Sample records for boiling nuclear superheater

  1. Environmental Assessment for Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear Superheat (BONUS) Reactor Building, Rincon, Puerto Rico

    SciTech Connect (OSTI)

    N /A

    2003-02-24T23:59:59.000Z

    The U.S. Department of Energy (DOE) proposes to consent to a proposal by the Puerto Rico Electric Power Authority (PREPA) to allow public access to the Boiling Nuclear Superheat (BONUS) reactor building located near Rincon, Puerto Rico for use as a museum. PREPA, the owner of the BONUS facility, has determined that the historical significance of this facility, as one of only two reactors of this design ever constructed in the world, warrants preservation in a museum, and that this museum would provide economic benefits to the local community through increased tourism. Therefore, PREPA is proposing development of the BONUS facility as a museum.

  2. Conceptual design of an annular-fueled superheat boiling water reactor

    E-Print Network [OSTI]

    Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

    2011-01-01T23:59:59.000Z

    The conceptual design of an annular-fueled superheat boiling water reactor (ASBWR) is outlined. The proposed design, ASBWR, combines the boiler and superheater regions into one fuel assembly. This ensures good neutron ...

  3. Boiling of nuclear liquid in the micro-canonical ensemble

    E-Print Network [OSTI]

    -dependence of the nuclear symmetry energy and the critical temperature of 1 #12;Boiling of nuclear liquid in the microBoiling of nuclear liquid in the micro-canonical ensemble K. Miyazaki E-mail: miyazakiro@rio.odn.ne.jp Abstract New calculus of the liquid-gas phase transition is developed for the boiling of nuclear liquid

  4. 2010 Inspection and Status Report for the Boiling Nuclear Superheater (BONUS) Reactor Facility, Rincon, Puerto Rico

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTuba City, Arizona, DisposalFourthN V4100U.S.Indian10WayneAnnual Inspection

  5. Experimental design for study of nucleate boiling in porous structures

    E-Print Network [OSTI]

    Kelley, Mitchell Joseph

    2011-01-01T23:59:59.000Z

    The superheat required to initiate nucleate boiling inside porous wicks is not well understood in practice. This thesis reports the design of an experimental setup for investigating the onset of vapor nucleation in sintered ...

  6. Superheat Control: A Hybrid Approach

    E-Print Network [OSTI]

    Elliot, M.S.; Rasmussen, B.P.; Walton, Z.; Bolding, B.

    temperature. This is generally referred to as 5°C (9°F) of superheat. The metering of refrigerant in air conditioning, refrigeration, or heat-pump systems is gener- ally achieved by a number of different valve types, which vary in expense and design sophistica.... The proposed valve has the potential for widespread use throughout the air-conditioning, refrigeration, and heat-pump industry as an alternative to other actively controlled valves, while preserving sufficient flexibility to allow for passive flow control...

  7. Alternate Materials for Recovery Boiler Superheater Tubes

    SciTech Connect (OSTI)

    Keiser, James R [ORNL; Kish, Joseph [McMaster University; Singbeil, Douglas [FPInnovations

    2009-01-01T23:59:59.000Z

    The ever escalating demands for increased efficiency of all types of boilers would most sensibly be realized by an increase in the steam parameters of temperature and pressure. However, materials and corrosion limitations in the steam generating components, particularly the superheater tubes, present major obstacles to boiler designers in achieving systems that can operate under the more severe conditions. This paper will address the issues associated with superheater tube selection for many types of boilers; particularly chemical recovery boilers, but also addressing the similarities in issues for biomass and coal fired boilers. It will also review our recent study of materials for recovery boiler superheaters. Additional, more extensive studies, both laboratory and field, are needed to gain a better understanding of the variables that affect superheater tube corrosion and to better determine the best means to control this corrosion to ultimately permit operation of recovery boilers at higher temperatures and pressures.

  8. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    SciTech Connect (OSTI)

    Not Available

    1986-09-01T23:59:59.000Z

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised.

  9. Superheater Corrosion Produced By Biomass Fuels

    SciTech Connect (OSTI)

    Sharp, William (Sandy) [SharpConsultant] [SharpConsultant; Singbeil, Douglas [FPInnovations] [FPInnovations; Keiser, James R [ORNL] [ORNL

    2012-01-01T23:59:59.000Z

    About 90% of the world's bioenergy is produced by burning renewable biomass fuels. Low-cost biomass fuels such as agricultural wastes typically contain more alkali metals and chlorine than conventional fuels. Although the efficiency of a boiler's steam cycle can be increased by raising its maximum steam temperature, alkali metals and chlorine released in biofuel boilers cause accelerated corrosion and fouling at high superheater steam temperatures. Most alloys that resist high temperature corrosion protect themselves with a surface layer of Cr{sub 2}O{sub 3}. However, this Cr{sub 2}O{sub 3} can be fluxed away by reactions that form alkali chromates or volatilized as chromic acid. This paper reviews recent research on superheater corrosion mechanisms and superheater alloy performance in biomass boilers firing black liquor, biomass fuels, blends of biomass with fossil fuels and municipal waste.

  10. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    SciTech Connect (OSTI)

    NONE

    1995-08-01T23:59:59.000Z

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  11. Protective coatings for radiation control in boiling water nuclear power reactors

    SciTech Connect (OSTI)

    Rao, T.V.; Vook, R.W.; Meyer, W.; Wittwer, C.

    1987-07-01T23:59:59.000Z

    Stainless-steel surfaces (316 nuclear grade) develop /sup 60/Co-embedded oxide scales when exposed to a boiling water reactor (BWR) environment. Thin films such as Pd, Ni, Au, and Cr were found to drastically reduce the radioactive buildup. The films were prepared by vacuum evaporation and electroless deposition. The electroless coating consisted of a thin cathodically treated layer, followed by a nickel strike (--1 ..mu..m) and an electroless layer (--600 A). The present paper describes the results obtained from a transmission electron microscope replica study of the radioactive growths that formed on uncoated and thin-film-coated stainless-steel rods. The coated rods, when exposed to a simulated BWR environment, exhibited corrosion film growths ranging from large faceted grains (uncoated) and isolated islands of similar crystallites (Au coated) to extremely small nucleated growths (Pd, Ni coated). Also, it was found that chromium oxide films, which generally form a protective oxide on stainless steel, do not completely stop either the corrosion film growth or the associated radioactive buildup. The morphologies of the corrosion film growth were correlated with the relative /sup 60/Co activity, and the substrate topography. The best coating to date was found to be a Pd thin film, 1000 A thick, which reduced the radioactive buildup by a factor of --13.

  12. Recovery Boiler Superheater Ash Corrosion Field Study

    SciTech Connect (OSTI)

    Keiser, James R [ORNL] [ORNL; Kish, Joseph [McMaster University] [McMaster University; Singbeil, Douglas [FPInnovations] [FPInnovations

    2010-01-01T23:59:59.000Z

    With the trend towards increasing the energy efficiency of black liquor recovery boilers operated in North America, there is a need to utilize superheater tubes with increased corrosion resistance that will permit operation at higher temperatures and pressures. In an effort to identify alloys with improved corrosion resistance under more harsh operating conditions, a field exposure was conducted that involved the insertion of an air-cooled probe, containing six candidate alloys, into the superheater section of an operating recovery boiler. A metallographic examination, complete with corrosion scale characterization using EMPA, was conducted after a 1,000 hour exposure period. Based on the results, a ranking of alloys based on corrosion performance was obtained.

  13. Validation and Calibration of Nuclear Thermal Hydraulics Multiscale Multiphysics Models - Subcooled Flow Boiling Study

    SciTech Connect (OSTI)

    Anh Bui; Nam Dinh; Brian Williams

    2013-09-01T23:59:59.000Z

    In addition to validation data plan, development of advanced techniques for calibration and validation of complex multiscale, multiphysics nuclear reactor simulation codes are a main objective of the CASL VUQ plan. Advanced modeling of LWR systems normally involves a range of physico-chemical models describing multiple interacting phenomena, such as thermal hydraulics, reactor physics, coolant chemistry, etc., which occur over a wide range of spatial and temporal scales. To a large extent, the accuracy of (and uncertainty in) overall model predictions is determined by the correctness of various sub-models, which are not conservation-laws based, but empirically derived from measurement data. Such sub-models normally require extensive calibration before the models can be applied to analysis of real reactor problems. This work demonstrates a case study of calibration of a common model of subcooled flow boiling, which is an important multiscale, multiphysics phenomenon in LWR thermal hydraulics. The calibration process is based on a new strategy of model-data integration, in which, all sub-models are simultaneously analyzed and calibrated using multiple sets of data of different types. Specifically, both data on large-scale distributions of void fraction and fluid temperature and data on small-scale physics of wall evaporation were simultaneously used in this work’s calibration. In a departure from traditional (or common-sense) practice of tuning/calibrating complex models, a modern calibration technique based on statistical modeling and Bayesian inference was employed, which allowed simultaneous calibration of multiple sub-models (and related parameters) using different datasets. Quality of data (relevancy, scalability, and uncertainty) could be taken into consideration in the calibration process. This work presents a step forward in the development and realization of the “CIPS Validation Data Plan” at the Consortium for Advanced Simulation of LWRs to enable quantitative assessment of the CASL modeling of Crud-Induced Power Shift (CIPS) phenomenon, in particular, and the CASL advanced predictive capabilities, in general. This report is prepared for the Department of Energy’s Consortium for Advanced Simulation of LWRs program’s VUQ Focus Area.

  14. Summary and bibliography of safety-related events at boiling-water nuclear power plants as reported in 1980

    SciTech Connect (OSTI)

    McCormack, K.E.; Gallaher, R.B.

    1982-03-01T23:59:59.000Z

    This document presents a bibliography that contains 100-word abstracts of event reports submitted to the US Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1980. The 1547 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor. Full-size keyword and permuted-title indexes to facilitate location of individual abstracts are provided following the text. Tables that summarize the information contained in the bibliography are also provided. The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item. Similar information is given for the various kinds of instrumentation and systems, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction).

  15. Film boiling on spheres in single- and two-phase flows.

    SciTech Connect (OSTI)

    Liu, C.; Theofanous, T. G.

    2000-08-29T23:59:59.000Z

    Film boiling on spheres in single- and two-phase flows was studied experimentally and theoretically with an emphasis on establishing the film boiling heat transfer closure law, which is useful in the analysis of nuclear reactor core melt accidents. Systematic experimentation of film boiling on spheres in single-phase water flows was carried out to investigate the effects of liquid subcooling (from 0 to 40 C), liquid velocity (from 0 to 2 m/s), sphere superheat (from 200 to 900 C), sphere diameter (from 6 to 19 mm), and sphere material (stainless steel and brass) on film boiling heat transfer. Based on the experimental data a general film boiling heat transfer correlation is developed. Utilizing a two-phase laminar boundary-layer model for the unseparated front film region and a turbulent eddy model for the separated rear region, a theoretical model was developed to predict the film boiling heat transfer in all single-phase regimes. The film boiling from a sphere in two-phase flows was investigated both in upward two-phase flows (with void fraction from 0.2 to 0.65, water velocity from 0.6 to 3.2 m/s, and steam velocity from 3.0 to 9.0 m/s) and in downward two-phase flows (with void fraction from 0.7 to 0.95, water velocity from 1.9 to 6.5 m/s, and steam velocity from 1.1 to 9.0 m/s). The saturated single-phase heat transfer correlation was found to be applicable to the two-phase film boiling data by making use of the actual water velocity (water phase velocity), and an adjustment factor of (1 - {alpha}){sup 1/4} (with a being the void fraction) for downward flow case only. Slight adjustments of the Reynolds number exponents in the correlation provided an even better interpretation of the two-phase data. Preliminary experiments were also conducted to address the influences of multi-sphere structure on the film boiling heat transfer in single- and two-phase flows.

  16. Superheater Corrosion In Biomass Boilers: Today's Science and Technology

    SciTech Connect (OSTI)

    Sharp, William (Sandy) [SharpConsultant

    2011-12-01T23:59:59.000Z

    This report broadens a previous review of published literature on corrosion of recovery boiler superheater tube materials to consider the performance of candidate materials at temperatures near the deposit melting temperature in advanced boilers firing coal, wood-based fuels, and waste materials as well as in gas turbine environments. Discussions of corrosion mechanisms focus on the reactions in fly ash deposits and combustion gases that can give corrosive materials access to the surface of a superheater tube. Setting the steam temperature of a biomass boiler is a compromise between wasting fuel energy, risking pluggage that will shut the unit down, and creating conditions that will cause rapid corrosion on the superheater tubes and replacement expenses. The most important corrosive species in biomass superheater corrosion are chlorine compounds and the most corrosion resistant alloys are typically FeCrNi alloys containing 20-28% Cr. Although most of these materials contain many other additional additions, there is no coherent theory of the alloying required to resist the combination of high temperature salt deposits and flue gases that are found in biomass boiler superheaters that may cause degradation of superheater tubes. After depletion of chromium by chromate formation or chromic acid volatilization exceeds a critical amount, the protective scale gives way to a thick layer of Fe{sub 2}O{sub 3} over an unprotective (FeCrNi){sub 3}O{sub 4} spinel. This oxide is not protective and can be penetrated by chlorine species that cause further acceleration of the corrosion rate by a mechanism called active oxidation. Active oxidation, cited as the cause of most biomass superheater corrosion under chloride ash deposits, does not occur in the absence of these alkali salts when the chloride is present as HCl gas. Although a deposit is more corrosive at temperatures where it is molten than at temperatures where it is frozen, increasing superheater tube temperatures through the measured first melting point of fly ash deposits does not necessarily produce a step increase in corrosion rate. Corrosion rate typically accelerates at temperatures below the first melting temperature and mixed deposits may have a broad melting temperature range. Although the environment at a superheater tube surface is initially that of the ash deposits, this chemistry typically changes as the deposits mature. The corrosion rate is controlled by the environment and temperature at the tube surface, which can only be measured indirectly. Some results are counter-intuitive. Two boiler manufacturers and a consortium have developed models to predict fouling and corrosion in biomass boilers in order to specify tube materials for particular operating conditions. It would be very useful to compare the predictions of these models regarding corrosion rates and recommended alloys in the boiler environments where field tests will be performed in the current program. Manufacturers of biomass boilers have concluded that it is more cost-effective to restrict steam temperatures, to co-fire biofuels with high sulfur fuels and/or to use fuel additives rather than try to increase fuel efficiency by operating with superheater tube temperatures above melting temperature of fly ash deposits. Similar strategies have been developed for coal fired and waste-fired boilers. Additives are primarily used to replace alkali metal chloride deposits with higher melting temperature and less corrosive alkali metal sulfate or alkali aluminum silicate deposits. Design modifications that have been shown to control superheater corrosion include adding a radiant pass (empty chamber) between the furnace and the superheater, installing cool tubes immediately upstream of the superheater to trap high chloride deposits, designing superheater banks for quick replacement, using an external superheater that burns a less corrosive biomass fuel, moving circulating fluidized bed (CFB) superheaters from the convective pass into the hot recirculated fluidizing medium and adding an insulating layer to superh

  17. Integrated boiler, superheater, and decomposer for sulfuric acid decomposition

    DOE Patents [OSTI]

    Moore, Robert (Edgewood, NM); Pickard, Paul S. (Albuquerque, NM); Parma, Jr., Edward J. (Albuquerque, NM); Vernon, Milton E. (Albuquerque, NM); Gelbard, Fred (Albuquerque, NM); Lenard, Roger X. (Edgewood, NM)

    2010-01-12T23:59:59.000Z

    A method and apparatus, constructed of ceramics and other corrosion resistant materials, for decomposing sulfuric acid into sulfur dioxide, oxygen and water using an integrated boiler, superheater, and decomposer unit comprising a bayonet-type, dual-tube, counter-flow heat exchanger with a catalytic insert and a central baffle to increase recuperation efficiency.

  18. Superheat effects on localized vessel breach enlargement during corium ejection

    SciTech Connect (OSTI)

    Sienicki, J.J.; Spencer, B.W.

    1986-01-01T23:59:59.000Z

    The evaluation of the consequences of hypothetical severe accident sequences in light water reactors includes those sequences in which molten corium is postulated to melt through the reactor pressure vessel (RPV) lower head and enter the region beneath the RPV. An important issue is the mode by which the lower head is breached and molten corium introduced into the reactor cavity (PWR) or pedestal (BWR). Reported here are the results of an investigation into the dependency of ablation-induced enlargement on the initial corium temperature, or more specifically, the initial corium superheat (i.e., excess temperature above the freezing temperature). A model is introduced here to predict the vessel erosion and is employed to scope the effects of variations in the superheat.

  19. On Reducing Evaporator Superheat Nonlinearity with Control Architecture

    E-Print Network [OSTI]

    Elliot, M.S.; Rasmussen, B.P.

    , and refrigeration VCC vapor compression cycle TEV thermostatic expansion valve EEV electronic expansion valve HEV hybrid expansion valve PID proportional-integral-derivative P, Pevap evaporator pressure Q(s) transfer function from PSET to evaporator superheat (inner... by a number of different valve types, which vary in expense and design KFM product of KF times KM(v) KU gain from HEV position to mechanical pressure setpoint U HEV position MEMS micro-electrical-mechanical systemsregulating device also known...

  20. Void volumes in subcooled boiling systems

    E-Print Network [OSTI]

    Griffith, P.

    1958-01-01T23:59:59.000Z

    Introduction: Knowledge of the pressure drop in a channel and the resulting flow redistribution is essential in predicting the performance of a nuclear reactor. The pressure drop in a channel which is experiencing boiling ...

  1. EA-1394: Final Environmental Assessment

    Broader source: Energy.gov [DOE]

    Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear Superheat (BONUS) Reactor Building, Ricon, Puerto Rico

  2. MOLTEN SALT CORROSION OF SUPERHEATERS IN BLACK LIQUOR RECOVERY BOILERS John Bohling, University of Tennessee Georgia Tech SURF 2010 Fellow

    E-Print Network [OSTI]

    Li, Mo

    MOLTEN SALT CORROSION OF SUPERHEATERS IN BLACK LIQUOR RECOVERY BOILERS John Bohling, University If the temperature is above the first melting point of the scale, severe corrosion of the tubes can result temperatures, reducing efficiency. The corrosive nature of the superheater environment arises primarily from

  3. Boiling behavior of sodium-potassium alloy in a bench-scale solar receiver

    SciTech Connect (OSTI)

    Moreno, J.B.; Andraka, C.E.; Moss, T.A.

    1992-01-01T23:59:59.000Z

    During 1989-90, a 75-kW{sub t} sodium reflux pool-boiler solar receiver was successfully demonstrated at Sandia National Laboratories. Significant features of this receiver include (1) boiling sodium as the heat transfer medium and (2) electric-discharge-machined (EDM) cavities as artificial nucleation sites to stabilize boiling. Since this first demonstration, design of a second-generation pool-boiler receiver that will bring the concept closer to commercialization has begun. For long life, the new receiver uses Haynes Alloy 230. For increased safety factors against film boiling and flooding, it has a refined shape and somewhat larger dimensions. To eliminate the need for trace heating, the receiver will boil the sodium-potassium alloy NaK-78 instead of sodium. To reduce manufacturing costs, it will use one of a number of alternatives to EDM cavities for stabilization of boiling. To control incipient-boiling superheats, especially during hot restarts, it will contain a small amount of inert gas. Before the new receiver design could be finalized, bench-scale tests of some of the proposed changes were necessary. A series of bench-scale pool boilers were built from Haynes Alloy 230 and filled with NaK-78. Various boiling-stabilizer candidates were incorporated into them, including laser-drilled cavities and a number of different sintered-powder-metal coatings. These bench-scale pool boilers have been operated at temperatures up to 750{degree}C, heated by quartz lamps with incident radiant fluxes up to 95 W/cm{sup 2}. The effects of various orientations and added gases have been studied. results of these studies are presented. 15 refs.

  4. MANAGING OXIDE SCALE EXFOLIATION IN BOILERS WITH TP347H SUPERHEATER TUBES

    SciTech Connect (OSTI)

    Sabau, Adrian S [ORNL] [ORNL; Wright, Ian G. [WrightHT, Inc.] [WrightHT, Inc.; Shingledecker, John P. [Electric Power Research Institute (EPRI)] [Electric Power Research Institute (EPRI); Tortorelli, Peter F [ORNL] [ORNL

    2014-01-01T23:59:59.000Z

    A model based on a concept of fraction of exfoliated area as a function of oxide scale strain energy was developed to predict the extent of exfoliation of steam-side scale from boiler tube superheater loops. As compared with the Armitt diagram, which can be used to predict when scale damage and exfoliation would be likely to occur, a fraction of exfoliated area approach provides an estimation of mass of scale released and the fraction of tube likely to be blocked by the exfoliation. This paper show results for the extent of blockage expected in a single bend of a superheater loop was predicted as a function of operating time, bend geometry, and outlet steam temperature under realistic service conditions that include outages. The deposits of exfoliated scale were assumed to be distributed horizontally the tubes bends. Three types of bends were considered: regular bends, short bends, and hairpin bends. The progressive increase in steam and tube temperatures along a single loop of superheater tubing and the ensuing variation of oxide scale thickness are considered. Numerical simulation results for a superheater loop made of TP347H austenitic steel indicated that tube blockage fractions larger than 50% are likely to occur within the first two years of boiler operation (with regularly scheduled outages) for outlet tube temperatures of 540-570oC, which is consistent with practical experience. Higher blockage fractions were predicted for tubes with short bends and hairpin bends than for tubes with regular bends, of length that are larger than five internal tube diameters. Finally, the blockage model presented can be used with some confidence to devise operating schedules for managing the consequences of oxide scale exfoliation based on projections of time to some critical blockage fraction for specific boiler operating conditions.

  5. OXIDE SCALE EXFOLIATION AND REGROWTH IN TP347H SUPERHEATER TUBES

    SciTech Connect (OSTI)

    Sabau, Adrian S [ORNL; Wright, Ian G [ORNL; Shingledecker, John P. [Electric Power Research Institute (EPRI)

    2012-01-01T23:59:59.000Z

    This paper provides an introduction to a comprehensive model being developed to predict and control oxide scale exfoliation from the steam-side of superheater and reheater tubes in steam boilers. The model deals with the main phenomena involved in scale growth and failure in steam, and incorporates major variables related to boiler design and operation. The considerations used to calculate oxide growth under the specific constrains of small diameter tubes carrying high-pressure steam and operating with large temperature gradients under temperature and pressure cycling conditions, as well as the evolution of stresses and strains in the scales, are indicated but only a cursory description is given of the details of the analytical treatments. An example is presented of calculations made with the model to predict the extent of blockage expected in a single superheater loop as a function of time and outlet steam temperature under several realistic service conditions. The results suggest that problems due to scale exfoliation would be expected early in the operating life of superheater tubes made from austenitic steel TP347H.

  6. A study of electrowetting-assisted boiling

    E-Print Network [OSTI]

    Bralower, Harrison L. (Harrison Louis)

    2011-01-01T23:59:59.000Z

    The classical theory of boiling heat transfer based on bubble dynamics is explained and includes a full derivation of the Rohsenow boiling correlation. An alternative, more accurate correlation for determining boiling heat ...

  7. Fatigue Testing of Metallurgically-Bonded EBR-II Superheater Tubes

    SciTech Connect (OSTI)

    Terry C. Totemeier

    2006-12-01T23:59:59.000Z

    Fatigue crack growth tests were performed on 2¼Cr-1Mo steel specimens machined from ex-service Experimental Breeder Reactor – II (EBR-II) superheater duplex tubes. The tubes had been metallurgically bonded with a 100 µm thick Ni interlayer; the specimens incorporated this bond layer. Tests were performed at room temperature in air and at 400°C in air and humid Ar; cracks were grown at varied levels of constant ?K. Crack growth tests at a range of ?K were also performed on specimens machined from the shell of the superheater. In all conditions the presence of the Ni interlayer was found to result in a net retardation of growth as the crack passed through the interlayer. The mechanism of retardation was identified as a disruption of crack planarity and uniformity after passing through the porous interlayer. Full crack arrest was only observed in a single test performed at near-threshold ?K level (12 MPa?m) at 400°C. In this case the crack tip was blunted by oxidation of the base steel at the steel-interlayer interface.

