Defense in planning and implementation of these programs to make efficient use of objectives and capabilities where appropriate. Much of the work was conducted in industry and academia as well as Government laboratories. Many test articles and data-bases now exist as a result of this work. At the conclusion of the period, the body of work made it clear that continued research and technology development was warranted, because although not ready for a NASA system development decision, results continued to support the promise of air-breathing propulsion for access to space.

  20. Combustion Efficiency, Flameout Operability Limits and General Design Optimization for Integrated Ramjet-Scramjet Hypersonic Vehicles

    NASA Astrophysics Data System (ADS)

    Mbagwu, Chukwuka Chijindu

    High speed, air-breathing hypersonic vehicles encounter a varied range of engine and operating conditions traveling along cruise/ascent missions at high altitudes and dynamic pressures. Variations of ambient pressure, temperature, Mach number, and dynamic pressure can affect the combustion conditions in conflicting ways. Computations were performed to understand propulsion tradeoffs that occur when a hypersonic vehicle travels along an ascent trajectory. Proper Orthogonal Decomposition methods were applied for the reduction of flamelet chemistry data in an improved combustor model. Two operability limits are set by requirements that combustion efficiency exceed selected minima and flameout be avoided. A method for flameout prediction based on empirical Damkohler number measurements is presented. Operability limits are plotted that define allowable flight corridors on an altitude versus flight Mach number performance map; fixed-acceleration ascent trajectories were considered for this study. Several design rules are also presented for a hypersonic waverider with a dual-mode scramjet engine. Focus is placed on ''vehicle integration" design, differing from previous ''propulsion-oriented" design optimization. The well-designed waverider falls between that of an aircraft (high lift-to-drag ratio) and a rocket (high thrust-to-drag ratio). 84 variations of an X-43-like vehicle were run using the MASIV scramjet reduced order model to examine performance tradeoffs. Informed by the vehicle design study, variable-acceleration trajectory optimization was performed for three constant dynamic pressures ascents. Computed flameout operability limits were implemented as additional constraints to the optimization problem. The Michigan-AFRL Scramjet In-Vehicle (MASIV) waverider model includes finite-rate chemistry, applied scaling laws for 3-D turbulent mixing, ram-scram transition and an empirical value of the flameout Damkohler number. A reduced-order modeling approach is justified (in lieu of higher-fidelity computational simulations) because all vehicle forces are computed multiple thousands of times to generate multi-dimensional performance maps. The findings of this thesis work present a number of compelling conclusions. It is found that the ideal operating conditions of a scramjet engine are heavily dependent on the ambient and combustor pressure (and less strongly on temperature). Combustor pressures of approximately 1.0 bar or greater achieve the highest combustion efficiency, in line with industry standards of more than 0.5 bar. Ascent trajectory analysis of combustion efficiency and lean-limit flameout dictate best operation at higher dynamic pressures and lower altitudes, but these goals are traded off by current structural limitations whereby dynamic pressures must remain below 100 kPa. Hypersonic waverider designs varied between an "airplane" and a "rocket" are found to have better performance with the latter design, with controllability and minimum elevon/rudder surface area as a stability constraint for the vehicle trim. Ultimately, these findings are beneficial and contribute to the overall understanding of dynamically stable waverider vehicles at hypersonic speeds. These types of vehicles have a range of applications from technology demonstration, to earth-to-low orbit payload transit, to most compellingly another step in the development and realization of viable supersonic commercial transport.

  1. 40-MW(E) PROTOTYPE HIGH-TEMPERATURE GAS-COOLED REACTOR RESEARCH AND DEVELOPMENT PROGRAM. Quarterly Progress Report for the Period Ending June 30, 1962