  8. Boiling water reactor control rod

    SciTech Connect (OSTI)

    Wilson, J.F.; Doshi, P.K.

    1986-12-23T23:59:59.000Z

    This patent describes a control rod for a boiling water type nuclear power reactor, the improvement comprising: (a) an elongated central stem defining a longitudinally extending internal central gas plenum; (b) blades connected to and extending along and radially outward from the stem, each blade including an elongated body portion extending along the stem and terminating in an end tip portion; (c) means defining a series of internal cavities in each of the blades, the cavities being arranged in columns and rows across the length and width of the body and tip portions of the blade; (d) pellets of neutron absorbing material, each disposed within one of the cavities with each of the cavities being oversized in relation to the size of the pellet disposed therein to allow extra space for swelling of the pellet. The cavities and the pellets disposed therein are arranged to define a longer, constant worth section generally coextensive with the body portion of the blade and a shorter, reduced worth section generally coextensive with the end tip portion of the blade; and (e) means defined within each blade communicating each of the cavities with the central gas plenum for allowing any gases generated by irradiation of the pellets to expand from the cavities into the plenum.

  9. Subcooled flow boiling of fluorocarbons

    E-Print Network [OSTI]

    Murphy, Richard Walter

    1971-01-01T23:59:59.000Z

    A study was conducted of heat transfer and hydrodynamic behavior for subcooled flow boiling of Freon-113, one of a group of fluorocarbons suitable for use in cooling of high-power-density electronic components. Problems ...

  10. Formation of deposits on the surfaces of superheaters and economisers of MSW incinerator plants

    SciTech Connect (OSTI)

    Reichelt, J. [IBR, Obergrombacher Strasse 29, D-76646 Bruchsal (Germany); Pfrang-Stotz, G., E-mail: Gudrun.Pfrang-Stotz@kit.edu [Karlsruhe Institute of Technology (KIT), ITC, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Bergfeldt, B.; Seifert, H. [Karlsruhe Institute of Technology (KIT), ITC, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Knapp, P. [MVV O and M GmbH, Muellheizkraftwerk Mannheim, Otto-Hahn-Strasse 1, D-68169 Mannheim (Germany)

    2013-01-15T23:59:59.000Z

    Highlights: Black-Right-Pointing-Pointer Composition of deposits depends on the temperature profile and boiler geometry. Black-Right-Pointing-Pointer The mineralogy of deposits defines critical and uncritical zones in the boiler. Black-Right-Pointing-Pointer Critical zones in boilers can be characterised by a classification systems. Black-Right-Pointing-Pointer Specific measures to enhance energy efficiency can be defined. - Abstract: Mineralogical and chemical investigations of deposits from superheaters and economisers from a MSWI plant in Mannheim, Germany, lead to a classification system which provides information about the most critical parameters leading to fouling and corrosion. With the help of this classification system parameters like the geometry of boilers and the waste input can be changed in order to prolong run times between revisions and enhance energy efficiency of MSWI plants.

  11. Dispersed flow film boiling

    E-Print Network [OSTI]

    Yoder, Graydon L.

    1980-01-01T23:59:59.000Z

    Dispersed flow consists of small liquid droplets entrained in a flowing vapor. This flow regime can occur in cryogenic equipment, in steam generators, and during nuclear reactor loss of coolant accidents. A theoretical ...

  12. Pool boiling on nano-finned surfaces

    E-Print Network [OSTI]

    Sriraman, Sharan Ram

    2009-05-15T23:59:59.000Z

    The effect of nano-structured surfaces on pool boiling heat transfer is explored in this study. Experiments are conducted in a cubical test chamber containing fluoroinert coolant (PF5060, Manufacturer: 3M Co.) as the working fluid. Pool boiling...

  13. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings

    SciTech Connect (OSTI)

    Van Weele, S. (Foster Wheeler Development Corp., Livingston, NJ (United States))

    1991-08-01T23:59:59.000Z

    Fireside corrosion, caused by liquid alkali-iron trisulfates, has been an obstacle to higher steam temperatures and to efficient utilization of high-sulfur coals. Tests simulating the environment in the superheater bank of a pulverized-coal-fired boiler were conducted on several promising new alloys and claddings. Alloys were exposed to a variety of synthetic ash and simulated flue gas compositions at 650 and 700{degrees}C for times ranging up to 800 hours. Included in the testing program were new high-chromium/high-nickel alloys, modified commercial alloys, lean stainless steels (modified Type 316) clad with high-chromium/high-nickel alloys, and intermetallic aluminides. Thickness loss measurements indicated that resistance to attach improved with increasing chromium level. Silicon and aluminum were also helpful in resisting attack, while molybdenum was detrimental to the resistance of the alloys to attack. Three different attack modes were observed on the alloys tested. Alloys with low resistance to attack exhibited uniform wastage, while pitting was observed in more resistant alloys. In addition to surface fluxing by molten alkali-iron trisulfates, subsurface sulfur penetration and intergranular attack also occurred.

  14. Advanced nuclear reactor safety analysis: the simulation of a small break loss of coolant accident in the simplified boiling water reactor using RELAP5/MOD3.1.1 

    E-Print Network [OSTI]

    Faust, Christophor Randall

    1995-01-01T23:59:59.000Z

    The thermal hydraulic simulation code RELAP5/MOD3.1.1 was utilized to model General Electric's Simplified Boiling Water Reactor plant. The model of the plant was subjected to a small break loss of coolant accident occurring from a guillotine shear...

  15. Efficient Cooling in Engines with Nucleate Boiling

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    with Nucleate Boiling Principal Investigator: Wenhua Yu Coworkers: D. France and R. Smith Energy Systems Division Argonne National Laboratory OVT Merit Review February 28, 2008...

  16. Enhancement of Pool Boiling Heat Transfer in Confined Space

    E-Print Network [OSTI]

    Hsu, Chia-Hsiang

    2014-05-05T23:59:59.000Z

    Pool boiling is an effective method used in many technical applications for a long time. Its highly efficient heat transfer performance results from not only the convection effect but also the phase change process in pool boiling. Pool boiling...

  17. Boiling of the Interface between Two Immiscible Liquids below the Bulk Boiling Temperatures of Both Components

    E-Print Network [OSTI]

    Anastasiya V. Pimenova; Denis S. Goldobin

    2014-10-20T23:59:59.000Z

    We consider the problem of boiling of the direct contact of two immiscible liquids. An intense vapour formation at such a direct contact is possible below the bulk boiling points of both components, meaning an effective decrease of the boiling temperature of the system. Although the phenomenon is known in science and widely employed in technology, the direct contact boiling process was thoroughly studied (both experimentally and theoretically) only for the case where one of liquids is becoming heated above its bulk boiling point. On the contrary, we address the case where both liquids remain below their bulk boiling points. In this paper we construct the theoretical description of the boiling process and discuss the actualisation of the case we consider for real systems.

  18. PACCAR CRADA: Experimental Investigation in Coolant Boiling in...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    PACCAR CRADA: Experimental Investigation in Coolant Boiling in a Half-Heated Circular Tube PACCAR CRADA: Experimental Investigation in Coolant Boiling in a Half-Heated Circular...

  19. CRADA with PACCAR Experimental Investigation in Coolant Boiling...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    PACCAR Experimental Investigation in Coolant Boiling in a Half-Heated Circular Tube CRADA with PACCAR Experimental Investigation in Coolant Boiling in a Half-Heated Circular Tube...

  20. Enhancement of Heat Transfer with Pool and Spray Impingement Boiling on Microporous and Nanowire Surface Coatings

    SciTech Connect (OSTI)

    Thiagarajan, S. J.; Wang, W.; Yang, R.; Narumanchi, S.; King, C.

    2010-09-01T23:59:59.000Z

    The DOE National Renewable Energy Laboratory (NREL) is leading a national effort to develop next-generation cooling technologies for hybrid vehicle electronics. The goal is to reduce the size, weight, and cost of power electronic modules that convert direct current from batteries to alternating current for the motor, and vice versa. Aggressive thermal management techniques help to increase power density and reduce weight and volume, while keeping chip temperatures within acceptable limits. The viability of aggressive cooling schemes such as spray and jet impingement in conjunction with enhanced surfaces is being explored. Here, we present results from a series of experiments with pool and spray boiling on enhanced surfaces, such as a microporous layer of copper and copper nanowires, using HFE-7100 as the working fluid. Spray impingement on the microporous coated surface showed an enhancement of 100%-300% in the heat transfer coefficient at a given wall superheat with respect to spray impingement on a plain surface under similar operating conditions. Critical heat flux also increased by 7%-20%, depending on flow rates.

  1. Acoustically Enhanced Boiling Heat Transfer

    E-Print Network [OSTI]

    Z. W. Douglas; M. K. Smith; A. Glezer

    2008-01-07T23:59:59.000Z

    An acoustic field is used to increase the critical heat flux (CHF) of a flat-boiling-heat-transfer surface. The increase is a result of the acoustic effects on the vapor bubbles. Experiments are performed to explore the effects of an acoustic field on vapor bubbles in the vicinity of a rigid-heated wall. Work includes the construction of a novel heater used to produce a single vapor bubble of a prescribed size and at a prescribed location on a flatboiling surface for better study of an individual vapor bubble's reaction to the acoustic field. Work also includes application of the results from the single-bubble heater to a calibrated-copper heater used for quantifying the improvements in CHF.

  2. Nucleate boiling bubble growth and departure

    E-Print Network [OSTI]

    Staniszewski, Bogumil E.

    1959-01-01T23:59:59.000Z

    The vapor bubble formation on the heating surface during pool boiling has been studied experimentally. Experiments were made at the atmospheric pressure 28 psi and 40 psi, using degassed distilled water and ethanol. The ...

  3. Pool boiling heat transfer characteristics of nanofluids

    E-Print Network [OSTI]

    Kim, Sung Joong, Ph. D. Massachusetts Institute of Technology

    2007-01-01T23:59:59.000Z

    Nanofluids are engineered colloidal suspensions of nanoparticles in water, and exhibit a very significant enhancement (up to 200%) of the boiling Critical Heat Flux (CHF) at modest nanoparticle concentrations (50.1% by ...

  4. Pool boiling on nano-finned surfaces

    E-Print Network [OSTI]

    Sriraman, Sharan Ram

    2008-10-10T23:59:59.000Z

    is explored in this study. Experiments are conducted in a cubical test chamber containing fluoroinert coolant (PF5060, Manufacturer: 3M Co.) as the working fluid. Pool boiling experiments are conducted for saturation and subcooled conditions. Three...). .................................................................................................68 Fig. 19. Plots comparing the heat flux through the enhanced test surfaces (qw”) with the heat flux through the bare test surface (qb”) under 10 °C sub-cooling condition for both nulceate and film boiling regimes. Heat flux data...

  5. Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

    SciTech Connect (OSTI)

    Huang, Jun-Lin; Zhou, Ke-Yi, E-mail: boiler@seu.edu.cn; Xu, Jian-Qun [Key Laboratory of Energy Thermal Conversion and Control of Ministry of Education, School of Energy and Environment, Southeast University, Nanjing 210096, Jiangsu Province (China); Wang, Xin-Meng; Tu, Yi-You [School of Materials Science and Engineering, Southeast University, Nanjing 210096, Jiangsu Province (China)

    2014-07-28T23:59:59.000Z

    Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

  6. Hydrodynamics and heat transfer during flow boiling instabilities in a single microchannel

    E-Print Network [OSTI]

    Aussillous, Pascale

    Hydrodynamics and heat transfer during flow boiling instabilities in a single microchannel July 2008 Keywords: Boiling Microchannels Visualisation Flow boiling instabilities Heat transfer a b intensification heat removal. Flow boiling heat transfer in microchannel geometry and the associated flow

  7. Transition boiling heat transfer from a horizontal surface

    E-Print Network [OSTI]

    Berenson Paul Jerome

    1960-01-01T23:59:59.000Z

    An experiment, utilising a condensing fluid as the heat source, was performed to determine the heat flux vs. temperature difference curve for transition pool boiling from a horisontal surface. The boiling cure was determined ...

  8. Boiling and condensation in a liquid-filled enclosure

    E-Print Network [OSTI]

    Bar-Cohen Avram

    1971-01-01T23:59:59.000Z

    A combined experimental and analytical investigation of boiling and condensation in a liquid-filled enclosure, with water and Freon- 113 as the working fluids, is described. The operating characteristics of a boiling system, ...

  9. PNNL Enhanced Pool-Boiling Heat Transfer Using Nanostructured Surfaces

    ScienceCinema (OSTI)

    None

    2012-12-31T23:59:59.000Z

    Close-up video of boiling taking place on a nanostructured surface in a controlled laboratory experiment.

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

    SciTech Connect (OSTI)

    Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-03-01T23:59:59.000Z

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

  11. Boiling heat transfer with three fluids in small circular and rectangular channels

    SciTech Connect (OSTI)

    Tran, T.N.; Wambsganss, M.W. [Argonne National Lab., IL (United States); France, D.M. [Illinois Univ., Chicago, IL (United States). Dept. of Mechanical Engineering

    1995-01-01T23:59:59.000Z

    Small circular and noncircular channels are representative of flow passages act evaporators and condensers. This report describes results of an ental study on heat transfer to the flow boiling of refrigerants (R-12) and refrigerant-134a (R-134a) in a small horizontal circular-cross-section tube. The tube diameter of 2.46 mm was chosen to approximate the hydraulic diameter of a 4.06 {times} 1.70 mm rectangular channel previously studied with R-12, and a 2.92-mm-diameter circular tube previously studied with R-113. The objective of this study was to assess the effects of channel geometry and fluid properties on the heat transfer coefficient and to obtain additional insights relative to the heat transfer mechanism(s). The current circular flow channel for the R-12 and R-134a tests was made of brass and had an overall length of 0.9 in. The channel wall was electrically heated, and thermocouples were installed on the channel wall and in the bulk fluid stream. Voltage taps were located at the same axial locations as the stream thermocouples to allow testing over an exit quality range to 0.94 and a large range of mass flux (58 to 832 kg/m{sup 2}s) and heat flux (3.6 to 59 kW/m{sup 2}). Saturation pressure was nearly constant, averaging 0.82 MPa for most of the testing, with some tests performed at a lower pressure of 0.4--0.5 MPa. Local heat transfer coefficients were determined experimentally as a function of quality along the length of the test section. Analysis of all data for three tubes and three fluids supported the conclusion that a nucleation mechanism dominates for flow boiling in small channels. Nevertheless, a convection-dominant region was obtained experimentally in this study at very low values of wall superheat (< {approx} 2.75{degrees}C). The circular and rectangular tube data for three fluids were successfully correlated in the nucleation-dominant region.

  12. Mechanism of nucleate pool boiling heat transfer to sodium and the criterion for stable boiling

    E-Print Network [OSTI]

    Shai, Isaac

    1967-01-01T23:59:59.000Z

    A comparison between liquid metals and other common fluids, like water, is made as regards to the various stages of nucleate pool boiling. It is suggested that for liquid metals the stage of building the thermal layer plays ...

  13. Transition from film boiling to nucleate boiling in forced convection vertical flow

    E-Print Network [OSTI]

    Iloeje, Onwuamaeze C.

    1972-01-01T23:59:59.000Z

    The mechanism of collapse of forced cnnvection annular vertical flow film boiling, with liquid core, is investigated using liquid nitrogen at low pressures. The report includes the effect of heat flux from the buss bar. ...

  14. Boiling heat transfer in a vertical microchannel: Local estimation during flow boiling with a non intrusive method

    E-Print Network [OSTI]

    Boiling heat transfer in a vertical microchannel: Local estimation during flow boiling with a non the results of experimental and numerical studies concerning boiling heat transfer inside vertical in minichannels for several gravity levels (µg, 1g, 2g). To fully understand the high heat transfer potential

  15. POOL BOILING OF HIGH-FREQUENCY CONDUCTORS

    SciTech Connect (OSTI)

    Wright, S. E. (Spencer E.); Konecni, S. (Snezana); Ammerman, C. N. (Curtt N.); Sims, J. R. (James R.)

    2001-01-01T23:59:59.000Z

    This study presents development of a unique, powerful method for cooling high-frequency, AC conductors that can benefit end users of transformer windings, electrical machine windings, and magnet coils. This method of heat removal involves boiling a dielectric, fluorinert refrigerant that is in direct contact with litz wire conductors. A pool boiling test vessel is constructed, which provides for temperature control of the pool of fluorinert liquid. The test vessel is fitted with viewing ports so that the experiments are observed and studied with the aid of high-speed photography. Tests are performed on a variety of litz wire conductors. The boiling heat transfer coefficient is dependent on the conductor surface roughness. The size of the features on the conductor surface depends on the single-strand wire gage from which the conductor is constructed. All tests are performed with the conductors mounted horizontally. These tests are performed using a DC power supply. The results of these experiments will aid in the design of future cooling systems.

  16. Flow boiling test of GDP replacement coolants

    SciTech Connect (OSTI)

    Park, S.H. [comp.

    1995-08-01T23:59:59.000Z

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C{sub 4}F{sub 10} and C{sub 4}F{sub 8}, were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C{sub 4}F{sub 10} mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C{sub 4}F{sub 10} weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd.

  17. Nucleate boiling pressure drop in an annulus: Book 8

    SciTech Connect (OSTI)

    Not Available

    1992-11-01T23:59:59.000Z

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of tables of temperature measurements.

  18. Nucleate boiling pressure drop in an annulus: Book 6

    SciTech Connect (OSTI)

    Not Available

    1992-11-01T23:59:59.000Z

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of a summary of temperature measurements to include recorded minima, maxima, averages and standard deviations.

  19. Nucleate boiling pressure drop in an annulus: Book 7

    SciTech Connect (OSTI)

    Not Available

    1992-11-01T23:59:59.000Z

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists solely of tables of temperature measurements; minima, maxima, averages and standard deviations being measured.

  20. Nucleate boiling pressure drop in an annulus: Book 4

    SciTech Connect (OSTI)

    Block, J.A.; Crowley, C.; Dolan, F.X.; Sam, R.G.; Stoedefalke, B.H.

    1992-11-01T23:59:59.000Z

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of data plots and summary files of temperature measurements.

  1. Nucleate boiling pressure drop in an annulus: Book 3

    SciTech Connect (OSTI)

    Block, J.A.; Crowley, C.; Dolan, F.X.; Sam, R.G.; Stoedefalke, B.H.

    1992-11-01T23:59:59.000Z

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of data plots and summary files of temperature measurements.

  2. Nucleate boiling pressure drop in an annulus: Book 2

    SciTech Connect (OSTI)

    Block, J.A.; Crowley, C.; Dolan, F.X.; Sam, R.G.; Stoedefalke, B.H.

    1992-11-01T23:59:59.000Z

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. Nineteen test series and a total of 178 tests were performed. Testing addressed the effects of: Heat flux; pressure; helium gas; power tilt; ribs; asymmetric heat flux.

  3. actual coolant boiling: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  4. abwr advanced boiling: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  5. advanced boiling curve: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  6. automatic boiling water: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  7. advanced boiling water: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  8. Cooling Boiling in Head Region - PACCAR Integrated Underhood...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Integrated External Aerodynamic and Underhood Thermal Analysis for Heavy Vehicles CRADA with PACCAR Experimental Investigation in Coolant Boiling in a Half-Heated Circular...

  9. Conversion of direct process high-boiling residue to monosilanes

    DOE Patents [OSTI]

    Brinson, Jonathan Ashley (Vale of Glamorgan, GB); Crum, Bruce Robert (Madison, IN); Jarvis, Jr., Robert Frank (Midland, MI)

    2000-01-01T23:59:59.000Z

    A process for the production of monosilanes from the high-boiling residue resulting from the reaction of hydrogen chloride with silicon metalloid in a process typically referred to as the "direct process." The process comprises contacting a high-boiling residue resulting from the reaction of hydrogen chloride and silicon metalloid, with hydrogen gas in the presence of a catalytic amount of aluminum trichloride effective in promoting conversion of the high-boiling residue to monosilanes. The present process results in conversion of the high-boiling residue to monosilanes. At least a portion of the aluminum trichloride catalyst required for conduct of the process may be formed in situ during conduct of the direct process and isolation of the high-boiling residue.

  10. Multi-cycle boiling water reactor fuel cycle optimization

    SciTech Connect (OSTI)

    Ottinger, K.; Maldonado, G.I. [University of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States)

    2013-07-01T23:59:59.000Z

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

  11. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    SciTech Connect (OSTI)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20T23:59:59.000Z

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm lithium metaborate solution respectively at the saturation temperature for 1000 psi (68.9 bar) coolant pressure. Boiling tests also revealed the formation of fine deposits of boron and lithium on the cladding surface which degraded the heat transfer rates. The boron and lithium metaborate precipitates after a 5 day test at 5000 ppm concentration and 1000 psi (68.9 bar) operating pressure reduced the heat transfer rate 21% and 30%, respectively for the two solutions.

  12. A study of boiling water flow regimes at low pressures

    E-Print Network [OSTI]

    Fiori, Mario P.

    1966-01-01T23:59:59.000Z

    "A comprehensive experimental program to examine flow regimes at pressures below 100 psia for boiling of water in tubes was carried out. An electrical probe, which measures the resistance of the fluid between the centerline ...

  13. Film Boiling on Downward Quenching Hemisphere of Varying Sizes

    SciTech Connect (OSTI)

    Chan S. Kim; Kune Y. Suh; Joy L. Rempe; Fan-Bill Cheung; Sang B. Kim

    2004-04-01T23:59:59.000Z

    Film boiling heat transfer coefficients for a downward-facing hemispherical surface are measured from the quenching tests in DELTA (Downward-boiling Experimental Laminar Transition Apparatus). Two test sections are made of copper to maintain low Biot numbers. The outer diameters of the hemispheres are 120 mm and 294 mm, respectively. The thickness of all the test sections is 30 mm. The effect of diameter on film boiling heat transfer is quantified utilizing results obtained from the test sections. The measured data are compared with the numerical predictions from laminar film boiling analysis. The measured heat transfer coefficients are found to be greater than those predicted by the conventional laminar flow theory on account of the interfacial wavy motion incurred by the Helmholtz instability. Incorporation of the wavy motion model considerably improves the agreement between the experimental and numerical results in terms of heat transfer coefficient. In addition, the interfacial wavy motion and the quenching process are visualized through a digital camera.

  14. Pressure drop with surface boiling in small-diameter tubes

    E-Print Network [OSTI]

    Dr?mer, Thomas

    1964-01-01T23:59:59.000Z

    Pressure drop for water flowing in small-diameter tubes under isothermal, nonboiling, and surface-boiling conditions was investigated. Experimental results for local pressure gradient and heattransfer coefficients are ...

  15. Suppression and Enhancement of Boiling Associated with Multiple Droplet Impingement

    E-Print Network [OSTI]

    Yang, Yuxuan

    2014-05-01T23:59:59.000Z

    are at play simultaneously. Furthermore, experiments with single streams of droplets have not been able to elucidate the effects of the onset of boiling (ONB) during the droplet impingement process. Therefore, efforts have been undertaken to consider...

  16. Pool boiling studies on nanotextured surfaces under highly subcooled conditions

    E-Print Network [OSTI]

    Sathyamurthi, Vijaykumar

    2009-05-15T23:59:59.000Z

    Subcooled pool boiling on nanotextured surfaces is explored in this study. The experiments are performed in an enclosed viewing chamber. Two silicon wafers are coated with Multiwalled Carbon Nanotubes (MWCNT), 9 microns (Type-A) and 25 microns (Type...

  17. Thermal boundary layer development in dispersed flow film boiling

    E-Print Network [OSTI]

    Hull, Lawrence M.

    1982-01-01T23:59:59.000Z

    Dispersed flow film boiling consists of a dispersion of droplets which are carried over a very hot surface by their vapor. This process occurs in cryogenic equipment and wet steam turbines. It is also of interest in the ...

  18. Dryout droplet distribution and dispersed flow film boiling

    E-Print Network [OSTI]

    Hill, Wayne S.

    1982-01-01T23:59:59.000Z

    Dispersed flow film boiling is characterized by liquid-phase droplets entrained in a continuous vapor-phase flow. In a previous work at MIT, a model of dispersed flow heat transfer was developed, called the Local Conditions ...

  19. The effects of surface conditions on boiling characteristics

    E-Print Network [OSTI]

    Lorenz, J. J.

    1972-01-01T23:59:59.000Z

    A unified model relating surface variables to the nucleate pool boiling characteristics was developed. A simple vapor trapping mechanism was postulated and a geometrical model constructed for idealized conical cavities ...

  20. Forced-convection, dispersed-flow film boiling

    E-Print Network [OSTI]

    Hynek, Scott Josef

    1969-01-01T23:59:59.000Z

    This report presents the latest results of an investigation of the characteristics of dispersed flow film boiling. Heat transfer data are presented for vertical upflow of nitrogen in an electrically heated tube, 0.4 in. ...

  1. Hydrodynamics, heat transfer and flow boiling instabilities in microchannels 

    E-Print Network [OSTI]

    Barber, Jacqueline Claire

    2010-01-01T23:59:59.000Z

    Boiling in microchannels is a very efficient mode of heat transfer with high heat and mass transfer coefficients achieved. Less pumping power is required for two-phase flows than for single-phase liquid flows to achieve ...

  2. Infrared thermometry study of nanofluid pool boiling phenomena

    E-Print Network [OSTI]

    Gerardi, Craig

    Abstract Infrared thermometry was used to obtain first-of-a-kind, time- and space-resolved data for pool boiling phenomena in water-based nanofluids with diamond and silica nanoparticles at low concentration (<0.1 vol.%). ...

  3. Determination of pool boiling Critical Heat Flux enhancement in nanofluids

    E-Print Network [OSTI]

    Truong, Bao H. (Bao Hoai)

    2007-01-01T23:59:59.000Z

    Nanofluids are engineered colloids composed of nano-size particles dispersed in common fluids such as water or refrigerants. Using an electrically controlled wire heater, pool boiling Critical Heat Flux (CHF) of Alumina ...

  4. Effects of surface parameters on boiling heat transfer phenomena

    E-Print Network [OSTI]

    Truong, Bao H. (Bao Hoai)

    2011-01-01T23:59:59.000Z

    Nanofluids, engineered colloidal dispersions of nanoparticles in fluid, have been shown to enhance pool and flow boiling CHF. The CHF enhancement was due to nanoparticle deposited on the heater surface, which was verified ...

  5. Boiling heat transfer on superhydrophilic, superhydrophobic, and superbiphilic surfaces

    E-Print Network [OSTI]

    Attinger, Daniel

    Boiling heat transfer on superhydrophilic, superhydrophobic, and superbiphilic surfaces Amy Rachel in revised form 30 October 2012 Accepted 31 October 2012 Keywords: Superhydrophobic Superhydrophilic Biphilic- and nanofabrication, superhydrophilic and superhydrophobic surfaces have been developed. The statics and dynamics

  6. Burnout in forced convection nucleate boiling of water

    E-Print Network [OSTI]

    Reynolds John Mitchell

    1957-01-01T23:59:59.000Z

    Data are presented for burnout in forced coivection nucleate boiling of water at pressures above 500 psia. A dimensionless correlation is devised for. the M.I.T. data which is found to be valid for certain recent data ...

  7. The correlation of nucleate boiling burn-out data

    E-Print Network [OSTI]

    Griffith, P.

    1957-01-01T23:59:59.000Z

    A dimensionless correlation is developed for nucleate boiling buzrnout data including the following ranges of variables. Fluids - Water Bensene n - Heptane n - Pentane Ethanol Pressure - 0.0045 to 0.96 of critical pressure ...

  8. RELAP5/MOD3 subcooled boiling model assessment

    SciTech Connect (OSTI)

    Devkin, A.S.; Podosenov, A.S. [Russian Research Center, Moscow (Russian Federation). Nuclear Safety Inst.

    1998-05-01T23:59:59.000Z

    This report presents the assessment of the RELAP5/Mod3 (5m5 version) code subcooled boiling process model which is based on a variety of experiments. The accuracy of the model is confirmed for a wide range of regime parameters for the case of uniform heating along the channel. The condensation rate is rather underpredicted, which may lead to considerable errors in void fraction behavior prediction in subcooled boiling regimes for nonuniformly or unheated channels.

  9. Enhanced boiling heat transfer in horizontal test bundles

    SciTech Connect (OSTI)

    Trewin, R.R.; Jensen, M.K.; Bergles, A.E.

    1994-08-01T23:59:59.000Z

    Two-phase flow boiling from bundles of horizontal tubes with smooth and enhanced surfaces has been investigated. Experiments were conducted in pure refrigerant R-113, pure R-11, and mixtures of R-11 and R-113 of approximately 25, 50, and 75% of R-113 by mass. Tests were conducted in two staggered tube bundles consisting of fifteen rows and five columns laid out in equilateral triangular arrays with pitch-to-diameter ratios of 1.17 and 1.5. The enhanced surfaces tested included a knurled surface (Wolverine`s Turbo-B) and a porous surface (Linde`s High Flux). Pool boiling tests were conducted for each surface so that reference values of the heat transfer coefficient could be obtained. Boiling heat transfer experiments in the tube bundles were conducted at pressures of 2 and 6 bar, heat flux values from 5 to 80 kW/m{sup 2}s, and qualities from 0% to 80%, Values of the heat transfer coefficients for the enhanced surfaces were significantly larger than for the smooth tubes and were comparable to the values obtained in pool boiling. It was found that the performance of the enhanced tubes could be predicted using the pool boiling results. The degradation in the smooth tube heat transfer coefficients obtained in fluid mixtures was found to depend on the difference between the molar concentration in the liquid and vapor.

  10. CORQUENCH: A model for gas sparging-enhanced, melt-water, film-boiling heat transfer

    SciTech Connect (OSTI)

    Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.

    1990-01-01T23:59:59.000Z

    In evaluation of severe-accident sequences for water-cooled nuclear reactors, molten core materials may be postulated to be released into the containment and accumulate on concrete. The heatup and decomposition of concrete is accompanied by the release of water vapor and carbon dioxide gases. Gases flowing through the melt upper surface can influence the rates of heat transfer to water overlying the melt. In particular, the gas flow through the interface can be envisioned to enhance the heat removal from the melt. A mechanistic model (CORQUENCH) has been developed to describe film-boiling heat transfer between a molten pool and an overlying coolant layer in the presence of sparging gas. The model favorably predicts the lead-Feron 11 data of Greene and Greene et al. for which the calculations indicate that area enhancement in the conduction heat transfer across the film is the predominant mechanism leading to augmentation in the heat flux as the gas velocity increases. Predictions for oxidic corium indicate a rapid increase in film-boiling heat flux as the gas velocity rises. The predominant mode of heat transfer for this case is radiation, and the increase in heat flux with gas velocity is primarily a result of interfacial area enhancement of the radiation component of the overall heat transfer coefficient. The CORQUENCH model has been incorporated into the MELTSPREAD-1 computer code{sup 6} for the analysis of transient spreading in containments.

  11. Design, fabrication, packaging and testing of thin film thermocouples for boiling studies

    E-Print Network [OSTI]

    Sinha, Nipun

    2009-06-02T23:59:59.000Z

    Boiling is the most efficient form of heat transfer. Thermo-fluidic transport mechanisms at different length and time scales govern the nature of boiling. This study was conducted to enhance the understanding of the surface temperature variations...

  12. Mechanism and behavior of nucleate boiling heat transfer to the alkalai liquid metals

    E-Print Network [OSTI]

    Deane, Charles William

    1969-01-01T23:59:59.000Z

    A model of boiling heat transfer to the alkali liquid metals is postulated from an examination of the events and phases of the nucleate boiling cycle. The model includes the important effect of microlayer evaporation which ...

  13. Forced-convection surface-boiling heat transfer and burnout in tubes of small diameters

    E-Print Network [OSTI]

    Bergles A. E.

    1962-01-01T23:59:59.000Z

    A basic heat-transfer apparatus was designed and constructed for the study of forced-convection boiling in small channels. The various regions of forced-convection surface boiling were studied experimentally and analytically. ...

  14. Alumina Nanoparticle Pre-coated Tubing Ehancing Subcooled Flow Boiling Cricital Heat Flux

    E-Print Network [OSTI]

    Truong, Bao H.

    Nanofluids are engineered colloidal dispersions of nano-sized particle in common base fluids. Previous pool boiling studies have shown that nanofluids can improve critical heat flux (CHF) up to 200% for pool boiling and ...

  15. Acoustic emission feedback control for control of boiling in a microwave oven

    DOE Patents [OSTI]

    White, Terry L. (Oak Ridge, TN)

    1991-01-01T23:59:59.000Z

    An acoustic emission based feedback system for controlling the boiling level of a liquid medium in a microwave oven is provided. The acoustic emissions from the medium correlated with surface boiling is used to generate a feedback control signal proportional to the level of boiling of the medium. This signal is applied to a power controller to automatically and continuoulsly vary the power applied to the oven to control the boiling at a selected level.

  16. NUMERICAL SIMULATION OF POOL BOILING FOR STEADY STATE AND TRANSIENT HEATING

    E-Print Network [OSTI]

    Maruyama, Shigeo

    1 NUMERICAL SIMULATION OF POOL BOILING FOR STEADY STATE AND TRANSIENT HEATING Ying He, Masahiro role in nucleate and transition boiling heat transfer at high heat flux. Many experiments have been in the numerical simulation of boiling heat transfer. In this study, based on the macrolayer evaporation model

  17. ECI International Conference on Boiling Heat Transfer Florianpolis-SC-Brazil, 3-7 May 2009

    E-Print Network [OSTI]

    Yanikoglu, Berrin

    . Recently the study of heat transfer in micro pin fin heat sinks has been extended to flow boiling by Koar and Peles (2006c), who studied boiling heat transfer in a hydrofoil-based micro pin fin heat sinkECI International Conference on Boiling Heat Transfer Florianópolis-SC-Brazil, 3-7 May 2009

  18. Flow Boiling Heat Transfer Coefficient In Minichannels Correlation and Trends Satish G. Kandlikar

    E-Print Network [OSTI]

    Kandlikar, Satish

    Flow Boiling Heat Transfer Coefficient In Minichannels ­ Correlation and Trends Satish G. Kandlikar York 14623, USA The flow boiling heat transfer in small diameter passages is being applied in many boiling heat transfer coefficient with the correlations developed for conventional channels. It is found

  19. Numerical study of high heat ux pool boiling heat transfer Ying He a,*, Masahiro Shoji b

    E-Print Network [OSTI]

    Maruyama, Shigeo

    Numerical study of high heat ¯ux pool boiling heat transfer Ying He a,*, Masahiro Shoji b , Shigeo simulation model of boiling heat transfer is proposed based on a numerical macrolayer model [S. Maruyama, M. Shoji, S. Shimizu, A numerical simulation of transition boiling heat transfer, in: Proceedings

  20. Enhancement of Pool Boiling Heat Transfer in Confined Space 

    E-Print Network [OSTI]

    Hsu, Chia-Hsiang

    2014-05-05T23:59:59.000Z

    on pool boiling. In the study, confinement was achieved by placing a flat plate over heated surface. The flat plate has a hole in the middle, and there is a gap between the flat plate and the heater. The diameters of hole are 2 mm, 3 mm, and 4 mm; the gap...

  1. Jet Flows Around Microbubbles In Subcooled Boiling , Xiaofeng Pengb

    E-Print Network [OSTI]

    Kihm, IconKenneth David

    Jet Flows Around Microbubbles In Subcooled Boiling Hao Wanga , Xiaofeng Pengb , David M Strong jet flows were observed emanating from micro bubbles on a 100 µm diameter wire during subcooled analysis. The bubble-top jet flows were characterized by a single jet at the bubble top. Both experiments

  2. Modeling acid-gas generation from boiling chloride brines

    E-Print Network [OSTI]

    Zhang, Guoxiang

    2010-01-01T23:59:59.000Z

    solutions, Journal of nuclear technology 2008, 164:180– 35.Modeling and Analysis. Nuclear Technology 1993, 104: 418- 2.

  3. NUMERICAL ANALYSIS OF VAPOR BUBBLE GROWTH AND WALL HEAT TRANSFER DURING FLOW BOILING OF WATER IN A MICROCHANNEL

    E-Print Network [OSTI]

    Kandlikar, Satish

    . (2004) developed three- zone flow boiling heat transfer model to describe evaporation of elongated

  4. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    SciTech Connect (OSTI)

    Not Available

    1994-07-01T23:59:59.000Z

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  5. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    SciTech Connect (OSTI)

    Not Available

    1994-07-01T23:59:59.000Z

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  6. Tour of Entergy's Nuclear Power Plant in River Bend Owner: Entergy Gulf States Inc.

    E-Print Network [OSTI]

    Ervin, Elizabeth K.

    Tour of Entergy's Nuclear Power Plant in River Bend Owner: Entergy Gulf States Inc. Reactor Type a nuclear power plant. Plant was Entergy, a Boiling Water Reactor (BWR) type. Built in the 80's, it has of the veteran plant workers. The presentation gave the nuclear plant engineering basics and built

  7. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    SciTech Connect (OSTI)

    Not Available

    1994-06-01T23:59:59.000Z

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

  8. The Neutronics Design and Analysis of a 200-MW(electric) Simplified Boiling Water Reactor Core

    SciTech Connect (OSTI)

    Tinkler, Daniel R.; Downar, Thomas J. [Purdue University (United States)

    2003-06-15T23:59:59.000Z

    A 200-MW(electric) simplified boiling water reactor (SBWR) was designed and analyzed under sponsorship of the U.S. Department of Energy Nuclear Energy Research Initiative program. The compact size of a 200-MW(electric) reactor makes it attractive for countries with a less well developed engineering infrastructure, as well as for developed countries seeking to tailor generation capacity more closely to the growth of their electricity demand. The 200-MW(electric) core design reported here is based on the 600-MW(electric) General Electric SBWR core, which was first analyzed in the work performed here in order to qualify the computer codes used in the analysis. Cross sections for the 8 x 8 fuel assembly design were generated with the HELIOS lattice physics code, and core simulation was performed with the U.S. Nuclear Regulatory Commission codes RELAP5/PARCS. In order to predict the critical heat flux, the Hench-Gillis correlation was implemented in the RELAP5 code. An equilibrium cycle was designed for the 200-MW(electric) core, which provided a cycle length of more than 2 yr and satisfied the minimum critical power ratio throughout the core life.

  9. Thermal stability of a spherical shell heated by convection and cooled by boiling

    E-Print Network [OSTI]

    Qaim-Maqami, Hassan

    1973-01-01T23:59:59.000Z

    such high heat transfer rates by forced convection would require extremely high veloc- ities with subsequent high pressure dropsy With nuclee, boiling, however, such high heat transfe. rates an be obtained at much smaller velocities. Boiling heat...THERMAL STABILITY OF A. SPHERICAL SHELL HEATED BY CONVECTION AND COOLED BY BOILING A Thesis HASSAN @AIM-MAQAMI Submitted to the Graduate College of Texas ASM University in partial fulfillment of the requirement for the degree of MASTER...

  10. Great Boiling Springs Geothermal Area | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdfGetec AG Contracting JumpGoveNebraska: EnergyGratingsGreat BasinGreat Boiling

  11. Efficient Cooling in Engines with Nucleate Boiling | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't Your Destiny:RevisedAdvisory BoardNucleate Boiling Efficient Cooling in Engines with Nucleate

  12. Efficient Cooling in Engines with Nucleated Boiling | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't Your Destiny:RevisedAdvisory BoardNucleate Boiling Efficient Cooling in Engines with

  13. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    SciTech Connect (OSTI)

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06T23:59:59.000Z

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are similar to those obtained with the traditional UO2 nuclear fuel.

  14. Water inventory management in condenser pool of boiling water reactor

    DOE Patents [OSTI]

    Gluntz, Douglas M. (San Jose, CA)

    1996-01-01T23:59:59.000Z

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

  15. Boiling water neutronic reactor incorporating a process inherent safety design

    DOE Patents [OSTI]

    Forsberg, Charles W. (Kingston, TN)

    1987-01-01T23:59:59.000Z

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  16. Boiling water neutronic reactor incorporating a process inherent safety design

    DOE Patents [OSTI]

    Forsberg, C.W.

    1985-02-19T23:59:59.000Z

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  17. Water inventory management in condenser pool of boiling water reactor

    DOE Patents [OSTI]

    Gluntz, D.M.

    1996-03-12T23:59:59.000Z

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

  18. Comparison of thorium-based fuels with different fissile components in existing boiling water reactors

    E-Print Network [OSTI]

    Demazière, Christophe

    Comparison of thorium-based fuels with different fissile components in existing boiling water, SE-412 96 Göteborg, Sweden Keywords: Thorium BWR Neutronics a b s t r a c t With the aim of investigating the technical feasibility of fuelling a conventional BWR (Boiling Water Reactor) with thorium

  19. INVESTIGATING THE EFFECT OF HEATING METHOD ON POOL BOILING HEAT TRANSFER

    E-Print Network [OSTI]

    Kandlikar, Satish

    INVESTIGATING THE EFFECT OF HEATING METHOD ON POOL BOILING HEAT TRANSFER Satish G. Kandlikar Department Rochester Institute of Technology Rochester, NY, USA Phone: (716) 475-6728; Fax: (716) 475-7710 E-mail: SGKEME@RIT.EDU ABSTRACT Pool boiling experiments are generally conducted with electrically heated

  20. Effect of surface conditions on boiling heat transfer of refrigerants in shell-and-tube evaporators

    SciTech Connect (OSTI)

    Danilova, G.N.; Dyundin, V.A.; Borishanskaya, A.V.; Soloviyov, A.G.; Vol'nykh, Y.A.; Kozyrev, A.A.

    1990-01-01T23:59:59.000Z

    Experimental results are presented for the boiling heat transfer performance of R 22 and R 717 on surfaces with porous metallized coatings. A calculational-theoretical model is given for predicting the heat transfer of refrigerants boiling on a bundle of finned tubes.

  1. Boiling heat transfer in a hydrofoil-based micro pin fin heat sink

    E-Print Network [OSTI]

    Peles, Yoav

    transfer in a horizontal tube bundle and reported an increase in local heat transfer coefficient boiling heat transfer of water in small horizontal tube bundles at low velocities. How- ever, the dataBoiling heat transfer in a hydrofoil-based micro pin fin heat sink Ali Kosßar, Yoav Peles

  2. Vapor pressure and boiling point elevation of slash pine black liquors: Predictive models with statistical approach

    SciTech Connect (OSTI)

    Zaman, A.A.; McNally, T.W.; Fricke, A.L. [Univ. of Florida, Gainesville, FL (United States)] [Univ. of Florida, Gainesville, FL (United States)

    1998-01-01T23:59:59.000Z

    Vapor-liquid equilibria and boiling point elevation of slash pine kraft black liquors over a wide range of solid concentrations (up to 85% solids) has been studied. The liquors are from a statistically designed pulping experiment for pulping slash pine in a pilot scale digester with four cooking variables of effective alkali, sulfidity, cooking time, and cooking temperature. It was found that boiling point elevation of black liquors is pressure dependent, and this dependency is more significant at higher solids concentrations. The boiling point elevation data at different solids contents (at a fixed pressure) were correlated to the dissolved solids (S/(1 {minus} S)) in black liquor. Due to the solubility limit of some of the salts in black liquor, a change in the slope of the boiling point elevation as a function of the dissolved solids was observed at a concentration of around 65% solids. An empirical method was developed to describe the boiling point elevation of each liquor as a function of pressure and solids mass fraction. The boiling point elevation of slash pine black liquors was correlated quantitatively to the pulping variables, using different statistical procedures. These predictive models can be applied to determine the boiling point rise (and boiling point) of slash pine black liquors at processing conditions from the knowledge of pulping variables. The results are presented, and their utility is discussed.

  3. Boiling Crisis as a Non-Equilibrium Drying V. S. Nikolayev, D. A. Beysens, J. Hegseth

    E-Print Network [OSTI]

    Nikolayev, Vadim S.

    crisis is the formation of a vapor film between the heater and the liquid when the heat supply exceeds in the coordinates heat supply ­ heater temperature is sketched in Fig. 1 for the case of stationary boiling experiment, the so called "pool boiling". When the heat supply to the fluid pool is small, only a fluid

  4. A Review Of Recent Progress On Nano/micro Scale Nucleate Boiling Fundamentals

    E-Print Network [OSTI]

    Chung, J. N.

    2011-01-01T23:59:59.000Z

    Recent research progress in the area of nano/micro scale nucleate boiling is reviewed and an up-to-date summary is provided with a focus on the advances of fundamental boiling physics. This review examines nano/micro scale ...

  5. Effects of Carbon Nanotube Coating on Bubble Departure Diameter and Frequency in Pool Boiling on a Flat, Horizontal Heater 

    E-Print Network [OSTI]

    Glenn, Stephen T.

    2011-08-08T23:59:59.000Z

    The effects of a carbon nanotube (CNT) coating on bubble departure diameter and frequency in pool boiling experiments was investigated and compared to those on a bare silicon wafer. The pool boiling experiments were performed at liquid subcooling...

  6. Effects of Carbon Nanotube Coating on Bubble Departure Diameter and Frequency in Pool Boiling on a Flat, Horizontal Heater

    E-Print Network [OSTI]

    Glenn, Stephen T.

    2011-08-08T23:59:59.000Z

    The effects of a carbon nanotube (CNT) coating on bubble departure diameter and frequency in pool boiling experiments was investigated and compared to those on a bare silicon wafer. The pool boiling experiments were performed at liquid subcooling...

  7. Nuclear reactor safety heat transfer

    SciTech Connect (OSTI)

    Jones, O.C.

    1982-07-01T23:59:59.000Z

    Reviewed is a book which has 5 parts: Overview, Fundamental Concepts, Design Basis Accident-Light Water Reactors (LWRs), Design Basis Accident-Liquid-Metal Fast Breeder Reactors (LMFBRs), and Special Topics. It combines a historical overview, textbook material, handbook information, and the editor's personal philosophy on safety of nuclear power plants. Topics include thermal-hydraulic considerations; transient response of LWRs and LMFBRs following initiating events; various accident scenarios; single- and two-phase flow; single- and two-phase heat transfer; nuclear systems safety modeling; startup and shutdown; transient response during normal and upset conditions; vapor explosions, natural convection cooling; blockages in LMFBR subassemblies; sodium boiling; and Three Mile Island.

  8. Rod bundle film boiling and steam cooling data base and correlation evaluation

    SciTech Connect (OSTI)

    Yoder, G.L.

    1986-08-01T23:59:59.000Z

    Film boiling and steam cooling data from several rod bundle experiments have been compiled from fourteen references. Steam cooling data are presented for five references, while film boiling data are presented for ten references (two of these for tube geometry). The compilation includes experimental parameters necessary for characterization of the local rod heat transfer. These data have been compared to several commonly used film boiling and pure vapor heat transfer coefficients. Results show that the Dougall-Rohsenow correlation tends to overpredict film boiling heat transfer, while Condie-Bengston IV and the Groeneveld 5. series correlations do a reasonable job of predicting film boiling heat fluxes. Of the three vapor correlations examined, the Dittus-Boelter correlation is least conservative, and the Sozer correlation most conservative.