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    1963-10-31

    Research and development progress specifically directed toward the construction of a 40-Mw(e) prototype power plant employing a high-temperature, gas-cooled, graphitemoderated reactor known as the HTGR is reported. Irradiation of element III-B in the in-pile loop continued satisfactorily. The element has generated a total of l36.3 Mw-hr of fission heat. The gross activity in the purge stream increased slightly to about 350 mu C/cm/sup 3/. By taking larger gas samples than were previously taken, a value of 0.02 VC/cm/sup 3/ was obtained for the gross activity of the primary loop. Element III-A, which was removed from the loop after generating 133more » Mw-hr of fission heat, was disassembled and examined. No fuel-compact damage of any type was visible. Determination of the distribution of fission products in the element is under way, Fissionproduct- release data for in-pile-loop element III-A were calculated. During the 133 Mw- hr of operation, the release fraction increased by approximately one order of magnitude. Also calculated were the xenon and krypton release data for the first 100 Mw-hr of III-B operation. The release rate for the longer-lived isotopes increased bv about a factor of 10 and that of the shorter-lived isotopes by about a factor of 100. A test was run in which the in-pileloop purge flow, was stopped. The primariy-loop activity level rose sharply during the first hour, increased at a slower rate for the next 11 hr, and then appeared to level off. When purge flow was resumed, the gross activity in the primary loop was cleaned up with a half life of about 2.2 hr. An attempt was made to identify Cs/sup 137/ and Ba/ sup 140/ plateout in portions of the in-pile loop. A very small amount of cesium (less than a monolayer) was found, but no barium could be detected. The validity of two basic assumptions made in the one-dimensional burnup code FEVER was investigated. As a result of extensive lifetime studies and power-distribution and temperaturecoefficient calculations, the initial fuel loading for the Peach Bottom core was specified. A series of control-rodworth calculations and a recalculation of the postulated rod-fall accident were made for this loading. The test of the prototype control rod and drive was satisfactorily coNonempleted during the quarter. During the course of the test the drive completed 590,641 starts and stops, 5,756 scrams, and more than 2.6 million inches of random regulating motion in helium at reactor temperatures. These totals far exceed the expected life requirements of the system. Preparations are being made for testing the prototype emergency shutdown rod and drive. The apparatus for the barium permeation experiment with a full-diameter sleeve was completed, and preliminary calibration runs were started. Following these runs, the system will be operated until an equilibrium distribution of barium is reached. At that time, a series of corings will be made on all of the compacts and the sleeve to evaluate the overall barium and strontium distribution. Other experiments on barium behavior, including permeation experiments with reducedscale fuel elements and exVeriments on the vaporization, sorption, and diffusion of barium, were continued, and the data are being analyzed. Measurements were made to compare the room-temperature back diffusion of argon, krypton, and xenon through a sample of sleeve graphite against a helium pressure difference. The results show that the difference between the effective back-diffusion coefficients of krypton and xenon seems to increase with increasing helium pressure difference across the sleeve. The argon and krypton back-diffusion data at an average pressure of 3 atm are essentially the same, A FORTRAN code was written to recalculate the retention of neutron poison material and fission products in the core as well as their condensation on and revaporization from the upper reflector following a complete loss-of-coolant-circulation accident. (auth)« less

  2. The Acoustic Emission signal acquired by the microphones placed in the CABRI test device along the fourteen last R.I.A. experiments: an example of reproducible research in nuclear science

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Laurent Pantera, Oumar Traore

    The CABRI facility is an experimental nuclear reactor of the French Atomic Energy Commission. It is located at the Cadarache Research Centre in southern France and it is designed to act as a support to the French nuclear infrastructure. The purpose of the new testing programme termed, 'CABRI International Programme' (CIP) is to study the behaviour of PWR-type fuel rods at high burnup, equipped with an 'advanced' cladding, under Reactivity Initiated Accident (RIA) conditions (such as the scenario of a control rod ejection). Within the framework of this programme, piloted and funded by the French Institute of Nuclear Radioprotection andmore » Safety (IRSN), ten tests are to be conducted with a frequency of two tests per year. The LPRE laboratory of the CEA which is in charge of the Preparation, realisation and breakdown of the test results studies the possibility to set up a new test analysis based on the processing of signals coming from sensors placed within the test equipment. During the experimental phase, the behaviour of the fuel element generates acoustic waves which can be detected by two microphones placed upstream and downstream of the test device. Studies showed the interest to realize temporal and spectral analyses on these signals by showing the existence of signatures which can be correlated with physical phenomena as the rod failure or the test shutdown (i.e. scram). The work presented in this article results from the will to consolidate these studies. Since the main phenomenon to be tracked is the fuel rod failure, the aim would be to highlight specific events which would have been precursors of the rod failure in order to use in the future these signals for further interpretation. Such an antecedent works resumption leads to a better understanding of the experimental needs and constitutes a good initial state to prepare the new very fast digital data acquisition systems. Although all the raw data are accessible in the form of text files, analyses and graphics representations were not straightforward to reproduce from the ancient studies since that, on one hand, people who were in charge of the original work left the laboratory and on the other hand because it is not easy when the time passes, even with our own work, to be able to remember the steps of data manipulations and the exact setup: - During the ancient experiments the use of analog data acquisition systems required to digitize tapes to be able to realize computer treatments. That had had for consequence to lose the initial dating. This one must be correctly edited to do temporal comparisons. - Analyses require functions for calculations whose parameters has to be well-known to reach the same results. We thus wished to manage our workflow in the idea that it can be easily reproducible on all the experiments. The object of the work presented in this article was to put in practice this strong bind between the data, treatments and generation of the document in order not to hesitate to do the iteration principle in action. We do not have to be afraid by the data driven analyses. According to the philosophy of the literate programming, the text of the technical document is woven with the computer code that produces all the printed output as tables, graphs for the study eliminating hence the unrealistic cut and paste. This difficulty is not specific to the nuclear domain. For many years, researchers have been worked out solutions to this mundane issue. And, presently, new technologies and high-level programming languages offer us actual answers. We will firstly present the tools applied in our laboratory to implement this workflow, then we will describe the global perception carried out to continue the study of the Acoustic Emission signals recorded by the two microphones during the fourteen last CABRI R.I.A. test.« less