  9. Experimental investigation of nucleate boiling heat transfer mechanisms for cylinders in water and FC-72

    SciTech Connect (OSTI)

    Ammerman, C.N.; You, S.M.; Hong, Y.S. [Univ. of Texas, Arlington, TX (United States). Dept. of Mechanical and Aerospace Engineering

    1995-12-31T23:59:59.000Z

    A recently developed photographic method is used to quantify vapor volumetric flow rate above a boiling wire. The volumetric flow rate is combined with additional analyses to determine the overall contributions to the total heat flux from four nucleate boiling heat transfer mechanisms (latent heat, natural convection, Marangoni flow, and micro-convection). This technique is used to quantify the boiling heat transfer mechanisms versus heat flux for a 510-{micro}m wire immersed in saturated water and in water with a small amount of liquid soap added. These data are compared with similar data taken for a 75-{micro}m wire boiling in saturated FC-72. For all cases, latent heat is the dominant heat transfer mechanism in the fully developed nucleate boiling regime. In addition, the latent heat component is significantly increased by the addition of small amounts of soap (surfactant).

  10. Determination of the boiling enhancement mechanism caused by surfactant addition to water

    SciTech Connect (OSTI)

    Ammerman, C.N.; You, S.M. [Univ. of Texas, Arlington, TX (United States). Dept. of Mechanical and Aerospace Engineering

    1995-12-31T23:59:59.000Z

    In the present investigation, boiling heat transfer coefficients are measured for an electrically heated 390-{micro}m, platinum wire immersed in saturated water, and in water mixed with three different concentrations of sodium dodecyl sulfate (an anionic surfactant). The addition of a surfactant to water is known to enhance boiling heat transfer. A recently developed photographic/laser Doppler anemometry measurement technique is used to quantify the vapor volumetric flowrate departing from the wire during the boiling process. The volumetric flowrate data are combined with results from additional analyses to determine the overall contributions to the total heat flux from three nucleate boiling heat transfer mechanisms. Comparisons are made to determine which heat transfer mechanisms are affected by the surfactant addition, and thus, which mechanisms promote boiling enhancement.

  11. Wavelet-Based Method for Instability Analysis in Boiling Water Reactors

    SciTech Connect (OSTI)

    Espinosa-Paredes, Gilberto; Prieto-Guerrero, Alfonso; Nunez-Carrera, Alejandro; Amador-Garcia, Rodolfo

    2005-09-15T23:59:59.000Z

    This paper introduces a wavelet-based method to analyze instability events in a boiling water reactor (BWR) during transient phenomena. The methodology to analyze BWR signals includes the following: (a) the short-time Fourier transform (STFT) analysis, (b) decomposition using the continuous wavelet transform (CWT), and (c) application of multiresolution analysis (MRA) using discrete wavelet transform (DWT). STFT analysis permits the study, in time, of the spectral content of analyzed signals. The CWT provides information about ruptures, discontinuities, and fractal behavior. To detect these important features in the signal, a mother wavelet has to be chosen and applied at several scales to obtain optimum results. MRA allows fast implementation of the DWT. Features like important frequencies, discontinuities, and transients can be detected with analysis at different levels of detail coefficients. The STFT was used to provide a comparison between a classic method and the wavelet-based method. The damping ratio, which is an important stability parameter, was calculated as a function of time. The transient behavior can be detected by analyzing the maximum contained in detail coefficients at different levels in the signal decomposition. This method allows analysis of both stationary signals and highly nonstationary signals in the timescale plane. This methodology has been tested with the benchmark power instability event of Laguna Verde nuclear power plant (NPP) Unit 1, which is a BWR-5 NPP.

  12. Feasibility study on the thorium fueled boiling water breeder reactor

    SciTech Connect (OSTI)

    PetrusTakaki, N. [Dept. of Applied Science, Tokai Univ., Kitakaname, Hiratsuka, Kanagawa 259-1292 (Japan)

    2012-07-01T23:59:59.000Z

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  13. Nonlinear dynamics and chaos in boiling water reactors

    SciTech Connect (OSTI)

    March-Leuba, J.

    1988-01-01T23:59:59.000Z

    There are currently 72 commercial boiling water reactors (BWRs) in operation or under construction in the western world, 37 of them in the United States. Consequently, a great effort has been devoted to the study of BWR systems under a wide range of plant operating conditions. This paper represents a contribution to this ongoing effort; its objective is to study the basic dynamic processes in BWR systems, with special emphasis on the physical interpretation of BWR dynamics. The main thrust in this work is the development of phenomenological BWR models suited for analytical studies performed in conjunction with numerical calculations. This approach leads to a deeper understanding of BWR dynamics and facilitates the interpretation of numerical results given by currently available sophisticated BWR codes. 6 refs., 14 figs., 2 tabs.

  14. Boiling Heat Transfer on Superhydrophilic, Superhydrophobic, and Superbiphilic Surfaces

    E-Print Network [OSTI]

    Betz, Amy Rachel; Kim, Chang-Jin 'CJ'; Attinger, Daniel

    2012-01-01T23:59:59.000Z

    With recent advances in micro- and nanofabrication, superhydrophilic and superhydrophobic surfaces have been developed. The statics and dynamics of fluids on these surfaces have been well characterized. However, few investigations have been made into the potential of these surfaces to control and enhance other transport phenomena. In this article, we characterize pool boiling on surfaces with wettabilities varied from superhydrophobic to superhydrophilic, and provide nucleation measurements. The most interesting result of our measurements is that the largest heat transfer coefficients are reached not on surfaces with spatially uniform wettability, but on biphilic surfaces, which juxtapose hydrophilic and hydrophobic regions. We develop an analytical model that describes how biphilic surfaces effectively manage the vapor and liquid transport, delaying critical heat flux and maximizing the heat transfer coefficient. Finally, we manufacture and test the first superbiphilic surfaces (juxtaposing superhydrophobic ...

  15. Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

    SciTech Connect (OSTI)

    NONE

    1994-04-30T23:59:59.000Z

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

  16. Identification of pool boiling heat transfer mechanisms in FC-72 using a single-photo method

    SciTech Connect (OSTI)

    Ammerman, C.N.; You, S.M.; Hong, Y.S. [Univ. of Texas, Arlington, TX (United States). Dept. of Mechanical and Aerospace Engineering

    1995-10-01T23:59:59.000Z

    A unique method to determine the vapor flow rate above a boiling cylinder utilizing a single photograph is developed and discussed. This method is applied to a 75-{micro}m wire immersed in a saturated, highly wetting liquid (FC-72) to determine bubble departure diameter, frequency, and nucleation site density. Using the experimental results, an analysis is performed to evaluate individual heat flux contributions of the four pool boiling mechanisms: latent heat, natural convection, Marangoni flow, and micro-convection. Latent heat is identified as the dominant mechanism throughout most of the nucleate boiling regime.

  17. CORQUENCH: A model for gas sparging-enhanced melt-water, film boiling heat transfer

    SciTech Connect (OSTI)

    Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.

    1990-01-01T23:59:59.000Z

    A phenomenological model (CORQUENCH) has been developed to describe the gas-sparging enhanced film boiling heat transfer between a molten pool of corium and an overlying water layer. The model accounts for thermal radiation across the vapor film, bulk liquid subcooling, interfacial area enhancement due to sparging gas, and melt entrainment into the overlying water layer. In this paper, the modeling approach is described, and a comparison with the lead-Freon 11 and lead-water film boiling experiment data of Greene is made. Predictions are then made for the case of film boiling over corium in the presence of sparging concrete decomposition gases. 15 refs., 3 figs.

  18. New flow boiling heat transfer model for hydrocarbons evaporating inside horizontal tubes

    SciTech Connect (OSTI)

    Chen, G. F.; Gong, M. Q.; Wu, J. F.; Zou, X. [Key Laboratory of Cryogenics, Technical Institute of Physics and Chemistry, Chinese Academy of Sciences, P.O. Box 2711, 35, Beijing, 100190 (China); Wang, S. [Key Laboratory of Cryogenics, Technical Institute of Physics and Chemistry, Chinese Academy of Sciences, P.O. Box 2711, 35, Beijing, 100190 (China); University of Chinese Academy of Science, No. 19 YuQuan Road, Beijing, 100049 (China)

    2014-01-29T23:59:59.000Z

    Hydrocarbons have high thermodynamic performances, belong to the group of natural refrigerants, and they are the main components in mixture Joule-Thomson low temperature refrigerators (MJTR). New evaluations of nucleate boiling contribution and nucleate boiling suppression factor in flow boiling heat transfer have been proposed for hydrocarbons. A forced convection heat transfer enhancement factor correlation incorporating liquid velocity has also been proposed. In addition, the comparisons of the new model and other classic models were made to evaluate its accuracy in heat transfer prediction.

  19. Method of controlling crystallite size in nuclear-reactor fuels

    DOE Patents [OSTI]

    Lloyd, Milton H. (Oak Ridge, TN); Collins, Jack L. (Knoxville, TN); Shell, Sam E. (Oak Ridge, TN)

    1985-01-01T23:59:59.000Z

    Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules.

  20. Experimental Two-Phase Flow Characterization of Subcooled Boiling in a Rectangular Channel

    E-Print Network [OSTI]

    Estrada Perez, Carlos E.

    2010-01-16T23:59:59.000Z

    On the efforts to provide a reliable source of experimental information on turbulent subcooled boiling ow, time resolved Particle Tracking Velocimetry (PTV) experiments were carried out using HFE-301 refrigerant ow through a vertical rectangular...

  1. Development of a model to predict flow oscillations in low-flow sodium boiling

    E-Print Network [OSTI]

    Levin, Alan Edward

    1980-01-01T23:59:59.000Z

    An experimental and analytical program has been carried out in order to better understand the cause and effect of flow oscillations in boiling sodium systems. These oscillations have been noted in previous experiments with ...

  2. Bubble behavior in subcooled flow boiling on surfaces of variable wettability

    E-Print Network [OSTI]

    Tow, Emily W

    2012-01-01T23:59:59.000Z

    Flow boiling is important in energy conversion and thermal management due to its potential for very high heat fluxes. By improving understanding of the conditions leading to bubble departure, surfaces can be designed that ...

  3. Stability analysis of the boiling water reactor : methods and advanced designs

    E-Print Network [OSTI]

    Hu, Rui, Ph. D. Massachusetts Institute of Technology

    2010-01-01T23:59:59.000Z

    Density Wave Oscillations (DWOs) are known to be possible when a coolant undergoes considerable density reduction while passing through a heated channel. In the development of boiling water reactors (BWRs), there has been ...

  4. Visualization of flow boiling in an annular heat exchanger under reduced gravity conditions

    E-Print Network [OSTI]

    Westheimer, David Thomas

    2000-01-01T23:59:59.000Z

    This work examines the effects of gravitational acceleration on the flow boiling process. A test facility focusing on an annular heat exchanger was designed, built out of borosilicate glass, and flown on NASA's KC-135 reduced gravity airplane...

  5. A visualization comparison of convective flow boiling heat transfer augmentation devices

    E-Print Network [OSTI]

    Lundy, Brian Franklin

    1998-01-01T23:59:59.000Z

    The qualitative effects of inset-table heat transfer phics. augmentation devices on vertical in-tube convective flow boiling flow regimes, transition mechanisms, and heat transfer are presented in this study. Three twisted tapes with twist ratios...

  6. Experimental Observation and Measurements of Pool Boiling Heat Transfer using PIV, Shadowgraphy, RICM Techniques

    E-Print Network [OSTI]

    Di, Yuan 1988-

    2012-12-05T23:59:59.000Z

    This present study seeks to contribute detailed visualization data on a pool boiling experiments using HFE-7000. Particle Image Velocimetry (PIV) was used to measure the time resolved whole field liquid velocity. Bubble dynamic parameters...

  7. Experimental & Numerical Investigation of Pool Boiling on Engineered Surfaces with Integrated Thin-flim Temperature Sensors

    E-Print Network [OSTI]

    Sathyamurthi, Vijaykumar

    2011-02-22T23:59:59.000Z

    of subcooling and micro/nano-scale surface texturing using MWCNT coatings on the dynamics of pool boiling is assessed. Dynamic invariants such as correlation dimensions and Lyapunov spectrum are evaluated for the reconstructed attractor. A non-linear noise...

  8. Prediction of departure from nucleate boiling in PWR fast power transients

    E-Print Network [OSTI]

    Lenci, Giancarlo

    2013-01-01T23:59:59.000Z

    An assessment is conducted of the differences in predicted results between use of steady state versus transient Departure from Nucleate Boiling (DNB) models, for fast power transients under forced convective heat exchange ...

  9. A study of system-induced instabilities in forced-convection flows with subcooled boiling

    E-Print Network [OSTI]

    Maulbetsch, John S.

    1965-01-01T23:59:59.000Z

    A combined analytical and experimental program was carried out to investigate the problem of hydrodynamic stability of forcedconvection flows with boiling. The study was restricted to the flow of water in small channels ...

  10. Film boiling of saturated liquid flowing upward through a heated tube : high vapor quality range

    E-Print Network [OSTI]

    Laverty, W. F.

    1964-01-01T23:59:59.000Z

    Film boiling of saturated liquid flowing upward through a uniformly heated tube has been studied for the case in which pure saturated liquid enters the tube and nearly saturated vapor is discharged. Since a previous study ...

  11. Thermal non-equilibrium in dispersed flow film boiling in a vertical tube

    E-Print Network [OSTI]

    Forslund, Robert Paul

    1966-01-01T23:59:59.000Z

    The departure from thermal equilibrium between a dispersed liquid phase and its vapor at high quality during film boiling is investigated, The departure from equilibruim is manifested by the high resistance to heat transfer ...

  12. Experimental & Numerical Investigation of Pool Boiling on Engineered Surfaces with Integrated Thin-flim Temperature Sensors 

    E-Print Network [OSTI]

    Sathyamurthi, Vijaykumar

    2011-02-22T23:59:59.000Z

    The objective of this investigation is to measure and analyze surface temperature fluctuations in pool boiling. The surface temperature fluctuations were recorded on silicon surfaces with and without multi-walled carbon nanotubes (MWCNT). Novel Thin...

  13. Nano-engineering the boiling surface for optimal heat transfer rate and critical heat flux

    E-Print Network [OSTI]

    Phillips, Bren Andrew

    2011-01-01T23:59:59.000Z

    The effects on pool boiling characteristics such as critical heat flux and the heat transfer coefficient of different surface characteristics such as surface wettability, roughness, morphology, and porosity are not well ...

  14. Enhanced convective and film boiling heat transfer by surface gas injection

    SciTech Connect (OSTI)

    Duignan, M.R.; Greene, G.A. [Brookhaven National Lab., Upton, NY (United States); Irvine, T.F., Jr. [State Univ. of New York, Stony Brook, NY (United States). Dept. of Mechanical Engineering

    1992-04-01T23:59:59.000Z

    Heat transfer measurements were made for stable film boiling of water over a horizontal, flat stainless steel plate from the minimum film boiling point temperature, T{sub SURFACE} {approximately}500K, to T{sub SURFACE} {approximately}950K. The pressure at the plate was approximately 1 atmosphere and the temperature of the water pool was maintained at saturation. The data were compared to the Berenson film-boiling model, which was developed for minimum film-boiling-point conditions. The model accurately represented the data near the minimum film-boiling point and at the highest temperatures measured, as long it was corrected for the heat transferred by radiation. On the average, the experimental data lay within {plus_minus}7% of the model. Measurements of heat transfer were made without film boiling for nitrogen jetting into an overlying pool of water from nine 1-mm- diameter holes, drilled in the heat transfer plate. The heat flux was maintained constant at approximately 26.4 kW/m{sup 2}. For water-pool heights of less than 6cm the heat transfer coefficient deceased linearly with a decrease in heights. Above 6cm the heat transfer coefficient was unaffected. For the entire range of gas velocities measured [0 to 8.5 cm/s], the magnitude of the magnitude of the heat transfer coefficient only changed by approximately 20%. The heat transfer data bound the Konsetov model for turbulent pool heat transfer which was developed for vertical heat transfer surfaces. This agreement suggests that surface orientation may not be important when the gas jets do not locally affect the surface heat transfer. Finally, a database was developed for heat transfer from the plate with both film boiling and gas jetting occurring simultaneously, in a pool of water maintained at its saturation temperature. The effect of passing nitrogen through established film boiling is to increase the heat transfer from that surface. 60 refs.

  15. Enhanced convective and film boiling heat transfer by surface gas injection

    SciTech Connect (OSTI)

    Duignan, M.R.; Greene, G.A. (Brookhaven National Lab., Upton, NY (United States)); Irvine, T.F., Jr. (State Univ. of New York, Stony Brook, NY (United States). Dept. of Mechanical Engineering)

    1992-04-01T23:59:59.000Z

    Heat transfer measurements were made for stable film boiling of water over a horizontal, flat stainless steel plate from the minimum film boiling point temperature, T{sub SURFACE} {approximately}500K, to T{sub SURFACE} {approximately}950K. The pressure at the plate was approximately 1 atmosphere and the temperature of the water pool was maintained at saturation. The data were compared to the Berenson film-boiling model, which was developed for minimum film-boiling-point conditions. The model accurately represented the data near the minimum film-boiling point and at the highest temperatures measured, as long it was corrected for the heat transferred by radiation. On the average, the experimental data lay within {plus minus}7% of the model. Measurements of heat transfer were made without film boiling for nitrogen jetting into an overlying pool of water from nine 1-mm- diameter holes, drilled in the heat transfer plate. The heat flux was maintained constant at approximately 26.4 kW/m{sup 2}. For water-pool heights of less than 6cm the heat transfer coefficient deceased linearly with a decrease in heights. Above 6cm the heat transfer coefficient was unaffected. For the entire range of gas velocities measured (0 to 8.5 cm/s), the magnitude of the magnitude of the heat transfer coefficient only changed by approximately 20%. The heat transfer data bound the Konsetov model for turbulent pool heat transfer which was developed for vertical heat transfer surfaces. This agreement suggests that surface orientation may not be important when the gas jets do not locally affect the surface heat transfer. Finally, a database was developed for heat transfer from the plate with both film boiling and gas jetting occurring simultaneously, in a pool of water maintained at its saturation temperature. The effect of passing nitrogen through established film boiling is to increase the heat transfer from that surface. 60 refs.

  16. Camera Inspection Arm for Boiling Water Reactors - 13330

    SciTech Connect (OSTI)

    Martin, Scott; Rood, Marc [S.A. Technology, 3985 S. Lincoln Ave, Loveland, CO 80537 (United States)] [S.A. Technology, 3985 S. Lincoln Ave, Loveland, CO 80537 (United States)

    2013-07-01T23:59:59.000Z

    Boiling Water Reactor (BWR) outage maintenance tasks can be time-consuming and hazardous. Reactor facilities are continuously looking for quicker, safer, and more effective methods of performing routine inspection during these outages. In 2011, S.A. Technology (SAT) was approached by Energy Northwest to provide a remote system capable of increasing efficiencies related to Reactor Pressure Vessel (RPV) internal inspection activities. The specific intent of the system discussed was to inspect recirculation jet pumps in a manner that did not require manual tooling, and could be performed independently of other ongoing inspection activities. In 2012, SAT developed a compact, remote, camera inspection arm to create a safer, more efficient outage environment. This arm incorporates a compact and lightweight design along with the innovative use of bi-stable composite tubes to provide a six-degree of freedom inspection tool capable of reducing dose uptake, reducing crew size, and reducing the overall critical path for jet pump inspections. The prototype camera inspection arm unit is scheduled for final testing in early 2013 in preparation for the Columbia Generating Station refueling outage in the spring of 2013. (authors)

  17. Generic risk insights for General Electric boiling water reactors

    SciTech Connect (OSTI)

    Travis, R.; Taylor, J. (Brookhaven National Lab., Upton, NY (USA)); Chung, J. (Nuclear Regulatory Commission, Washington, DC (USA). Office of Nuclear Reactor Regulation)

    1991-05-01T23:59:59.000Z

    A methodology has been developed to extract generic risk-based information from probabilistic risk assessments (PRAs) of General Electric boiling water rectors and applying the insights gained to plants that have not been subjected to a PRA. The available risk assessments (six plants) were examined to identify the most probable, i.e., dominant accident sequences at each plants. The goal was to include all sequences which represented at least 80% of core damage frequency. If the same plant specific dominant accident sequence appeared within this boundary in at least two plant PRAs, the sequence was considered to be a representative sequence. Eight sequences met this definition. From these sequences, the most important component failures and human error that contributed to each sequence have been prioritized. Guidance is provided to prioritize the representative sequences and modify selected basic events that have been shown to be sensitive to the plant specific design or operating variations of the contributing PRAs. This risk-based guidance can be used for utility and NRC activities including operator training, maintenance, design review, and inspections. 13 refs., 6 tabs.

  18. Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

    SciTech Connect (OSTI)

    NONE

    1997-05-01T23:59:59.000Z

    This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff`s review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design.

  19. Bottom head to shell junction assembly for a boiling water nuclear reactor

    DOE Patents [OSTI]

    Fife, Alex Blair (San Jose, CA); Ballas, Gary J. (San Jose, CA)

    1998-01-01T23:59:59.000Z

    A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening.

  20. Gas bubbling-enhanced film boiling of Freon-11 on liquid metal pools. [PWR; BWR

    SciTech Connect (OSTI)

    Greene, G.A.

    1985-01-01T23:59:59.000Z

    In the analysis of severe core damage accidents in LWRs, a major driving force which must be considered in evaluating containment loading and fission product transport is the ex-vessel interaction between molten core debris and structural concrete. Two computer codes have been developed for this purpose, the CORCON-MOD2 model of ex-vessel, core concrete interactions and the VANESA model for aerosol generation and fission product release as a result of molten core-concrete interactions. Under a wide spectrum of reactor designs and accident sequences, it is possible for water to come into contact with the molten core debris and form a coolant pool overlying the core debris which is attacking the concrete. As the concrete decomposes, noncondensable gases are released, which bubble through the melt and across the boiling interface, affecting the liquid-liquid boiling process. Currently, the CORCON code includes the classical Berenson model for film boiling over a horizontal flat plate for this phenomenon. The objectives of this activity are to investigate the influence of transverse noncondensable gas flux on the magnitude of the stable liquid-liquid film boiling heat flux and develop a gas flux-enhanced, liquid-liquid film boiling model for incorporation into the CORCON-MOD2 computer code to replace or modify the Berenson model.

  1. Influences of electrical field on boiling-condensation heat transfer system

    SciTech Connect (OSTI)

    Yang Jiaxiang; Ding Lijian; Chi Xiaochun; Liu Ji [Harbin Inst. of Electrical Technology (China). Dept. of Electrical Materials Engineering; Yang He [Harbin Inst. of Tech. (China). Dept. of Thermal Energy Engineering

    1996-12-31T23:59:59.000Z

    In this paper, the influences of electrical field on boiling-condensation heat transfer system have been investigated using a cylinder heat transfer model. Freon-11 is selected as working fluid. The condensation heat transfer coefficient, the boiling heat flux and the saturation pressure are measured in this investigation. According to the experimental results, it is found that the electrical field can influence heat transfer system. The boiling heat transfer is enhanced by the applied voltage, and the saturate vapor of working fluid is condensed on the high voltage electrode directly when the applied voltage is higher than 6 kv. The experimental results have been discussed, and it is considered that the high electrical field strength change the thermal properties of working fluid.

  2. BOILING CRISIS:THEORY,SIMULATION,AND EXPERIMENTSBOILING CRISIS: THEORY, SIMULATION, AND EXPERIMENTS Boiling is a very efficient way to transfer heat from a heater to the liquid heat carrier. We discuss the

    E-Print Network [OSTI]

    Nikolayev, Vadim S.

    that separates the liquid from the heater. Film boiling can be also observed in the kitchen by sprinkling water onto a hot frying pan. In spite of the large temperature of the heater (the pan) the water drops Boiling is a very efficient way to transfer heat from a heater to the liquid heat carrier. We discuss

  3. Enhancement of pool boiling heat transfer with electrohydrodynamics and its fundamental study

    E-Print Network [OSTI]

    Raghupathi, Sri Laxmi Priya

    1998-01-01T23:59:59.000Z

    , and 10kV. To conduct this study, an existing low pressure pool boiling apparatus was modified and another high pressure apparatus was designed and built. The fluids were tested on the smooth tube at operating temperatures of 4'C and 20'C, and 19 fins... TWl Tvvo Heat flux of EHD enhancement of boiling (W/m ) Total heat flux with EHD at 10kV (W/m ) Total heat flux with no EHD (W/m ) Minimum cavity mouth radius (m) Rayleigh number Fluid temperature ('C) Pool Saturation temperature ('C) Tube...

  4. The use of the probability distribution function to analyze surface temperature fluctuations in pool boiling

    E-Print Network [OSTI]

    Tu, Chau Qui

    1976-01-01T23:59:59.000Z

    accepted. 3. Mass-transfer model: As early as 1952, the possi- 1. 1 bi) ity of mass transfer across the bubble was visualized oy Snyder and presented at a conference in 1856 for the ca. , e of subcool ed nucleate boiling in forced convection l qf. He... heat transfer in forced convection, found negligible latent heat transport compared to the total heat transferred from the heated surface. Ruling out the latent heat transport as major factor, they attributed the high heat transfer in boiling...

  5. Investigation of boiling heat transfer at a surface with a system of cylindrical cavities under conditions of free motion

    SciTech Connect (OSTI)

    Danilova, G.N.; Reznikov, V.I.

    1988-01-01T23:59:59.000Z

    The authors propose a mathematical model for the intensification of boiling heat transfer and the subsequent increase in thermal efficiency of the cylindrical heat transfer surfaces in an evaporative cooling system. The boiling curves for water, ethanol, and freon 113 are calculated for a surface with artificial nucleation sites. The model incorporates such coolant properties as surface tension, specific heat, and vaporization heat.

  6. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    SciTech Connect (OSTI)

    M. Ishii; S. T. Revankar; T. Downar; Y. Xu, H. J. Yoon; D. Tinkler; U. S. Rohatgi

    2003-06-16T23:59:59.000Z

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral system scaling analysis, design parameters were obtained and designs of the compact modular 200 MWe SBWR and the full size 1200 MWe SBWR were developed. These reactors are provided with passive safety systems. A new passive vacuum breaker check valve was designed to replace the mechanical vacuum beaker check valve. The new vacuum breaker check valve was based on a hydrostatic head, and was fail safe. The performance of this new valve was evaluated both by the thermal-hydraulic code RELAP5 and by the experiments in a scaled SBWR facility, PUMA. In the core neutronic design a core depletion model was implemented to PARCS code. A lattice design for the SBWR fuel assemblies was performed. Design improvements were made to the neutronics/thermal-hydraulics models of SBWR-200 and SBWR-1200, and design analyses of these reactors were performed. The design base accident analysis and evaluation of all the passive safety systems were completed as scheduled in tasks 4 and 5. Initial conditions for the small break loss of coolant accidents (LOCA) and large break LOCA using REALP5 code were obtained. Small and large break LOCA tests were performed and the data was analyzed. An anticipated transient with scram was simulated using the RELAP5 code for SBWR-200. The transient considered was an accidental closure of the main steam isolation valve (MSIV), which was considered to be the most significant transient. The evaluation of the RELAP5 code against experimental data for SBWR-1200 was completed. In task 6, the instability analysis for the three SBWR designs (SBWR-1200, SBWR-600 and SBWR-200) were simulated for start-up transients and the results were similar. Neither the geysering instability, nor the loop type instability was predicted by RAMONA-4B in the startup simulation following the recommended procedure by GE. The density wave oscillation was not observed at all because the power level used in the simulation was not high enough. A study was made of the potential instabilities by imposing an unrealistically high power ramp in a short time period, as suggested by GE. RAMON

  7. Nanoscale modification of key surface parameters to augment pool boiling heat transfer and critical heat flux in water and dielectric fluids

    E-Print Network [OSTI]

    Forrest, Eric Christopher

    2009-01-01T23:59:59.000Z

    Surface effects on pool boiling heat transfer and the critical heat flux are well documented but poorly understood. This study investigates the pool boiling characteristics of various fluids, and demonstrates that surface ...

  8. Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

    DOE Patents [OSTI]

    Tokarz, R.D.

    1981-10-27T23:59:59.000Z

    This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

  9. Simulation of subcooled boiling at low pressure conditions with RELAP5-3D computer program 

    E-Print Network [OSTI]

    Reza, S.M. Mohsin

    2002-01-01T23:59:59.000Z

    Simulation of subcooled boiling was carried out using RELAP5 thermal hydraulic computer programs. Both one-dimensional and three-dimensional analyses were carried out with one-dimensional RELAP5/MOD3.2 and three-dimensional RELAP5-3D code...

  10. "Hot Water" in Lassen Volcanic National Park--Fumaroles, Steaming Ground, and Boiling Mudpots

    E-Print Network [OSTI]

    Torgersen, Christian

    "Hot Water" in Lassen Volcanic National Park-- Fumaroles, Steaming Ground, and Boiling Mudpots U, ydrothermal (hot water) and steaming ground. These features are re- lated to active volcanism, the largest fumarole (steam and volcanic-gas vent) in the park. The temperature of the high-velocity steam

  11. QSPR models of boiling point, octanolwater partition coefficient and retention time index of polycyclic aromatic hydrocarbons

    E-Print Network [OSTI]

    Ferreira, Márcia M. C.

    QSPR models of boiling point, octanol­water partition coefficient and retention time index Structure­Property Relationship (QSPR) analysis and study of polycyclic aromatic hydrocarbons (PAHs (bp), octanol­ water partition coefficient ðlog KowÞ and retention time index (RI) for reversed

  12. Development of a small-channel nucleate-boiling heat transfer correlation

    SciTech Connect (OSTI)

    Kasza, K.E.; Wambsganss, M.W.

    1994-06-01T23:59:59.000Z

    Development of an improved semimechanistic-based set of correlation parameters for nucleation-dominant flow-boiling heat transfer in small channels is described. Formulation of these parameters is on the basis of a recently published open-literature model for vapor bubble growth at a heated surface. This work is part of a program directed at obtaining an understanding of the physical mechanisms that influence boiling in compact heat exchangers through the use of high-speed video and microscope optics to characterize bubble nucleation, growth, and interaction with the confining walls of small heat transfer passages. The correlation parameters presented here represent the first step in the development of an improved boiling correlation for geometrically confined small-channel flows. In such flows, the nucleating bubbles can become nominally the same size as the channel cross section, thereby invalidating existing correlations that are based on large-channel data. Initial efforts to correlate small-channel-boiling data obtained at Argonne National Laboratory from nontransparent electrically heated metal tube tests appear promising.

  13. 2007-No54-BoilingPoint Health and Greenhouse Gas Impacts of Biomass and Fossil Fuel

    E-Print Network [OSTI]

    Kammen, Daniel M.

    2007-No54-BoilingPoint Theme Health and Greenhouse Gas Impacts of Biomass and Fossil Fuel Energy of fossil-fuel energy systems. These scenarios are analysed for various environmental and health impacts from fossil fuels and other energy sources reported by IEA []. In all of these countries except Kenya

  14. Hydrodynamic cavitation and boiling in refrigerant (R-123) flow inside microchannels

    E-Print Network [OSTI]

    Peles, Yoav

    Hydrodynamic cavitation and boiling in refrigerant (R-123) flow inside microchannels Brandon cavitation has on heat transfer. The fluid medium is refrigerant R-123 flowing through 227 lm hydraulic diameter microchannels. The cavitation is instigated by the inlet orifice. Adiabatic tests were con- ducted

  15. Numerical simulation of boiling during the quenching process N. Kosseifi1

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    by computing the heat flux that arises from the different modes of boiling. They coupled the phase change to heat dissipation of the metal thanks to an interface coupling code. The heat transfer coefficient using a level set method. Surface tension is evaluated using the continuum surface force. The proposed

  16. Nuclear Engineering Nuclear Criticality Safety

    E-Print Network [OSTI]

    Kemner, Ken

    development, Nuclear Operations Division (NOD) waste management and storage activities and other laboratoryNuclear Engineering Nuclear Criticality Safety The Nuclear Engineering Division (NE) of Argonne National Laboratory is experienced in performing criticality safety and shielding evaluations for nuclear

  17. Pressure drop, heat transfer, critical heat flux, and flow stability of two-phase flow boiling of water and ethylene glycol/water mixtures - final report for project "Efficent cooling in engines with nucleate boiling."

    SciTech Connect (OSTI)

    Yu, W.; France, D. M.; Routbort, J. L. (Energy Systems)

    2011-01-19T23:59:59.000Z

    Because of its order-of-magnitude higher heat transfer rates, there is interest in using controllable two-phase nucleate boiling instead of conventional single-phase forced convection in vehicular cooling systems to remove ever increasing heat loads and to eliminate potential hot spots in engines. However, the fundamental understanding of flow boiling mechanisms of a 50/50 ethylene glycol/water mixture under engineering application conditions is still limited. In addition, it is impractical to precisely maintain the volume concentration ratio of the ethylene glycol/water mixture coolant at 50/50. Therefore, any investigation into engine coolant characteristics should include a range of volume concentration ratios around the nominal 50/50 mark. In this study, the forced convective boiling heat transfer of distilled water and ethylene glycol/water mixtures with volume concentration ratios of 40/60, 50/50, and 60/40 in a 2.98-mm-inner-diameter circular tube has been investigated in both the horizontal flow and the vertical flow. The two-phase pressure drop, the forced convective boiling heat transfer coefficient, and the critical heat flux of the test fluids were determined experimentally over a range of the mass flux, the vapor mass quality, and the inlet subcooling through a new boiling data reduction procedure that allowed the analytical calculation of the fluid boiling temperatures along the experimental test section by applying the ideal mixture assumption and the equilibrium assumption along with Raoult's law. Based on the experimental data, predictive methods for the two-phase pressure drop, the forced convective boiling heat transfer coefficient, and the critical heat flux under engine application conditions were developed. The results summarized in this final project report provide the necessary information for designing and implementing nucleate-boiling vehicular cooling systems.

  18. Study of bubble growth in water pool boiling through synchronized, infrared thermometry and high-speed video

    E-Print Network [OSTI]

    Gerardi, Craig

    High-speed video and infrared thermometry were used to obtain time- and space-resolved information on bubble nucleation and heat transfer in pool boiling of water. The bubble departure diameter and frequency, growth and ...

  19. Development of models for the two-dimensional, two-fluid code for sodium boiling NATOF-2D

    E-Print Network [OSTI]

    Zielinski, R. G.

    1981-01-01T23:59:59.000Z

    Several features were incorporated into NATOF-2D, a twodimensional, two fluid code developed at M.I.T. for the purpose of analysis of sodium boiling transients under LMFBR conditions. They include improved interfacial mass, ...

  20. Multi-Scale Experiments in Turbulent Subcooled Boiling Flow Through a Square Channel with a Single Heated Wall

    E-Print Network [OSTI]

    Estrada Perez, Carlos Eduardo

    2014-12-12T23:59:59.000Z

    In this work, visualization experimental techniques that provide whole-field and multi-scale measurements of the liquid turbulence parameters, liquid and heater wall temperatures, and gas phase local parameters, were used to study subcooled boiling...

  1. Development of a general purpose subgrid wall boiling model from improved physical understanding for use in computational fluid dynamics

    E-Print Network [OSTI]

    Gilman, Lindsey Anne

    2014-01-01T23:59:59.000Z

    Advanced modeling capabilities were developed for application to subcooled flow boiling through this work. The target was to introduce, and demonstrate, all necessary mechanisms required to accurately predict the temperature ...

  2. A four-equation two-phase flow model for sodium boiling simulation of LMFBR fuel assemblies

    E-Print Network [OSTI]

    Schor, Andrei L.

    1982-01-01T23:59:59.000Z

    A three-dimensional numerical model for the simulation of sodium boiling transients has been developed. The model uses mixture mass and energy equations, while employing a separate momentum equation for each phase. Thermal ...

  3. Film boiling on the inside of vertical tubes with upward flow of the fluid at low qualities

    E-Print Network [OSTI]

    Dougall, R. S.

    1963-01-01T23:59:59.000Z

    Flow regimes, local heat transfer coefficients, and temperature distributions along the wall have been studied for film boiling inside a vertical tube with upward flow of a saturated liquid. The area of interest has been ...

  4. Subcooled flow boiling heat transfer and critical heat flux in water-based nanofluids at low pressure

    E-Print Network [OSTI]

    Kim, Sung Joong, Ph. D. Massachusetts Institute of Technology

    2009-01-01T23:59:59.000Z

    A nanofluid is a colloidal suspension of nano-scale particles in water, or other base fluids. Previous pool boiling studies have shown that nanofluids can improve the critical heat flux (CHF) by as much as 200%. In this ...

  5. Effect of combined nanoparticle and polymeric dispersions on critical heat flux, nucleate boiling heat transfer coefficient, and coating adhesion

    E-Print Network [OSTI]

    Edwards, Bronwyn K

    2009-01-01T23:59:59.000Z

    An experimental study was performed to determine thermal performance and adhesion effects of a combined nanoparticle and polymeric dispersion coating. The critical heat flux (CHF) values and nucleate boiling heat transfer ...

  6. Experimental investigation of subcooled flow boiling using synchronized high speed video, infrared thermography, and particle image velocimetry

    E-Print Network [OSTI]

    Phillips, Bren Andrew

    2014-01-01T23:59:59.000Z

    Subcooled flow boiling of water was experimentally investigated using high-speed video (HSV), infrared (IR) thermography, and particle image velocimetry (PIV) to generate a unique database of synchronized data. HSV allowed ...

  7. Two dimensional, two fluid model for sodium boiling in LMFBR fuel assemblies

    SciTech Connect (OSTI)

    Granziera, M.R.; Kazimi, M.S.

    1980-05-01T23:59:59.000Z

    A two dimensional numerical model for the simulation of sodium boiling transient was developed using the two fluid set of conservation equations. A semiimplicit numerical differencing scheme capable of handling the problems associated with the ill-posedness implied by the complex characteristic roots of the two fluid problems was used, which took advantage of the dumping effect of the exchange terms. Of particular interest in the development of the model was the identification of the numerical problems caused by the strong disparity between the axial and radial dimensions of fuel assemblies. A solution to this problem was found which uses the particular geometry of fuel assemblies to accelerate the convergence of the iterative technique used in the model. Three sodium boiling experiments were simulated with the model, with good agreement between the experimental results and the model predictions.

  8. Multi-scale Control and Enhancement of Reactor Boiling Heat Flux by Reagents and Nanoparticles

    SciTech Connect (OSTI)

    Manglik, R M; Athavale, A; Kalaikadal, D S; Deodhar, A; Verma, U

    2011-09-02T23:59:59.000Z

    The phenomenological characterization of the use of non-invasive and passive techniques to enhance the boiling heat transfer in water has been carried out in this extended study. It provides fundamental enhanced heat transfer data for nucleate boiling and discusses the associated physics with the aim of addressing future and next-generation reactor thermal-hydraulic management. It essentially addresses the hypothesis that in phase-change processes during boiling, the primary mechanisms can be related to the liquid-vapor interfacial tension and surface wetting at the solidliquid interface. These interfacial characteristics can be significantly altered and decoupled by introducing small quantities of additives in water, such as surface-active polymers, surfactants, and nanoparticles. The changes are fundamentally caused at a molecular-scale by the relative bulk molecular dynamics and adsorption-desorption of the additive at the liquid-vapor interface, and its physisorption and electrokinetics at the liquid-solid interface. At the micro-scale, the transient transport mechanisms at the solid-liquid-vapor interface during nucleation and bubblegrowth can be attributed to thin-film spreading, surface-micro-cavity activation, and micro-layer evaporation. Furthermore at the macro-scale, the heat transport is in turn governed by the bubble growth and distribution, macro-layer heat transfer, bubble dynamics (bubble coalescence, collapse, break-up, and translation), and liquid rheology. Some of these behaviors and processes are measured and characterized in this study, the outcomes of which advance the concomitant fundamental physics, as well as provide insights for developing control strategies for the molecular-scale manipulation of interfacial tension and surface wetting in boiling by means of polymeric reagents, surfactants, and other soluble surface-active additives.

  9. 167Int. J. Nuclear Energy Science and Technology, Vol. 2, No. 3, 2006 Analysis methods for the determination of possible

    E-Print Network [OSTI]

    Demazière, Christophe

    167Int. J. Nuclear Energy Science and Technology, Vol. 2, No. 3, 2006 Analysis methods-dependence due to the localised character of the perturbation. Several techniques relying on the analysis. Keywords: neutron noise analysis; Boiling Water Reactor (BWR); stability; Decay Ratio (DR); space

  10. Enhancement of flow boiling of subcooled water on transverse ribbed surface

    SciTech Connect (OSTI)

    Fujita, Yasunobu; Deguchi, Hironobu

    1999-07-01T23:59:59.000Z

    The present work deals with enhancement of flow boiling heat transfer using transverse ribs of rectangular cross section attached on a flat heating surface. High flux cooling is envisioned in the field of the leading edge technology such as high power laser applications, advanced metallurgical processes, fusion reactors, integrated circuit chips and so forth. In such advanced cooling devices subcooled boiling of water at high velocity is expected as one of most efficient and convenient means for heat removal exceeding 10{sup 7}W/m{sup 2}. In the present experiment, a sheet of stainless-steel (10mm wide, 0.2mm thick and 80mm in heated length) was flush mounted on one wall of a vertical rectangular channel (a cross-section 20mm x 30 mm) and used as a heating surface by passing a direct current. Transverse-ribs made of plastic plate (0.5mm thick and 30mm in lateral length) were attached at an equal longitudinal spacing on the heating surface. Longitudinal spacing of ribs was varied from 2.5, 5, 10, 20mm to infinity (a flat surface without ribs), and the rib height was 2.5mm and 5.0mm. Experiments were conducted with water at a pressure of 0.12MPa in the range of mass velocity from 500kg/m{sup 2}s to 2,000kg/m{sup 2}s (water velocity from 0.5m/s to 2m/s) and subcooling from 20K to 50K. It was found that the ribbed structure strongly affects heat transfer both in the non-boiling and partial nucleate boiling regimes. Enhancement rate of heat transfer coefficient varied from 5% to 50% in excess compared with that for the flat heating plate. According to visual observation of boiling bubbles the circulating flow occurred in a space between each consecutive ribs and seemed to enhance heat transfer. For narrow rib spacing a large coalesced bubble filled up each rib space and impeded the exchange of vapor with liquid, leading to heat transfer deterioration. In the present experimental range of the ribs, a 50% increase of heat transfer coefficient was attained on the ribbed surface with 5mm height and a spacing of 10mm.

  11. Nuclear Resonance Fluorescence for Nuclear Materials Assay

    E-Print Network [OSTI]

    Quiter, Brian Joseph

    2010-01-01T23:59:59.000Z

    Potential of Nuclear Resonance Fluorescence . . . . . . . .2.9.1 Nuclear ThomsonSections . . . . . . . . . . . . . . . Nuclear Resonance

  12. Analysis of the magnetic corrosion product deposits on a boiling water reactor cladding

    SciTech Connect (OSTI)

    Orlov, Andrey [Paul Scherrer Institut, Villigen (Switzerland); Degueldre, Claude, E-mail: claude.degueldre@psi.ch [Paul Scherrer Institut, Villigen (Switzerland); Kaufmann, Wilfried [Kernkraftwerk Leibstadt, Leibstadt (Switzerland)

    2013-01-15T23:59:59.000Z

    The buildup of corrosion product deposits (CRUD) on the fuel cladding of the boiling water reactor (BWR) before and after zinc injection has been investigated by applying local experimental analytical techniques. Under the BWR water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}] spinel solid solutions. X-ray absorption spectroscopy (XAS) revealed inversion ratios of cation distribution in spinels deposited from the solid solution. Based on this information, a two-site ferrite spinel solid solution model is proposed. Electron probe microanalysis (EPMA) and extended X-ray absorption fine structure (EXAFS) findings suggest the zinc-rich ferrite spinels formation on BWR fuel cladding mainly at lower pin. - Graphical Abstract: Analysis of spinels in corrosion product deposits on boiling water reactor fuel rod. Combining EPMA and XAFS results: schematic representation of the ferrite spinels in terms of the end members and their extent of inversion. Note that the ferrites are represented as a surface between the normal (upper plane, M[Fe{sub 2}]O{sub 4}) and the inverse (lower plane, Fe[MFe]O{sub 4}). Actual compositions red Black-Small-Square for the specimen at low elevation (810 mm), blue Black-Small-Square for the specimen at mid elevation (1800 mm). The results have an impact on the properties of the CRUD material. Highlights: Black-Right-Pointing-Pointer Buildup of corrosion product deposits on fuel claddings of a boiling water reactor (BWR) are investigated. Black-Right-Pointing-Pointer Under BWR water conditions, Zn addition with Ni and Mn induced formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}]. Black-Right-Pointing-Pointer X-Ray Adsorption Spectroscopy (XAS) revealed inversion of cations in spinel solid solutions. Black-Right-Pointing-Pointer Zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations.

  13. The use of the probability distribution function to analyze surface temperature fluctuations in pool boiling 

    E-Print Network [OSTI]

    Tu, Chau Qui

    1976-01-01T23:59:59.000Z

    par t of his t ime and labor to assist in the welding of thermocoupl-. junctions to the copper disk. And to all of those directly connected and to many others who helped by givino words of suggestion, there will always remain a debt o.... DEDICATION. AC (QU OWL EDGI"IENT S. TABLE OF CONTENTS. L1ST OF TABLES. LIST OF F IGUR ES. vi vii 1x CHAPTER I INTRODUCTION. CHAPTER II MECHANISMS OF NUCLEATE POOL BOILING. . . . . Bubble agitation model. Vapor-liquid exchange model. 3. M...

  14. Nuclear Physics

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Underground Research Facility in South Dakota, which will search for neutrinoless double-beta decay. Strong Los Alamos programs in nuclear data and nuclear theory supports...

  15. Nuclear weapons and nuclear war

    SciTech Connect (OSTI)

    Cassel, C.; McCally, M.; Abraham, H.

    1984-01-01T23:59:59.000Z

    This book examines the potential radiation hazards and environmental impacts of nuclear weapons. Topics considered include medical responsibility and thermonuclear war, the threat of nuclear war, nuclear weaponry, biological effects, radiation injury, decontamination, long-term effects, ecological effects, psychological aspects, the economic implications of nuclear weapons and war, ethics, civil defense, arms control, nuclear winter, and long-term biological consequences of nuclear war.

  16. Nuclear Matter and Nuclear Dynamics

    E-Print Network [OSTI]

    M Colonna

    2009-02-26T23:59:59.000Z

    Highlights on the recent research activity, carried out by the Italian Community involved in the "Nuclear Matter and Nuclear Dynamics" field, will be presented.

  17. Magnetic thaw-down and boil-off due to magneto acceptors in 2DEG

    SciTech Connect (OSTI)

    Chaubet, C.; Raymond, A. [L2C UMR 5221, CNRS-Université Montpellier 2, Place E. Bataillon, 34090 Montpellier cedex 05 (France); Bisotto, I. [LNCMI, UPR 3228, CNRS-INSA-UJF-UPS, BP166, 38042 Grenoble, Cedex 9 (France); Harmand, J. C. [LPN, CNRS, route de Nozay, 91460 Marcoussis (France); Kubisa, M. [Institute of Physics, Wroclaw University of Technology, 50-370 Wroclaw (Poland); Zawadzki, W. [Institute of Physics, Polish Academy of Sciences, 02668 Warsaw (Poland)

    2013-12-04T23:59:59.000Z

    The Quantum Hall Effect (QHE) and Shubnikov-de Haas effect are investigated experimentally using n type modulation-doped GaAs/GaAlAs quantum wells (QWs) additionally doped in the well with beryllium acceptor atoms. It is presently shown that the localized magneto-acceptor (MA) states which possess discrete energies above the corresponding Landau levels (LLs) lead to two observable effects in magneto-transport: magnetic thaw-down and magnetic boil-off of 2D electrons. Both effects are related to the fact that electrons occupying the localized MA states cannot conduct. Thus in the thaw-down effect the electrons fall down from the MA states to the free Landau states. This leads to a shift of the Hall plateau towards higher magnetic fields as a consequence of an increase of the 2D electron density N{sub S}. In the boil-off effect the electrons are pushed from the free Landau states to the empty MA states under high enough Hall electric field. This process has an avalanche character leading to a dramatic increase of magneto-resistance, consequence of a decrease of N{sub S}.