  3. Development of the ANL plant dynamics code and control strategies for the supercritical carbon dioxide Brayton cycle and code validation with data from the Sandia small-scale supercritical carbon dioxide Brayton cycle test loop.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Moisseytsev, A.; Sienicki, J. J.

    2011-11-07

    Significant progress has been made in the ongoing development of the Argonne National Laboratory (ANL) Plant Dynamics Code (PDC), the ongoing investigation and development of control strategies, and the analysis of system transient behavior for supercritical carbon dioxide (S-CO{sub 2}) Brayton cycles. Several code modifications have been introduced during FY2011 to extend the range of applicability of the PDC and to improve its calculational stability and speed. A new and innovative approach was developed to couple the Plant Dynamics Code for S-CO{sub 2} cycle calculations with SAS4A/SASSYS-1 Liquid Metal Reactor Code System calculations for the transient system level behavior onmore » the reactor side of a Sodium-Cooled Fast Reactor (SFR) or Lead-Cooled Fast Reactor (LFR). The new code system allows use of the full capabilities of both codes such that whole-plant transients can now be simulated without additional user interaction. Several other code modifications, including the introduction of compressor surge control, a new approach for determining the solution time step for efficient computational speed, an updated treatment of S-CO{sub 2} cycle flow mergers and splits, a modified enthalpy equation to improve the treatment of negative flow, and a revised solution of the reactor heat exchanger (RHX) equations coupling the S-CO{sub 2} cycle to the reactor, were introduced to the PDC in FY2011. All of these modifications have improved the code computational stability and computational speed, while not significantly affecting the results of transient calculations. The improved PDC was used to continue the investigation of S-CO{sub 2} cycle control and transient behavior. The coupled PDC-SAS4A/SASSYS-1 code capability was used to study the dynamic characteristics of a S-CO{sub 2} cycle coupled to a SFR plant. Cycle control was investigated in terms of the ability of the cycle to respond to a linear reduction in the electrical grid demand from 100% to 0% at a rate of 5%/minute. It was determined that utilization of turbine throttling control below 50% load improves the cycle efficiency significantly. Consequently, the cycle control strategy has been updated to include turbine throttle valve control. The new control strategy still relies on inventory control in the 50%-90% load range and turbine bypass for fine and fast generator output adjustments, but it now also includes turbine throttling control in the 0%-50% load range. In an attempt to investigate the feasibility of using the S-CO{sub 2} cycle for normal decay heat removal from the reactor, the cycle control study was extended beyond the investigation of normal load following. It was shown that such operation is possible with the extension of the inventory and the turbine throttling controls. However, the cycle operation in this range is calculated to be so inefficient that energy would need to be supplied from the electrical grid assuming that the generator could be capable of being operated in a motoring mode with an input electrical energy from the grid having a magnitude of about 20% of the nominal plant output electrical power level in order to maintain circulation of the CO{sub 2} in the cycle. The work on investigation of cycle operation at low power level will be continued in the future. In addition to the cycle control study, the coupled PDC-SAS4A/SASSYS-1 code system was also used to simulate thermal transients in the sodium-to-CO{sub 2} heat exchanger. Several possible conditions with the potential to introduce significant changes to the heat exchanger temperatures were identified and simulated. The conditions range from reactor scram and primary sodium pump failure or intermediate sodium pump failure on the reactor side to pipe breaks and valve malfunctions on the S-CO{sub 2} side. It was found that the maximum possible rate of the heat exchanger wall temperature change for the particular heat exchanger design assumed is limited to {+-}7 C/s for less than 10 seconds. Modeling in the Plant Dynamics Code has been compared with available data from the Sandia National Laboratories (SNL) small-scale S-CO{sub 2} Brayton cycle demonstration that is being assembled in a phased approach currently at Barber-Nichols Inc. and at SNL in the future. The available data was obtained with an earlier configuration of the S-CO{sub 2} loop involving only a single-turbo-alternator-compressor (TAC) instead of two TACs, a single low temperature recuperator (LTR) instead of both a LTR and a high temperature recuperator (HTR), and fewer than the later to be installed full set of electric heaters. Due to the absence of the full heating capability as well as the lack of a high temperature recuperator providing additional recuperation, the temperature conditions obtained with the loop are too low for the loop conditions to be prototypical of the S-CO{sub 2} cycle.« less

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Moisseytsev, A.; Sienicki, J. J.