  18. Pool boiling of R-114/oil mixtures from single tubes and tube bundles. Master's thesis

    SciTech Connect (OSTI)

    Murphy, T.J.

    1987-09-01T23:59:59.000Z

    An apparatus was designed, fabricated, and operated for the testing of horizontal tube bundles for boiling of R-114 with various concentrations of oil. Preliminary data were taken on the top tube in the bundle, with and without the other tubes in operation. Results showed up to a 37% increase in the boiling heat-transfer coefficient as a result of the favorable bundle effect. In a separate single-tube apparatus, three enhanced tubes were tested at a saturation temperature of 2.2 C with oil mass concentrations of 0, 1, 2, 3, 6 and 10%. The tubes were: 1) a finned tube with 1024 fins per meter, 2) a finned tube with 1575 fins per meter and 3) a Turbo-B tube. These tubes resulted in enhancement ratios in pure refrigerant of 2.8, 3.8 and 5.2, respectively, at a practical heat flux of 30 kW/sq. meter. With 3% oil, these ratios were decreased to 2.6, 3.5 and 5, while with 10% oil, these ratios were further reduced to 2.6, 3.2 and 4.7, respectively. Based on these results, the use of Turbo-B tubes is expected to result in significant savings in weight and size of evaporators over the finned tubes presently in use on board some naval vessels.

  19. Performance of Charcoal Cookstoves for Haiti Part 1: Results from the Water Boiling Test

    SciTech Connect (OSTI)

    Booker, Kayje; Han, Tae Won; Granderson, Jessica; Jones, Jennifer; Lsk, Kathleen; Yang, Nina; Gadgil, Ashok

    2011-06-01T23:59:59.000Z

    In April 2010, a team of scientists and engineers from Lawrence Berkeley National Lab (LBNL) and UC Berkeley, with support from the Darfur Stoves Project (DSP), undertook a fact-finding mission to Haiti in order to assess needs and opportunities for cookstove intervention. Based on data collected from informal interviews with Haitians and NGOs, the team, Scott Sadlon, Robert Cheng, and Kayje Booker, identified and recommended stove testing and comparison as a high priority need that could be filled by LBNL. In response to that recommendation, five charcoal stoves were tested at the LBNL stove testing facility using a modified form of version 3 of the Shell Foundation Household Energy Project Water Boiling Test (WBT). The original protocol is available online. Stoves were tested for time to boil, thermal efficiency, specific fuel consumption, and emissions of CO, CO{sub 2}, and the ratio of CO/CO{sub 2}. In addition, Haitian user feedback and field observations over a subset of the stoves were combined with the experiences of the laboratory testing technicians to evaluate the usability of the stoves and their appropriateness for Haitian cooking. The laboratory results from emissions and efficiency testing and conclusions regarding usability of the stoves are presented in this report.

  20. A correlation for local coefficients of heat transfer in boiling of R12 and R22 refrigerants on multirow bundles of smooth tubes

    SciTech Connect (OSTI)

    Rebrov, P.N.; Bukin, V.G.; Danilova, G.N.

    1989-07-01T23:59:59.000Z

    Experimental results on the boiling of R12 and R22 on bundles of 30 to 50 vertically stacked rows of smooth tubes are presented and correlated. Engineering equations for determining the coefficients of heat transfer in the pool boiling of freons on such multirow bundles are derived.

  1. Nuclear Resonance Fluorescence for Nuclear Materials Assay

    E-Print Network [OSTI]

    Quiter, Brian Joseph

    2010-01-01T23:59:59.000Z

    to Journal of Nuclear Technology. [46] C.J. Hagmann and J.Library for Nuclear Science and Technology,” Nuclear Dataof Standards and Technology daughter nuclear data processing

  2. IEEE TRANSACTIONS ON AUTOMATIC CONTROL, VOL. AC-20, NO. 6, DECEMBER 1975 127 Acoustics, Stability, and Compensation in Boiling

    E-Print Network [OSTI]

    Kwatny, Harry G.

    of steam pipe acoustics on the control of steam pressore in boiling waterreactors. A madel is developed forthe process which encompassesthe reactor vessel, the main steam piping, the bypasssystempiping in an increased steaming rate. The turbine admission valves would thenopen in responsetotheconsequent increase

  3. Boiling Water at Hot Creek--The Dangerous and Dynamic Thermal Springs in California's Long Valley Caldera

    E-Print Network [OSTI]

    Torgersen, Christian

    ). Conditions in Hot Creek can change very quickly. These fish--caught in a burst of high-temperature water" or intermittently spurting very hot, sediment-laden water as high as 6 feet (2 m) above the stream surface. At timesBoiling Water at Hot Creek--The Dangerous and Dynamic Thermal Springs in California's Long Valley

  4. Flow boiling of water in a circular staggered micro-pin fin heat sink Santosh Krishnamurthy, Yoav Peles *

    E-Print Network [OSTI]

    Peles, Yoav

    across a bank of heated tube bundles, have shown that the local two-phase heat transfer coefficient across a tube bundle and determined the void fraction, the frictional pressure drop, and the local heatFlow boiling of water in a circular staggered micro-pin fin heat sink Santosh Krishnamurthy, Yoav

  5. Minnesota Nuclear Profile - Monticello

    U.S. Energy Information Administration (EIA) Indexed Site

    date","License expiration date" 1,554,"4,695",96.7,"BWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ,554,"4,695",96.7 "Data for 2010" "BWR Boiling Water Reactor."...

  6. Operational experience with and post-irradiation examinations on boiling water reactor control rods

    SciTech Connect (OSTI)

    Eickelpash, N.; Mullauer, J.; Seepolt, R.W.; Spalthoff, W.

    1983-03-01T23:59:59.000Z

    The control rods of the KRB-I 250-MW (electric) boiling water reactor contain Vipac B/sub 4/C powder in Type 304 stainless steel tubes as a neutron-absorbing material. Because of an increase in the reactor coolant /sup 3/H activity, defective control rods were suspected. The hot cell examination of a highly exposed control rod revealed B/sub 4/C losses. The mechanism of failure was shown to be B/sub 4/C swelling and stress corrosion cracking of the absorber tubes, followed by B/sub 4/C washout. The B/sub 4/C volume swelling is given. The tube cracking starts at 30 to 35% and the B/sub 4/C washout at 50 to 55% local /sup 10/B burnup in the tubes.

  7. Boiling water reactor control rod programming using heuristic and mathematical methods

    SciTech Connect (OSTI)

    Hayase, T.; Motoda, H.

    1980-04-01T23:59:59.000Z

    OPROD, a computer code for automatic generation of control rod programming that has successfully been applied to an older boiling water reactor (BWR), has experienced some difficulties when applied to a BWR of larger power density and stronger heterogeneity. To improve the performance, a heuristic algorithm that is derived from accumulated experience has been introduced to search for a feasible rod pattern that satisfies all constraints. Application of this algorithm to an initial cycle of an 800-MW(e) BWR of high heterogeneity has been very successful. It has been demonstrated that the proposed algorithm is capable of finding a feasible rod pattern, even starting from an all-rods-out pattern. Some improvement was also made in the method of approximation programming (MAP) algorithm. The temporal constraint relaxation method is shown to be effective in finding an optimal control rod pattern in MAP starting from a guess pattern that is not feasible.

  8. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOE Patents [OSTI]

    Hill, P.R.

    1994-12-27T23:59:59.000Z

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  9. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOE Patents [OSTI]

    Hill, Paul R. (Tucson, AZ)

    1994-01-01T23:59:59.000Z

    A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

  10. LIQUID PROPANE GAS (LPG) STORAGE AREA BOILING LIQUID EXPANDING VAPOR EXPLOSION (BLEVE) ANALYSIS

    SciTech Connect (OSTI)

    PACE, M.E.

    2004-01-13T23:59:59.000Z

    The PHA and the FHAs for the SWOC MDSA (HNF-14741) identified multiple accident scenarios in which vehicles powered by flammable gases (e.g., propane), or combustible or flammable liquids (e.g., gasoline, LPG) are involved in accidents that result in an unconfined vapor cloud explosion (UVCE) or in a boiling liquid expanding vapor explosion (BLEVE), respectively. These accident scenarios are binned in the Bridge document as FIR-9 scenarios. They are postulated to occur in any of the MDSA facilities. The LPG storage area will be in the southeast corner of CWC that is relatively remote from store distaged MAR. The location is approximately 30 feet south of MO-289 and 250 feet east of 2401-W by CWC Gate 10 in a large staging area for unused pallets and equipment.

  11. A phenomenological model of the thermal hydraulics of convective boiling during the quenching of hot rod bundles

    SciTech Connect (OSTI)

    Nelson, R.A.; Unal, C.

    1991-01-01T23:59:59.000Z

    In this paper, a phenomenological model of the thermal hydraulics of convective boiling in the post-critical-heat-flux (post-CHF) regime is developed and discussed. The model was implemented in the TRAC-PF1/MOD2 computer code (an advanced best-estimate computer program written for the analysis of pressurized water reactor systems). The model was built around the determination of flow regimes downstream of the quench front. The regimes were determined from the flow-regime map suggested by Ishii and his coworkers. Heat transfer in the transition boiling region was formulated as a position-dependent model. The propagation of the CHF point was strongly dependent on the length of the transition boiling region. Wall-to-fluid film boiling heat transfer was considered to consist of two components: first, a wall-to-vapor convective heat-transfer portion and, second, a wall-to-liquid heat transfer representing near-wall effects. Each contribution was considered separately in each of the inverted annular flow (IAF) regimes. The interfacial heat transfer was also formulated as flow-regime dependent. The interfacial drag coefficient model upstream of the CHF point was considered to be similar to flow through a roughened pipe. A free-stream contribution was calculated using Ishii's bubbly flow model for either fully developed subcooled or saturated nucleate boiling. For the drag in the smooth IAF region, a simple smooth-tube correlation for the interfacial friction factor was used. The drag coefficient for the rough-wavy IAF was formulated in the same way as for the smooth IAF model except that the roughness parameter was assumed to be proportional to liquid droplet diameter entrained from the wavy interface. The drag coefficient in the highly dispersed flow regime considered the combined effects of the liquid droplets within the channel and a liquid film on wet unheated walls. 431 refs., 6 figs., 4 tabs.

  12. Measurement of Key Pool BOiling Parameters in nanofluids for Nuclerar Applications

    SciTech Connect (OSTI)

    Bang, In C [ORNL; Buongiorno, Jdacopo [Massachusetts Institute of Technology (MIT); Hu, Lin-wen [Massachusetts Institute of Technology (MIT); Wang, Hsin [ORNL

    2007-01-01T23:59:59.000Z

    Nanofluids, colloidal dispersions of nanoparticles in a base fluid such as water, can afford very significant Critical Heat Flux (CHF) enhancement. Such engineered fluids potentially could be employed in reactors as advanced coolants in safety systems with significant safety and economic advantages. However, a satisfactory explanation of the CHF enhancement mechanism in nanofluids is lacking. To close this gap, we have identified the important boiling parameters to be measured. These are the properties (e.g., density, viscosity, thermal conductivity, specific heat, vaporization enthalpy, surface tension), hydrodynamic parameters (i.e., bubble size, bubble velocity, departure frequency, hot/dry spot dynamics) and surface conditions (i.e., contact angle, nucleation site density). We have also deployed a pool boiling facility in which many such parameters can be measured. The facility is equipped with a thin indium-tin-oxide heater deposited over a sapphire substrate. An infra-red high-speed camera and an optical probe are used to measure the temperature distribution on the heater and the hydrodynamics above the heater, respectively. The first data generated with this facility already provide some clue on the CHF enhancement mechanism in nanofluids. Specifically, the progression to burnout in a pure fluid (ethanol in this case) is characterized by a smoothly-shaped and steadily-expanding hot spot. By contrast, in the ethanol-based nanofluid the hot spot pulsates and the progression to burnout lasts longer, although the nanofluid CHF is higher than the pure fluid CHF. The presence of a nanoparticle deposition layer on the heater surface seems to enhance wettability and aid hot spot dissipation, thus delaying burnout.

  13. Nuclear Engineer

    Broader source: Energy.gov [DOE]

    This position is located in the Nuclear Safety Division which has specific responsibility for managing the development, analysis, review, and approval of non-reactor nuclear facility safety bases...

  14. Nuclear Navy

    SciTech Connect (OSTI)

    Not Available

    1994-01-01T23:59:59.000Z

    This video tells the story of the Navy's development of nuclear power and its application in long-range submarines and the growing nuclear surface force. Narrated by Frank Blair.

  15. Nuclear Navy

    SciTech Connect (OSTI)

    NONE

    1994-12-31T23:59:59.000Z

    This video tells the story of the Navy`s development of nuclear power and its application in long-range submarines and the growing nuclear surface force. Narrated by Frank Blair.

  16. Nuclear Engineer

    Broader source: Energy.gov [DOE]

    This position is located in the Nuclear Safety Division (NSD) which has specific responsibility for managing the development, analysis, review, and approval of non-reactor nuclear facility safety...

  17. Nuclear-Coupled Flow Instabilities and Their Effects on Dryout

    SciTech Connect (OSTI)

    M. Ishii; X. Sunn; S. Kuran

    2004-09-27T23:59:59.000Z

    Nuclear-coupled flow/power oscillations in boiling water reactors (BWRs) are investigated experimentally and analytically. A detailed literature survey is performed to identify and classify instabilities in two-phase flow systems. The classification and the identification of the leading physical mechanisms of the two-phase flow instabilities are important to propose appropriate analytical models and scaling criteria for simulation. For the purpose of scaling and the analysis of the nonlinear aspects of the coupled flow/power oscillations, an extensive analytical modeling strategy is developed and used to derive both frequency and time domain analysis tools.

  18. Nuclear weapons, nuclear effects, nuclear war

    SciTech Connect (OSTI)

    Bing, G.F.

    1991-08-20T23:59:59.000Z

    This paper provides a brief and mostly non-technical description of the militarily important features of nuclear weapons, of the physical phenomena associated with individual explosions, and of the expected or possible results of the use of many weapons in a nuclear war. Most emphasis is on the effects of so-called ``strategic exchanges.``

  19. Neutron dosimetry at commercial nuclear plants. Final report of Subtask B: dosimeter response

    SciTech Connect (OSTI)

    Cummings, F.M.; Endres, G.W.R.; Brackenbush, L.W.

    1983-03-01T23:59:59.000Z

    As part of a larger program to evaluate personnel neutron dosimetry at commercial nuclear power plants, this study was designed to characterize neutron dosimeter responses inside the containment structure of commercial nuclear plants. In order to characterize those responses, dosimeters were irradiated inside containment at 2 pressurized water reactors and at pipe penetrations outside the biological shield at two boiling water reactors. The reactors were operating at full power during the irradiations. Measurements were also performed with electronic instruments, the tissue equivalent proportional counter (TEPC), and portable remmeters, SNOOPY, RASCAL and PNR-4.

  20. Synchronized High-Speed Video, Infrared Thermometry, and Particle Image Velocimetry Data for Validation of Interface-Tracking Simulations of Nucleate Boiling Phenomena

    E-Print Network [OSTI]

    Duan, Xiaoman

    Nucleation, growth, and detachment of steam bubbles during nucleate boiling of a water pool at atmospheric pressure is experimentally investigated using a combination of synchronized high-speed video, infrared thermography, ...

  1. The effects of orientation angle, subcooling, heat flux, mass flux, and pressure on bubble growth and detachment in subcooled flow boiling

    E-Print Network [OSTI]

    Sugrue, Rosemary M

    2012-01-01T23:59:59.000Z

    The effects of orientation angle, subcooling, heat flux, mass flux, and pressure on bubble growth and detachment in subcooled flow boiling were studied using a high-speed video camera in conjunction with a two-phase flow ...

  2. An investigation of the physical and numerical foundations of two-fluid representation of sodium boiling with applications to LMFBR experiments

    E-Print Network [OSTI]

    No, Hee Cheon

    1983-01-01T23:59:59.000Z

    This work involves the development of physical models for the constitutive relations of a two-fuid, three-dimensional sodium boiling code, THERMIT-6S. The code is equipped with a fluid conduction model, a fuel pin model, ...

  3. Investigation of the pool boiling heat transfer enhancement of nano-engineered fluids by means of high-speed infrared thermography

    E-Print Network [OSTI]

    Gerardi, Craig Douglas

    2009-01-01T23:59:59.000Z

    A high-speed video and infrared thermography based technique has been used to obtain detailed and fundamental time- and space-resolved information on pool boiling heat transfer. The work is enabled by recent advances in ...

  4. 2013 Nuclear Workforce Development ...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Nuclear Workforce Development Day Tuesday, October 22, 2013 Nuclear Medicine Topics: Pathways of Practice in Nuclear Medicine Radiopharmacy Patient Care ...

  5. Nuclear Counterterrorism

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2013-08-26T23:59:59.000Z

    The Order defines requirements for the protection of sensitive improvised nuclear device information and provides a framework to support DOE activities related to nuclear counterterrorism. (A supplemental DOE Manual, Control of and Access to Improvised Nuclear Device Information, provides requirements and procedures for protecting Sigma 20 information.) Appendices A and B are Official Use Only. Point of contact is Adam Boyd (NA-82), 202-586-0010. Cancels DOE O 457.1 and DOE M 457.1-1.

  6. Nuclear Power

    E-Print Network [OSTI]

    Vilhena and Bardo E.J. Bodmann Carbon-#1;? in Terrestrial and Aquatic Environment of Ignalina Nuclear Power Plant: Sources of Production, Releases and Dose Estimates #3;?? Jonas Mazeika Impact of radionuclide discharges from Temel?n Nuclear Power... (chapter 5), ? Instrumentation and control (chapter 6), ? Diagnostics (chapter 7), ? Safety evaluation methods (chapters 6, 8, 9 and 10), ? Environment and nuclear power plants (chapters 11 - 15), ? Human factors (chapter 16), ? Software development...

  7. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    SciTech Connect (OSTI)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01T23:59:59.000Z

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  8. Operational experience with and postirradiation examinations on boiling water reactor control rods

    SciTech Connect (OSTI)

    Eickelpasch, N.; Seepolt, R.W.; Muellauer, J.; Spalthoff, W.

    1983-03-01T23:59:59.000Z

    The control rods of the KRBI-I 250-MW(electric) boiling water reactor contain Vipac B/sub 4/C powder in Type 304 stainless steel tubes as a neutron-absorbing material. Because of an increase in the reactor coolant /sup 3/H activity, defective control rods were suspected. The hot cell examination of a highly exposed control rod revealed B/sub 4/C losses. The mechanism of failure was shown to be B/sub 4/C swelling and stress corrosion cracking of the absorber tubes, followed by B/sub 4/C washout. The B/sub 4/C volume swelling is ..delta..V(%) = 0.851x + 0.0449x/sup 2/ (x = /sup 10/B decays in 10/sup 21/(n,..cap alpha..)/cm/sup 3/). The tube cracking starts at 30 to 35% and the B/sub 4/C washout at 50 to 55% local /sup 10/B burnup in the tubes.

  9. Results of the DF-4 BWR (boiling water reactor) control blade-channel box test

    SciTech Connect (OSTI)

    Gauntt, R.O.; Gasser, R.D.

    1990-10-01T23:59:59.000Z

    The DF-4 in-pile fuel damage experiment investigated the behavior of boiling water reactor (BWR) fuel canisters and control blades in the high temperature environment of an unrecovered reactor accident. This experiment, which was carried out in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories, was performed under the USNRC's internationally sponsored severe fuel damage (SFD) program. The DF-4 test is described herein and results from the experiment are presented. Important findings from the DF-4 test include the low temperature melting of the stainless steel control blade caused by reaction with the B{sub 4}C, and the subsequent low temperature attack of the Zr-4 channel box by the relocating molten blade components. Hydrogen generation was found to continue throughout the experiment, diminishing slightly following the relocation of molten oxidizing zircaloy to the lower extreme of the test bundle. A large blockage which was formed from this material continued to oxidize while steam was being fed into the the test bundle. The results of this test have provided information on the initial stages of core melt progression in BWR geometry involving the heatup and cladding oxidation stages of a severe accident and terminating at the point of melting and relocation of the metallic core components. The information is useful in modeling melt progression in BWR core geometry, and provides engineering insight into the key phenomena controlling these processes. 12 refs., 12 figs.

  10. Thermal-hydraulic instabilities in pressure tube graphite-moderated boiling water reactors

    SciTech Connect (OSTI)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01T23:59:59.000Z

    Thermally induced two-phase instabilities in non-uniformly heated boiling charmers in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to {minus}150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  11. Metallurgical failure analysis of a propane tank boiling liquid expanding vapor explosion (BLEVE).

    SciTech Connect (OSTI)

    Kilgo, Alice C.; Eckelmeyer, Kenneth Hall; Susan, Donald Francis

    2005-01-01T23:59:59.000Z

    A severe fire and explosion occurred at a propane storage yard in Truth or Consequences, N.M., when a truck ran into the pumping and plumbing system beneath a large propane tank. The storage tank emptied when the liquid-phase excess flow valve tore out of the tank. The ensuing fire engulfed several propane delivery trucks, causing one of them to explode. A series of elevated-temperature stress-rupture tears developed along the top of a 9800 L (2600 gal) truck-mounted tank as it was heated by the fire. Unstable fracture then occurred suddenly along the length of the tank and around both end caps, along the girth welds connecting the end caps to the center portion of the tank. The remaining contents of the tank were suddenly released, aerosolized, and combusted, creating a powerful boiling liquid expanding vapor explosion (BLEVE). Based on metallography of the tank pieces, the approximate tank temperature at the onset of the BLEVE was determined. Metallurgical analysis of the ruptured tank also permitted several hypotheses regarding BLEVE mechanisms to be evaluated. Suggestions are made for additional work that could provide improved predictive capabilities regarding BLEVEs and for methods to decrease the susceptibility of propane tanks to BLEVEs.

  12. FORMOSA-B: A Boiling Water Reactor In-Core Fuel Management Optimization Package II

    SciTech Connect (OSTI)

    Karve, Atul A.; Turinsky, Paul J. [North Carolina State University (United States)

    2000-07-15T23:59:59.000Z

    As part of the continuing development of the boiling water reactor in-core fuel management optimization code FORMOSA-B, the fidelity of the core simulator has been improved and a control rod pattern (CRP) sampling capability has been added. The robustness of the core simulator is first demonstrated by benchmarking against core load-follow depletion predictions of both SIMULATE-3 and MICROBURN-B2 codes. The CRP sampling capability, based on heuristic rules, is next successfully tested on a fixed fuel loading pattern (LP) to yield a feasible CRP that removes the thermal margin and critical flow constraint violations. Its performance in facilitating a spectral shift flow operation is also demonstrated, and then its significant influence on the cost of thermal margin is presented. Finally, the heuristic CRP sampling capability is coupled with the stochastic LP optimization capability in FORMOSA-B - based on simulated annealing (SA) - to solve the combined CRP-LP optimization problem. Effectiveness of the sampling in improving the efficiency of the SA adaptive algorithm is shown by comparing the results to those obtained with the sampling turned off (i.e., only LP optimization is carried out for the fixed reference CRP). The results presented clearly indicate the successful implementation of the CRP sampling algorithm and demonstrate FORMOSA-B's enhanced optimization features, which facilitate the code's usage for broader optimization studies.

  13. PROBING DENSE NUCLEAR MATTER VIA NUCLEAR COLLISIONS

    E-Print Network [OSTI]

    Stocker, H.

    2012-01-01T23:59:59.000Z

    University of California. LBL-12095 Probing Dense NuclearMatter Nuclear Collisions* v~a H. Stocker, M.Gyulassy and J. Boguta Nuclear Science Division Lawrence

  14. Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN

    SciTech Connect (OSTI)

    Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci

    2013-09-01T23:59:59.000Z

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

  15. Nuclear fuel elements made from nanophase materials

    DOE Patents [OSTI]

    Heubeck, Norman B. (Schenectady, NY)

    1998-01-01T23:59:59.000Z

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

  16. Nuclear fuel elements made from nanophase materials

    DOE Patents [OSTI]

    Heubeck, N.B.