    2011-04-12

    The analysis of specific control strategies and dynamic behavior of the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle has been extended to the two reactor types selected for continued development under the Generation IV Nuclear Energy Systems Initiative; namely, the Very High Temperature Reactor (VHTR) and the Sodium-Cooled Fast Reactor (SFR). Direct application of the standard S-CO{sub 2} recompression cycle to the VHTR was found to be challenging because of the mismatch in the temperature drop of the He gaseous reactor coolant through the He-to-CO{sub 2} reactor heat exchanger (RHX) versus the temperature rise of the CO{sub 2} through themore » RHX. The reference VHTR features a large temperature drop of 450 C between the assumed core outlet and inlet temperatures of 850 and 400 C, respectively. This large temperature difference is an essential feature of the VHTR enabling a lower He flow rate reducing the required core velocities and pressure drop. In contrast, the standard recompression S-CO{sub 2} cycle wants to operate with a temperature rise through the RHX of about 150 C reflecting the temperature drop as the CO{sub 2} expands from 20 MPa to 7.4 MPa in the turbine and the fact that the cycle is highly recuperated such that the CO{sub 2} entering the RHX is effectively preheated. Because of this mismatch, direct application of the standard recompression cycle results in a relatively poor cycle efficiency of 44.9%. However, two approaches have been identified by which the S-CO{sub 2} cycle can be successfully adapted to the VHTR and the benefits of the S-CO{sub 2} cycle, especially a significant gain in cycle efficiency, can be realized. The first approach involves the use of three separate cascaded S-CO{sub 2} cycles. Each S-CO{sub 2} cycle is coupled to the VHTR through its own He-to-CO{sub 2} RHX in which the He temperature is reduced by 150 C. The three respective cycles have efficiencies of 54, 50, and 44%, respectively, resulting in a net cycle efficiency of 49.3 %. The other approach involves reducing the minimum cycle pressure significantly below the critical pressure such that the temperature drop in the turbine is increased while the minimum cycle temperature is maintained above the critical temperature to prevent the formation of a liquid phase. The latter approach also involves the addition of a precooler and a third compressor before the main compressor to retain the benefits of compression near the critical point with the main compressor. For a minimum cycle pressure of 1 MPa, a cycle efficiency of 49.5% is achieved. Either approach opens up the door to applying the SCO{sub 2} cycle to the VHTR. In contrast, the SFR system typically has a core outlet-inlet temperature difference of about 150 C such that the standard recompression cycle is ideally suited for direct application to the SFR. The ANL Plant Dynamics Code has been modified for application to the VHTR and SFR when the reactor side dynamic behavior is calculated with another system level computer code such as SAS4A/SYSSYS-1 in the SFR case. The key modification involves modeling heat exchange in the RHX, accepting time dependent tabular input from the reactor code, and generating time dependent tabular input to the reactor code such that both the reactor and S-CO{sub 2} cycle sides can be calculated in a convergent iterative scheme. This approach retains the modeling benefits provided by the detailed reactor system level code and can be applied to any reactor system type incorporating a S-CO{sub 2} cycle. This approach was applied to the particular calculation of a scram scenario for a SFR in which the main and intermediate sodium pumps are not tripped and the generator is not disconnected from the electrical grid in order to enhance heat removal from the reactor system thereby enhancing the cooldown rate of the Na-to-CO{sub 2} RHX. The reactor side is calculated with SAS4A/SASSYS-1 while the S-CO{sub 2} cycle is calculated with the Plant Dynamics Code with a number of iterations over a timescale of 500 seconds. It is found that the RHX undergoes a maximum cooldown rate of {approx} -0.3 C/s. The Plant Dynamics Code was also modified to decrease its running time by replacing the compressible flow form of the momentum equation with an incompressible flow equation for use inside of the cooler or recuperators where the CO{sub 2} has a compressibility similar to that of a liquid. Appendices provide a quasi-static control strategy for a SFR as well as the self-adaptive linear function fitting algorithm developed to produce the tabular data for input to the reactor code and Plant Dynamics Code from the detailed output of the other code.« less

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