    1998-09-08T23:59:59.000Z

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

  17. Method and apparatus for steam mixing a nuclear fueled electricity generation system

    DOE Patents [OSTI]

    Tsiklauri, Georgi V. (Richland, WA); Durst, Bruce M. (Kennewick, WA)

    1996-01-01T23:59:59.000Z

    A method and apparatus for improving the efficiency and performance of a nuclear electrical generation system that comprises the addition of steam handling equipment to an existing plant that results in a surprising increase in plant performance. More particularly, a gas turbine electrical generation system with heat recovery boiler is installed along with a micro-jet high pressure and a low pressure mixer superheater. Depending upon plant characteristics, the existing moisture separator reheater (MSR) can be either augmented or done away with. The instant invention enables a reduction in T.sub.hot without a derating of the reactor unit, and improves efficiency of the plant's electrical conversion cycle. Coupled with this advantage is a possible extension of the plant's fuel cycle length due to an increased electrical conversion efficiency. The reduction in T.sub.hot further allows for a surprising extension of steam generator life. An additional advantage is the reduction in erosion/corrosion of secondary system components including turbine blades and diaphragms. The gas turbine generator used in the instant invention can also replace or augment existing peak or emergency power needs. Another benefit of the instant invention is the extension of plant life and the reduction of downtime due to refueling.

  18. Method and apparatus for improving the performance of a nuclear power electrical generation system

    DOE Patents [OSTI]

    Tsiklauri, Georgi V. (Richland, WA); Durst, Bruce M. (Kennewick, WA)

    1995-01-01T23:59:59.000Z

    A method and apparatus for improving the efficiency and performance a of nuclear electrical generation system that comprises the addition of steam handling equipment to an existing plant that results in a surprising increase in plant performance. More particularly, a gas turbine electrical generation system with heat recovery boiler is installed along with a high pressure and a low pressure mixer superheater. Depending upon plant characteristics, the existing moisture separator reheater (MSR) can be either augmented or done away with. The instant invention enables a reduction in T.sub.hot without a derating of the reactor unit, and improves efficiency of the plant's electrical conversion cycle. Coupled with this advantage is a possible extension of the plant's fuel cycle length due to an increased electrical conversion efficiency. The reduction in T.sub.hot further allows for a surprising extension of steam generator life. An additional advantage is the reduction in erosion/corrosion of secondary system components including turbine blades and diaphragms. The gas turbine generator used in the instant invention can also replace or augment existing peak or emergency power needs.

  19. Systematic investigation of the effects of hydrophilic porosity on boiling heat transfer and critical heat flux

    E-Print Network [OSTI]

    Tetreault-Friend, Melanie

    2014-01-01T23:59:59.000Z

    Predicting the conditions of critical heat flux (CHF) is of considerable importance for safety and economic reasons in heat transfer units, such as in nuclear power plants. It is greatly advantageous to increase this thermal ...

  20. Alcohol-free alkoxide process for containing nuclear waste

    DOE Patents [OSTI]

    Pope, James M. (Monroeville, PA); Lahoda, Edward J. (Edgewood, PA)

    1984-01-01T23:59:59.000Z

    Disclosed is a method of containing nuclear waste. A composition is first prepared of about 25 to about 80%, calculated as SiO.sub.2, of a partially hydrolyzed silicon compound, up to about 30%, calculated as metal oxide, of a partially hydrolyzed aluminum or calcium compound, about 5 to about 20%, calculated as metal oxide, of a partially hydrolyzed boron or calcium compound, about 3 to about 25%, calculated as metal oxide, of a partially hydrolyzed sodium, potassium or lithium compound, an alcohol in a weight ratio to hydrolyzed alkoxide of about 1.5 to about 3% and sufficient water to remove at least 99% of the alcohol as an azeotrope. The azeotrope is boiled off and up to about 40%, based on solids in the product, of the nuclear waste, is mixed into the composition. The mixture is evaporated to about 25 to about 45% solids and is melted and cooled.

  1. National Nuclear Security Administration | National Nuclear Security...

    National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing...

  2. Nuclear war, nuclear proliferation, and their consequences

    SciTech Connect (OSTI)

    Aga Khan, S.

    1986-01-01T23:59:59.000Z

    This book presents papers on nuclear weapons proliferation. Topics considered include the Nuclear Non-Proliferation Treaty and its future, the spread of nuclear weapons among nations, the link between horizontal and vertical proliferation, national security, nuclear disarmament, the impact of nuclear weapons on Third World regional conflicts, the global effects of a nuclear war, medical effects on human populations, the nuclear winter, the nuclear arms race, and arms control.

  3. NUCLEAR PLANT OPERATIONS AND

    E-Print Network [OSTI]

    Pázsit, Imre

    NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: neutron flux, cur- rent noise, vibration diagnostics: Swedish Nuclear Powe

  4. Nuclear Counterterrorism

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2006-02-07T23:59:59.000Z

    The Order defines requirements for the protection of sensitive improvised nuclear device information and provides a framework to support DOE activities related to nuclear counterterrorism. (A supplemental DOE Manual, Control of and Access to Improvised Nuclear Device Information, provides requirements and procedures for protecting Sigma 20 information. The Manual is Official Use Only, and is not available on the Directives Portal. The point of contact for the Manual is Randall Weidman, NA-121.2, 202-586-4582.) Canceled by DOE O 457.1A

  5. Factors influencing the precoat filtration of boiling water reactor water streams

    SciTech Connect (OSTI)

    Hermansson, H.P. (Studsvik Material AB, Nykoeping (Sweden)); Persson, G. (Forsmark Nuclear Power Plant, Oesthammar (Sweden)); Reinvall, A. (Industrial Power Co. Ltd., Olkiluoto (Finland))

    1994-10-01T23:59:59.000Z

    A series of studies on precoat filtration were carried out on condensate and preheater drains in the Swedish and Finnish boiling water reactors (BWRs). The goal was to increase knowledge about the precoat filtration process and to find physical and chemical means to improve the performance of the precoat filters in the condensate polishing plants. To achieve this goal a number of parameters, such as type of resin, bed depth, pH, oxygen and organic contaminant concentrations (measured total organic carbon), and corrosion product particle characteristics, were selected for the study. The work was mainly carried out in the power plants using an experimental facility fed with on-line sampled condensates and drains taken from the plant sampling lines. The main results are that there is a varying influence on precoat filtration from all the aforementioned parameters. The oxygen concentration, the concentration of organic contaminants, and the type of corrosion products are, however, the factors that have the strongest influence within the parameter ranges that are representative for BWR operation. The results are rather similar when the different units are compared. There are, however, some differences that could be mainly attributed to deviations in operation parameters and the subsequent differences in the corrosion product spectra. The mechanism for precoat filtration of corrosion products in BWR condensate is complex. The filtration behavior is to a large extent governed by competition between depth filtration and electrostatic interactions. During the early stages of the filtration cycle, electrostatic interaction is of great importance, whereas depth filtration becomes more important with increasing operating time. Rapid pressure drop buildup rates have been demonstrated to be caused by the presence of amorphous corrosion products. An effect from the presence of organic contaminants has been found, although this should be of little significance.

  6. Nuclear Celebrations

    E-Print Network [OSTI]

    Hacker, Randi; Tsutsui, William

    2006-11-01T23:59:59.000Z

    Broadcast Transcript: The North Korean situation is frightening for many reasons but none, perhaps, more eerily disturbing than images of North Koreans celebrating in brightly colored costumes just days after the nation's underground nuclear test...

  7. Nuclear Nonproliferation

    SciTech Connect (OSTI)

    Atkins-Duffin, C E

    2008-12-10T23:59:59.000Z

    With an explosion equivalent of about 20kT of TNT, the Trinity test was the first demonstration of a nuclear weapon. Conducted on July 16, 1945 in Alamogordo, NM this site is now a Registered National Historic Landmark. The concept and applicability of nuclear power was demonstrated on December 20, 1951 with the Experimental Breeder Reactor Number One (EBR-1) lit four light bulbs. This reactor is now a Registered National Historic Landmark, located near Arco, ID. From that moment forward it had been clearly demonstrated that nuclear energy has both peaceful and military applications and that the civilian and military fuel cycles can overlap. For the more than fifty years since the Atoms for Peace program, a key objective of nuclear policy has been to enable the wider peaceful use of nuclear energy while preventing the spread of nuclear weapons. Volumes have been written on the impact of these two actions on the world by advocates and critics; pundits and practioners; politicians and technologists. The nations of the world have woven together a delicate balance of treaties, agreements, frameworks and handshakes that are representative of the timeframe in which they were constructed and how they have evolved in time. Collectively these vehicles attempt to keep political will, nuclear materials and technology in check. This paper captures only the briefest abstract of the more significant aspects on the Nonproliferation Regime. Of particular relevance to this discussion is the special nonproliferation sensitivity associated with the uranium isotope separation and spent fuel reprocessing aspects of the nuclear fuel cycle.

  8. Nuclear Science/Nuclear Chemistry

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItemResearch > TheNuclear Astrophysics One of the greatNuclear Science/Nuclear

  9. Aquatic impacts from operation of three midwestern nuclear power stations: Cooper Nuclear Station environmental appraisal report

    SciTech Connect (OSTI)

    Brice, J.R.

    1981-10-01T23:59:59.000Z

    Cooper Nuclear Station is located on the Nebraska side of the Missouri River in Nemaha County. The station utilizes a boiling water reactor and steam turbine generator to produce 778 MW (net) of electrical power. The cooling system is a once-through design that withdraws water from, and discharges to, the Missouri River. No significant adverse impacts to the biota of the Missouri River from the Cooper Nuclear Station discharge were detected. Localized effects in the vicinity of the discharge have been observed. These include changes in the diversity and productivity of phytoplankton, periphyton, and benthic invertebrates at certain times of the year. The station appears to entrain large numbers of catostomid larvae, but this loss is not reflected in the available commercial fisheries statistics. Large numbers of gizzard shad and freshwater drum are impinged annually by Cooper Nuclear Station, but neither of these species seem to be adversely affected. Bigmouth buffalo populations could potentially suffer losses, but as was the case with the other catostomids, commercial catches of bigmouth buffalo did not seem to be affected by station operation.

  10. NUCLEAR PROXIMITY FORCES

    E-Print Network [OSTI]

    Randrup, J.

    2011-01-01T23:59:59.000Z

    One might summarize of nuclear potential energy has beendegree of freedom) for the nuclear interaction between anyUniversity of California. Nuclear Proximity Forces 'I< at

  11. Nuclear scales

    SciTech Connect (OSTI)

    Friar, J.L.

    1998-12-01T23:59:59.000Z

    Nuclear scales are discussed from the nuclear physics viewpoint. The conventional nuclear potential is characterized as a black box that interpolates nucleon-nucleon (NN) data, while being constrained by the best possible theoretical input. The latter consists of the longer-range parts of the NN force (e.g., OPEP, TPEP, the {pi}-{gamma} force), which can be calculated using chiral perturbation theory and gauged using modern phase-shift analyses. The shorter-range parts of the force are effectively parameterized by moments of the interaction that are independent of the details of the force model, in analogy to chiral perturbation theory. Results of GFMC calculations in light nuclei are interpreted in terms of fundamental scales, which are in good agreement with expectations from chiral effective field theories. Problems with spin-orbit-type observables are noted.

  12. Nuclear winter

    SciTech Connect (OSTI)

    Ehrlich, A.

    1984-04-01T23:59:59.000Z

    The 13 speakers at the October 1983 Conference on the World After Nuclear War each contributed specialized knowledge to the climatic and biological effects of nuclear war. The author highlights the findings of the TTAPS (named for its authors) study and confirmation by Soviet scientists on the nuclear winter. Atmospheric consequences would come from debris blocking sunlight and creating conditions of cold and darkness that could preclude the continued existence of life. The biological consequences of cold and darkness would be reduced photosynthesis, devastating losses of food, damage and death from ionizing radiation, and a breakdown of ecosystems. Impacts on the human population would be intensified by a breakdown in social services. The author summarizes points of discussion during the conference. 4 references.

  13. Mississippi Nuclear Profile - Grand Gulf

    U.S. Energy Information Administration (EIA) Indexed Site

    expiration date" 1,"1,251","9,643",88.0,"BWR","applicationvnd.ms-excel","applicationvnd.ms-excel" ,"1,251","9,643",88.0 "Data for 2010" "BWR Boiling Water Reactor."...

  14. Nuclear Physics

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOERecovery NuclearSensor NodesNuclear

  15. Dispersed-flow film boiling in rod-bundle geometry: steady-state heat-transfer data and correlation comparisons. [PWR; BWR

    SciTech Connect (OSTI)

    Yoder, G. L.; Morris, D. G.; Mullins, C. B.; Ott, L. J.; Reed, D. A.

    1982-03-01T23:59:59.000Z

    Assessment of six film boiling correlations and one single-phase vapor correlation has been made using data from 22 steady state upflow rod bundle tests (series 3.07.9). Bundle fluid conditions were calculated using energy and mass conservation considerations. Results of the steady state film boiling tests support the conclusions reached in the analysis of prior transient tests 3.03.6AR, 3.06.6B, and 3.08.6C. Comparisons between experimentally determined and correlation-predicted heat transfer coefficients, are presented.

  16. 0-7803-XXXX-X/06/$20.00 2009 IEEE 25th IEEE SEMI-THERM Symposium Sub-Atmospheric Pressure Pool Boiling of Water on a Screen-Laminate Enhanced Surface

    E-Print Network [OSTI]

    Wirtz, Richard A.

    structures having wide ranging porosity and pore size. When deployed as a surface enhancement in a boiling pool-boiling experiments at one atmosphere and sub-atmospheric pressure assess the utility of fine factor of lamination [dimensionless] CHF = critical heat flux [W/cm2 ] Dh = pore hydraulic diameter [µm

  17. Nuclear Golf

    E-Print Network [OSTI]

    Hacker, Randi; Tsutsui, William

    2006-12-06T23:59:59.000Z

    Broadcast Transcript: Pay no attention to that nuclear warhead behind the 18th hole; just shout "Fore!" and drive your Titleist down the fairway. In a development that is bizarre even by North Korean standards, the country is making a move to sell...

  18. Proceedings of EPRI/DOE workshop on nuclear industry valve problems

    SciTech Connect (OSTI)

    Sprung, J.L. (ed.) [ed.

    1981-01-01T23:59:59.000Z

    Representatives from 29 nuclear industry organizations (11 valve manufacturers, 4 nuclear steam supply system vendors, 5 utilities, 3 national laboratories, 2 architect/engineering firms, the Department of Energy (DOE), EPRI, and 2 others) attended the workshop. Working sessions on key valves and on valve stem and seat leakage developed the following recommendations: (1) establish a small permanent expert staff to collect, analyze, and disseminate information about nuclear valve problems; (2) perform generic key valve programs for pressurized water reactors and for boiling water reactors, and several plant specific key valve programs, the latter to demonstrate the cost-effectiveness of such studies; (3) confirm the identity of, define, and initiate needed longer term research and development programs dealing with seat and stem leakage; and (4) establish an industry working group to review and advise on these efforts. Separate abstracts were prepared for three papers which are included in the appendix. (DLC)

  19. Nuclear Forensics | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Forensics | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy...

  20. Nuclear Incident Team | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Incident Team | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear...

  1. Integrated Boiler, Superheater & Decomposer Bayonet for Hydrogen Production

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville PowerCherries 82981-1cnHigh SchoolIn Other NewsSpin andInterim Data Changes in the2011 |- Energy

  2. WORKSHOP ON NUCLEAR DYNAMICS

    E-Print Network [OSTI]

    Myers, W.D.

    2010-01-01T23:59:59.000Z

    L. Wilets, "Theories of Nuclear Fission", Clarendon Press,of the nuclear force, result in lower calculated fission

  3. National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    FROM: SUBJECT: USIUK Memorandum of Understanding between National Nuclear Security Administration's (NNSA) Associate Administrator for Defense Nuclear Security (AADNS)...

  4. INSTRUCTIONS FOR SUBMITTING NUCLEAR

    E-Print Network [OSTI]

    waste management proceedings. Keywords Nuclear, nuclear power plant, spent fuel, nuclear waste, data of Submitted Data 3 NUCLEAR POWER PLANT DATA REQUESTS 6 A. Environmental Impacts 6 B. Spent Fuel Generation 8 C. Spent Nuclear Fuel Storage 9 D. Spent Nuclear Fuel Transport and Disposal Issues 10 E. Interim Spent

  5. PII S0016-7037(99)00441-X Sub sea floor boiling of Red Sea Brines: New indication from noble gas data

    E-Print Network [OSTI]

    Winckler, Gisela

    PII S0016-7037(99)00441-X Sub sea floor boiling of Red Sea Brines: New indication from noble gas in revised form December 8, 1999) Abstract--Hydrothermal brines from the Atlantis II Deep, Red Sea, have been depressions filled by highly saline brines (Hartmann et al., 1998a). The Atlantis II Deep, located

  6. Flow boiling enhancement on a horizontal heater using carbon nanotube coatings N. Singh, V. Sathyamurthy, W. Peterson, J. Arendt, D. Banerjee *

    E-Print Network [OSTI]

    Banerjee, Debjyoti

    with embedded cartridge heaters and K-type thermocouples. De-ion- ized (DI) water was used as the test fluidFlow boiling enhancement on a horizontal heater using carbon nanotube coatings N. Singh, V flux on a horizontal heater that is heated from below. The horizontal heater consists of either a bare

  7. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    SciTech Connect (OSTI)

    NONE

    1993-09-15T23:59:59.000Z

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  8. Heat transfer enhancement in single-phase forced convection with blockages and in two-phase pool boiling with nano-structured surfaces 

    E-Print Network [OSTI]

    Ahn, Hee Seok

    2007-09-17T23:59:59.000Z

    boiling on nano-structured surfaces. Experiments were conducted with three horizontal silicon surfaces, two of which were coated with vertically aligned multi-walled carbon nanotubes (MWCNT) with heights of 9 and 25 �¹m, respectively, and diameters...

  9. Nuclear Energy Research Initiative. Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants. Annual Report

    SciTech Connect (OSTI)

    Ritterbusch, S.E.

    2000-08-01T23:59:59.000Z

    The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-informed approach for the design and regulation of nuclear power plants. This approach will include the development and.lor confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRs) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go farther by focusing on the design of new plants.

  10. Nuclear photonics

    SciTech Connect (OSTI)

    Habs, D.; Guenther, M. M.; Jentschel, M.; Thirolf, P. G. [Ludwig-Maximilians-Universitaet Muenchen, D-85748 Garching (Germany); Max Planck Institut fuer Quantenoptik, D-85748 Garching (Germany); Institut Laue-Langevin, F-38042 Grenoble (Germany); Ludwig-Maximilians-Universitaet Muenchen, D-85748 Garching (Germany)

    2012-07-09T23:59:59.000Z

    With the planned new {gamma}-beam facilities like MEGa-ray at LLNL (USA) or ELI-NP at Bucharest (Romania) with 10{sup 13}{gamma}/s and a band width of {Delta}E{gamma}/E{gamma} Almost-Equal-To 10{sup -3}, a new era of {gamma} beams with energies up to 20MeV comes into operation, compared to the present world-leading HI{gamma}S facility at Duke University (USA) with 10{sup 8}{gamma}/s and {Delta}E{gamma}/E{gamma} Almost-Equal-To 3 Dot-Operator 10{sup -2}. In the long run even a seeded quantum FEL for {gamma} beams may become possible, with much higher brilliance and spectral flux. At the same time new exciting possibilities open up for focused {gamma} beams. Here we describe a new experiment at the {gamma} beam of the ILL reactor (Grenoble, France), where we observed for the first time that the index of refraction for {gamma} beams is determined by virtual pair creation. Using a combination of refractive and reflective optics, efficient monochromators for {gamma} beams are being developed. Thus, we have to optimize the total system: the {gamma}-beam facility, the {gamma}-beam optics and {gamma} detectors. We can trade {gamma} intensity for band width, going down to {Delta}E{gamma}/E{gamma} Almost-Equal-To 10{sup -6} and address individual nuclear levels. The term 'nuclear photonics' stresses the importance of nuclear applications. We can address with {gamma}-beams individual nuclear isotopes and not just elements like with X-ray beams. Compared to X rays, {gamma} beams can penetrate much deeper into big samples like radioactive waste barrels, motors or batteries. We can perform tomography and microscopy studies by focusing down to {mu}m resolution using Nuclear Resonance Fluorescence (NRF) for detection with eV resolution and high spatial resolution at the same time. We discuss the dominating M1 and E1 excitations like the scissors mode, two-phonon quadrupole octupole excitations, pygmy dipole excitations or giant dipole excitations under the new facet of applications. We find many new applications in biomedicine, green energy, radioactive waste management or homeland security. Also more brilliant secondary beams of neutrons and positrons can be produced.

  11. Results and phenomena observed from the DF-4 BWR (boiling water reactor) control blade-channel box test

    SciTech Connect (OSTI)

    Gauntt, R.O.; Gasser, R.D.; Fryer, C.P.; Walker, J.V.

    1988-01-01T23:59:59.000Z

    The DF-4 in-pile fuel damage experiment, one of a series of tests performed as part of the USNRC's internationally sponsored severe fuel damage (SFD) program, and carried out at Sandia National Laboratories, addressed the behavior of boiling water reactor (BWR) fuel canisters and control blades in the high temperature environment of an unrecovered loss of coolant accident. The DF-4 test is described in some detail herein and results from the experiment are presented. Significant results from prior experiments in the series are also summarized. Important findings from the DF-4 test include the low temperature melting of the stainless steel control blade caused by reaction with the B/sub 4/C, and the subsequent low temperature attack of the Zr-4 channel box by the relocating molten blade components. This early melt and relocation phenomena could significantly influence the progression of meltdown in BWRs and should be considered in SFD accident analyses. 14 refs., 9 figs.

  12. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    SciTech Connect (OSTI)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08T23:59:59.000Z

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  13. New flow boiling heat transfer correlation for CFC-114 and C{sub 4}F{sub 10}

    SciTech Connect (OSTI)

    Szady, A.J. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.

    1995-08-01T23:59:59.000Z

    The flow boiling tests being conducted at Cudo Technologies, Ltd. in Lexington, Kentucky are a part of The GDP Coolant Replacement Project. The tests are to be done with two alternates, C{sub 4}F{sub 10} and C{sub 4}F{sub 8}, as well as CFC-114. So far, tests were conducted with CFC-114 and C{sub 4}F{sub 10}. The CFC-114 data by Cudo shows better heat transfer performance than that predicted by the codes used in the numerical model which were based on a superposition model. The data was applied to an asymptotic model developed by Steiner and Taborek. The new correlation developed seems to fit better with the Cudo data as well as the Paducah cell test data. The model will be further investigated when C{sub 4}F{sub 8} data is available.

  14. Collecting and recirculating condensate in a nuclear reactor containment

    DOE Patents [OSTI]

    Schultz, Terry L. (Murrysville Boro, PA)

    1993-01-01T23:59:59.000Z

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

  15. Collecting and recirculating condensate in a nuclear reactor containment

    DOE Patents [OSTI]

    Schultz, T.L.

    1993-10-19T23:59:59.000Z

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures.

  16. Heat transfer characteristics of R410A-oil mixture flow boiling inside a 7 mm straight smooth tube

    SciTech Connect (OSTI)

    Hu, Haitao; Ding, Guoliang; Wei, Wenjian; Wang, Zhence [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Wang, Kaijian [Fujitsu General Institute of Air-Conditioning Technology Limited, Kawasaki 213-8502 (Japan)

    2008-01-15T23:59:59.000Z

    Two-phase flow patterns and heat transfer characteristics of R410A-oil mixture flow boiling inside a straight smooth tube with the outside diameter of 7.0 mm were investigated experimentally. The experimental conditions include the evaporation temperature of 5 C, the mass flux from 200 to 400 kg m{sup -2} s{sup -1}, the heat flux from 7.56 to 15.12 kW m{sup -2}, the inlet vapor quality from 0.2 to 0.7, nominal oil concentration from 0% to 5%. The test results show that the heat transfer coefficient of R410A-oil mixture increases with mass flux of refrigerant-oil mixture; the presence of oil enhances the heat transfer at the range of low and intermediate vapor qualities; there is a peak of local heat transfer coefficient at about 2-4% nominal oil concentration at higher vapor qualities, and the peak shifts to lower nominal oil concentration with the increasing of vapor qualities; higher nominal oil concentration gives more detrimental effect at high vapor qualities. The flow pattern map of R410A-oil mixture was developed based on refrigerant-oil mixture properties, and the observed flow patterns match well with the flow pattern map. New correlation to predict the local heat transfer of R410A-oil mixture flow boiling inside the straight smooth tube was developed based on flow patterns and local properties of refrigerant-oil mixture, and it agrees with 90% of the experiment data within the deviation of {+-}25%. (author)

  17. Security and Use Control of Nuclear Explosives and Nuclear Weapons...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    4C, Security and Use Control of Nuclear Explosives and Nuclear Weapons by LtCol Karl Basham Functional areas: Nuclear Explosives, Nuclear Weapons, Security The Order establishes...

  18. Nuclear Astrophysics

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItemResearch > TheNuclear Astrophysics One of the great scientific challenges is

  19. Nuclear Forensics

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItemResearch > TheNuclear Astrophysics One of the great scientific

  20. Nuclear Energy!

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOE Directives,838Nuclear Detectionmore

  1. Nuclear Nonproliferation,

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOERecovery NuclearSensor Nodes for

  2. NUCLEAR ENERGY

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently Asked QuestionsDepartment of Energyof Energy NREL:Education &NTSF NUCLEAR

  3. Nuclear reactor engineering

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1981-01-01T23:59:59.000Z

    Chapters are presented concerning energy from nuclear fission; nuclear reactions and radiations; diffusion and slowing-down of neutrons; principles of reactor analysis; nuclear reactor kinetics and control; energy removal; non-fuel reactor materials; the reactor fuel system; radiation protection and environmental effects; nuclear reactor shielding; nuclear reactor safety; and power reactor systems.

  4. NUCLEAR DEFORMATION ENERGIES

    E-Print Network [OSTI]

    Blocki, J.

    2009-01-01T23:59:59.000Z

    J.R. Nix, Theory of Nuclear Fission and Superheavy Nuclei,energy maps relevant for nuclear fission and nucleus-nucleusin connection with nuclear fission. The need for a better

  5. Office of Nuclear Safety

    Broader source: Energy.gov [DOE]

    The Office of Nuclear Safety establishes nuclear safety requirements and expectations for the Department to ensure protection of workers and the public from the hazards associated with nuclear operations with all Department operations.

  6. Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2015-01-26T23:59:59.000Z

    This Department of Energy (DOE) Order establishes requirements to implement the nuclear explosive safety (NES) elements of DOE O 452.1E, Nuclear Explosive and Weapon Surety Program, or successor directive, for routine and planned nuclear explosive operations (NEOs).

  7. Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2014-07-10T23:59:59.000Z

    The Order establishes requirements to implement the nuclear explosive safety (NES) elements of DOE O 452.1E, Nuclear Explosive and Weapon Surety Program, for routine and planned nuclear explosive operations (NEOs).

  8. NUCLEAR STRUCTURE DATABASE

    E-Print Network [OSTI]

    Firestone, R.B.

    2010-01-01T23:59:59.000Z

    CALIFORNIA NUCLEAR STRUCTURE DATABASE R. B. Firestone and E.11089 NUCLEAR STRUCTURE DATABASE by R.B. Firestone and E.iii- NUCLEAR STRUCTURE DATABASE R.B Firestone and E. Browne

  9. RELATIVISTIC NUCLEAR COLLISIONS: THEORY

    E-Print Network [OSTI]

    Gyulassy, M.

    2010-01-01T23:59:59.000Z

    Effects in Relativistic Nuclear Collisions", Preprint LBL-Pion Interferometry of Nuclear Collisions. 18.1 M.Gyulassy,was supported by the Office of Nuclear Physics of the U.S.

  10. Nuclear Safety (Pennsylvania)

    Broader source: Energy.gov [DOE]

    The Nuclear Safety Division conducts a comprehensive nuclear power plant oversight review program of the nine reactors at the five nuclear power sites in Pennsylvania. It also monitors the...

  11. Electrochemistry of Water-Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08T23:59:59.000Z

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  12. Nuclear Fuel Cycle & Vulnerabilities

    SciTech Connect (OSTI)

    Boyer, Brian D. [Los Alamos National Laboratory

    2012-06-18T23:59:59.000Z

    The objective of safeguards is the timely detection of diversion of significant quantities of nuclear material from peaceful nuclear activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection. The safeguards system should be designed to provide credible assurances that there has been no diversion of declared nuclear material and no undeclared nuclear material and activities.

  13. Nuclear Physics: Campaigns

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Free-Electron Laser (FEL) Medical Imaging Physics Topics Campaigns The Structure of the Nuclear Building Blocks The Structure of Nuclei Symmetry Tests in Nuclear Physics Meetings...

  14. Nuclear Power Overview

    Broader source: Energy.gov (indexed) [DOE]

    San Onofre Nuclear Generating Station San Onofre Nuclear Generating Station Bob Ashe-Everest Southern California Edison 10 Incoming New Fuel Inspecting New Fuel SONGS Unit 1 Fuel...

  15. Nuclear weapons modernizations

    SciTech Connect (OSTI)

    Kristensen, Hans M. [Federation of American Scientists, Washington, DC (United States)

    2014-05-09T23:59:59.000Z

    This article reviews the nuclear weapons modernization programs underway in the world's nine nuclear weapons states. It concludes that despite significant reductions in overall weapons inventories since the end of the Cold War, the pace of reductions is slowing - four of the nuclear weapons states are even increasing their arsenals, and all the nuclear weapons states are busy modernizing their remaining arsenals in what appears to be a dynamic and counterproductive nuclear competition. The author questions whether perpetual modernization combined with no specific plan for the elimination of nuclear weapons is consistent with the nuclear Non-Proliferation Treaty and concludes that new limits on nuclear modernizations are needed.

  16. Advancing Global Nuclear Security

    Broader source: Energy.gov [DOE]

    Today world leaders gathered at The Hague for the Nuclear Security Summit, a meeting to measure progress and take action to secure sensitive nuclear materials.

  17. Nuclear / Radiological Advisory Team | National Nuclear Security...

    National Nuclear Security Administration (NNSA)

    Radiological Advisory Team | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering...

  18. Nuclear Security Summit | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

  19. Nuclear Safeguards | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

  20. FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF

    SciTech Connect (OSTI)

    Brady Raap, Michaele C.; Lyons, Jennifer A.; Collins, Brian A.; Livingston, James V.

    2014-03-31T23:59:59.000Z

    This report documents the FY13 efforts to enhance a dataset of spent nuclear fuel isotopic composition data for use in developing intrinsic signatures for nuclear forensics. A review and collection of data from the open literature was performed in FY10. In FY11, the Spent Fuel COMPOsition (SFCOMPO) excel-based dataset for nuclear forensics (NF), SFCOMPO/NF was established and measured data for graphite production reactors, Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) were added to the dataset and expanded to include a consistent set of data simulated by calculations. A test was performed to determine whether the SFCOMPO/NF dataset will be useful for the analysis and identification of reactor types from isotopic ratios observed in interdicted samples.

  1. The Joys of Nuclear Engineering

    SciTech Connect (OSTI)

    Jon Carmack

    2009-10-02T23:59:59.000Z

    Nuclear fuels researcher Jon Carmack talks about the satisfactions of a career in nuclear engineering.

  2. The Joys of Nuclear Engineering

    ScienceCinema (OSTI)

    Jon Carmack

    2010-01-08T23:59:59.000Z

    Nuclear fuels researcher Jon Carmack talks about the satisfactions of a career in nuclear engineering.

  3. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    SciTech Connect (OSTI)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01T23:59:59.000Z

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE.

  4. ANNOUNCEMENT NUCLEAR ENGINEERING FACULTY POSITION

    E-Print Network [OSTI]

    Tennessee, University of

    ANNOUNCEMENT NUCLEAR ENGINEERING FACULTY POSITION The Department of Nuclear Engineering at the Assistant or Associate Professor level. These areas include, but are not limited to, nuclear system instrumentation & controls, monitoring and diagnostics, reactor dynamics, nuclear security, nuclear materials

  5. Standard test method for evaluating stress-corrosion cracking of stainless alloys with different nickel content in boiling acidified sodium chloride solution

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2000-01-01T23:59:59.000Z

    1.1 This test method describes a procedure for conducting stress-corrosion cracking tests in an acidified boiling sodium chloride solution. This test method is performed in 25% (by mass ) sodium chloride acidified to pH 1.5 with phosphoric acid. This test method is concerned primarily with the test solution and glassware, although a specific style of U-bend test specimen is suggested. 1.2 This test method is designed to provide better correlation with chemical process industry experience for stainless steels than the more severe boiling magnesium chloride test of Practice G36. Some stainless steels which have provided satisfactory service in many environments readily crack in Practice G36, but have not cracked during interlaboratory testing using this sodium chloride test method. 1.3 This boiling sodium chloride test method was used in an interlaboratory test program to evaluate wrought stainless steels, including duplex (ferrite-austenite) stainless and an alloy with up to about 33% nickel. It may also b...

  6. Load follow simulation of three-dimensional boiling water reactor core by PACS-32 parallel microprocessor system

    SciTech Connect (OSTI)

    Hoshino, T.; Shirakawa, T.

    1982-03-01T23:59:59.000Z

    The three-dimensional boiling water reactor (BWR) core following the daily load was simulated by the use of the processor array for continuum simulation (PACS-32), a newly developed parallel microprocessor system. The PACS system consists of 32 processing units (PUs) (microprocessors) and has a multiinstruction, multidata type architecture, being optimum to the numerical simulation of the partial differential equations. The BWR core model includes the modified twogroup finite difference, coarse-mesh model for neutronics, steady-state model for thermohydraulics, criticality control by core coolant flow, and the time-dependent solution of iodine-xenon dynamics with constant flux level. The analysis of the parallel processing program revealed that the overhead is independent from the number of PUs and that the efficiency of PUs, i.e., the ratio of effective calculation over total, amounts to 75%, even up to 90% if it is limited to the core part. Simulation was made on the daily load follow for 144 h including the weekend, which took 1.3 central processing unit hours by the PACS system. The PACS system demonstrated a computation speed nearly one-tenth that of the large-scale high-speed general purpose computer, with a 25 times better cost performance ratio and showed that the system could be used as the pratical BWR core simulator with more complicated core models.

  7. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    SciTech Connect (OSTI)

    Not Available

    1993-05-13T23:59:59.000Z

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  8. Simulation of in-core neutron noise measurements for axial void profile reconstruction in boiling water reactors

    SciTech Connect (OSTI)

    Dykin, V.; Pazsit, I. [Chalmers Univ. of Technology, Div. of Nuclear Engineering, Dept. of Applied Physics, SE-412 96 Gothenburg (Sweden)

    2012-07-01T23:59:59.000Z

    A possibility to reconstruct the axial void profile from the simulated in-core neutron noise which is caused by density fluctuations in a Boiling Water Reactor (BWR) heated channel is considered. For this purpose, a self-contained model of the two-phase flow regime is constructed which has quantitatively and qualitatively similar properties to those observed in real BWRs. The model is subsequently used to simulate the signals of neutron detectors induced by the corresponding perturbations in the flow density. The bubbles are generated randomly in both space and time using Monte-Carlo techniques. The axial distribution of the bubble production is chosen such that the mean axial void fraction and void velocity follow the actual values of BWRs. The induced neutron noise signals are calculated and then processed by the standard signal analysis methods such as Auto-Power Spectral Density (APSD) and Cross-Power Spectral Density (CPSD). Two methods for axial void and velocity profiles reconstruction are discussed: the first one is based on the change of the break frequency of the neutron auto-power spectrum with axial core elevation, while the second refers to the estimation of transit times of propagating steam fluctuations between different axial detector positions. This paper summarizes the principles of the model and presents a numerical testing of the qualitative applicability to estimate the required parameters for the reconstruction of the void fraction profile from the neutron noise measurements. (authors)

  9. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    SciTech Connect (OSTI)

    Trianti, Nuri, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Su'ud, Zaki, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Arif, Idam, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) (Indonesia); Riyana, EkaSapta [Nuclear Energy Regulatory Agency (BAPETEN) (Indonesia)

    2014-09-30T23:59:59.000Z

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  10. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B. (Severna Park, MD)

    2004-03-16T23:59:59.000Z

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

  11. Advanced nuclear fuel

    SciTech Connect (OSTI)

    Terrani, Kurt

    2014-07-14T23:59:59.000Z

    Kurt Terrani uses his expertise in materials science to develop safer fuel for nuclear power plants.

  12. Nuclear Reaction Data Centers

    SciTech Connect (OSTI)

    McLane, V.; Nordborg, C.; Lemmel, H.D.; Manokhin, V.N.

    1988-01-01T23:59:59.000Z

    The cooperating Nuclear Reaction Data Centers are involved in the compilation and exchange of nuclear reaction data for incident neutrons, charged particles and photons. Individual centers may also have services in other areas, e.g., evaluated data, nuclear structure and decay data, reactor physics, nuclear safety; some of this information may also be exchanged between interested centers. 20 refs., 1 tab.

  13. Catalysinganenergyrevolution Nuclear Failures

    E-Print Network [OSTI]

    Laughlin, Robert B.

    extraction, fuel manufacture and management of spent fuel and waste. Currently, CEA is a large FrenchCatalysinganenergyrevolution France's Nuclear Failures The great illusion of nuclear energy greenpeace.org #12;Contents 2 Greenpeace International France's Nuclear Failures The French nuclear industry

  14. Advanced nuclear fuel

    ScienceCinema (OSTI)

    Terrani, Kurt

    2014-07-15T23:59:59.000Z

    Kurt Terrani uses his expertise in materials science to develop safer fuel for nuclear power plants.

  15. Focus Article Nuclear winter

    E-Print Network [OSTI]

    Robock, Alan

    the climatic effects of nuclear war. Smoke from the fires started by nuclear weapons, especially the black in recorded human history. Although the number of nuclear weapons in the world has fallen from 70,000 at its and Russia could still produce nuclear winter. This theory cannot be tested in the real world. However

  16. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 1. Investigation and evaluation of stress corrosion cracking in piping of boiling water reactor plants

    SciTech Connect (OSTI)

    Not Available

    1984-08-01T23:59:59.000Z

    IGSCC in BWR piping is occurring owing to a combination of material, environment, and stress factors, each of which can affect both the initiation of a stress-corrosion crack and the rate of its subsequent propagation. In evaluating long-term solutions to the problem, one needs to consider the effects of each of the proposed remedial actions. Mitigating actions to control IGSCC in BWR piping must be designed to alleviate one or more of the three synergistic factors: sensitized material, the convention BWR environment, and high tensile stresses. Because mitigating actions addressing each of these factors may not be fully effective under all anticipated operating conditions, mitigating actions should address two and preferably all three of the causative factors; e.g., material plus some control of water chemistry, or stress reversal plus controlled water chemistry.

  17. Experimental investigation of effects of surface roughness, wettability and boiling-time on steady state and transient CHF for nanofluids

    E-Print Network [OSTI]

    Sharma, Vivek Inder

    2012-01-01T23:59:59.000Z

    Critical Heat Flux (CHF) is one of the primary design constraints in a nuclear reactor. Increasing the CHF of water can enhance the safety margins of the current fleet of Light Water Reactors (LWRs) and/or increase their ...

  18. The Nuclear Revolution, Relative Gains, and International Nuclear Assistance

    E-Print Network [OSTI]

    Kroenig, Matthew

    2006-01-01T23:59:59.000Z

    it would transfer nuclear technology. Washington Post. 26preferences: the export of sensitive nuclear technology.export of sensitive nuclear technology presents a kind of

  19. Dynamics of nuclear envelope and nuclear pore complex formation

    E-Print Network [OSTI]

    Anderson, Daniel J.

    2008-01-01T23:59:59.000Z

    Limited expression of nuclear pore membrane glycoprotein 210suggests cell-type specific nuclear pores in metazoans. Expand Dultz, E. (2008). Nuclear pore complex assembly through

  20. Wisconsin Nuclear Profile - Point Beach Nuclear Plant

    U.S. Energy Information Administration (EIA) Indexed Site

    Point Beach Nuclear Plant" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License expiration...

  1. Tennessee Nuclear Profile - Watts Bar Nuclear Plant

    U.S. Energy Information Administration (EIA) Indexed Site

    Watts Bar Nuclear Plant" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License expiration...

  2. Massachusetts Nuclear Profile - Pilgrim Nuclear Power Station

    U.S. Energy Information Administration (EIA) Indexed Site

    Pilgrim Nuclear Power Station" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer cpacity factor (percent)","Type","Commercial operation date","License...

  3. Arkansas Nuclear Profile - Arkansas Nuclear One

    U.S. Energy Information Administration (EIA) Indexed Site

    Nuclear One" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License expiration date"...

  4. Is Nuclear Energy the Solution?

    E-Print Network [OSTI]

    Saier, Milton H.; Trevors, Jack T.

    2010-01-01T23:59:59.000Z

    009-0270-y Is Nuclear Energy the Solution? Milton H. Saier &in the last 50 years, nuclear energy subsidies have totaledadministration, the Global Nuclear Energy Partnership (GNEP)

  5. Nuclear Safety Research and Development...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Nuclear Safety Research and Development Proposal Review and Prioritization Process and Criteria Nuclear Safety Research and Development Program Office of Nuclear Safety Office of...

  6. NUCLEAR SCIENCE ANNUAL REPORT 1975

    E-Print Network [OSTI]

    Authors, Various

    2010-01-01T23:59:59.000Z

    Gove and A. H. Wapstra, Nuclear Data Tables 11, 127 (1972).P. Jackson, Chalk River Nuclear Laboratories Report (1975)national Conference on Nuclear Structure and Spec­ troscopy,

  7. China's Nuclear Industry After Fukushima

    E-Print Network [OSTI]

    YUAN, Jingdong

    2013-01-01T23:59:59.000Z

    the previous year. NUCLEAR TECHNOLOGY AND FUEL CYCLES China’third-generation nuclear technology and reactor design, withs own third-generation nuclear technology. Westing- house,

  8. NUCLEAR CHEMISTRY ANNUAL REPORT 1970

    E-Print Network [OSTI]

    Authors, Various

    2011-01-01T23:59:59.000Z

    Nuclear Laboratories, AECL, Chalk River, Ontario, Canada.Nuclear Laboratories, AECL, Chalk River, Ontario, Canada. 1.Nuclear Laboratories, AECL, Chalk River, Ontario, Canada. 1.

  9. Reactor & Nuclear Systems Publications | ORNL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications The...

  10. A phenomenological model of thermal-hydraulics of convective boiling during the quenching of hot rod bundles

    SciTech Connect (OSTI)

    Unal, C.; Nelson, R.

    1991-01-01T23:59:59.000Z

    After completion of the thermal-hydraulic model developed in a companion paper, the authors performed developmental assessment calculation of the model using steady-state and transient post-critical heat flux (CHF) data. This paper discusses the results of those calculations. The overall interfacial drag model predicted reasonable drag coefficients for both the nucleate boiling and the inverted annular flow (IAF) regimes. The predicted pressure drops agreed reasonably well with the measured data of two transient experiments, CCTF Run 14 and a Lehigh reflood test. The thermal-hydraulic model for post-CHF convective heat transfer predicted the rewetting velocities reasonably well for both experiments. The predicted average slope of the wall temperature traces for these tests showed reasonable agreement with the measured data, indicating that the transient-calculated precursory cooling rates agreed with measured data. The hot-patch model, in conjunction with the other thermal-hydraulic models, was capable of modeling the Winfrith post-CHF hot-patch experiments. The hot-patch model kept the wall temperatures at the specified levels in the hot-patch regions and did not allow any quench-front propagation from either the bottom or the top of the test section. The interfacial heat-transfer model tended to slightly underpredict the vapor temperatures. The maximum difference between calculated and measured vapor temperatures was 20%, with a 10% difference for the remainder of the runs considered. The wall-to-fluid heat transfer was predicted reasonably well, and the predicted wall temperatures were in reasonable agreement with measured data with a maximum relative error of less than 13%.

  11. Nuclear fuel post-irradiation examination equipment package

    SciTech Connect (OSTI)

    DeCooman, W.J. [AREVA NP Inc., Lynchburg, VA (United States); Spellman, D.J. [UT-Battelle, LLC, Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    2007-07-01T23:59:59.000Z

    Hot cell capabilities in the U.S. are being reviewed and revived to meet today's demand for fuel reliability, tomorrow's demands for higher burnup fuel and future demand for fuel recycling. Fuel reliability, zero tolerance for failure, is more than an industry buzz. It is becoming a requirement to meet the rapidly escalating demands for the impending renaissance of nuclear power generation, fuel development, and management of new waste forms that will need to be dealt with from programs such as the Global Nuclear Energy Partnership (GNEP). Fuel performance data is required to license fuel for higher burnup; to verify recycled fuel performance, such as MOX, for wide-scale use in commercial reactors; and, possibly, to license fuel for a new generation of fast reactors. Additionally, fuel isotopic analysis and recycling technologies will be critical factors in the goal to eventually close the fuel cycle. This focus on fuel reliability coupled with the renewed interest in recycling puts a major spotlight on existing hot cell capabilities in the U.S. and their ability to provide the baseline analysis to achieve a closed fuel cycle. Hot cell examination equipment is necessary to determine the characteristics and performance of irradiated materials that are subjected to nuclear reactor environments. The equipment within the hot cells is typically operated via master-slave manipulators and is typically manually operated. The Oak Ridge National Laboratory is modernizing their hot cell nuclear fuel examination equipment, installing automated examination equipment and data gathering capabilities. Currently, the equipment has the capability to perform fuel rod visual examinations, length and diametrical measurements, eddy current examination, profilometry, gamma scanning, fission gas collection and void fraction measurement, and fuel rod segmentation. The used fuel postirradiation examination equipment was designed to examine full-length fuel rods for both Boiling Water Reactors and Pressurized Water Reactors. (authors)

  12. Working Group Report on - Space Nuclear Power Systems and Nuclear...

    Energy Savers [EERE]

    Working Group Report on - Space Nuclear Power Systems and Nuclear Waste Technology R&D Working Group Report on - Space Nuclear Power Systems and Nuclear Waste Technology R&D "Even...

  13. Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report

    SciTech Connect (OSTI)

    NONE

    2000-08-01T23:59:59.000Z

    OAK B188 Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report. The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-formed approach for the design and regulation of nuclear power plants. This approach will include the development and/or confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRS) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go further by focusing on the design of new plants.

  14. Thermoacoustic Thermometry for Nuclear Reactor Monitoring

    SciTech Connect (OSTI)

    James A. Smith; Dale K. Kotter; Steven L. Garrett; Randall A. Ali

    2013-06-01T23:59:59.000Z

    On Friday, March 11, 2011, at 2:46pm (Japan Standard Trme), the Tohoku region on the east coast of northern Japan experi­enced what would become known as the largest earthquake in the country's history at magnitude 9.0 on the Richter scale. The Fukushima Daiichi nuclear power plant suffered exten­sive and irreversible damage. Six operating units were at the site, each with a boiling water reactor. When the earthquake struck, three of the six reactors were operating and the others were in a periodic inspection outage phase. In one reactor, all of the fuel had been relocated to a spent fuel pool in the reactor building. The seismic acceleration caused by the earthquake brought the three operating units to an automatic shutdown. Since there was damage to the power transmission lines, the emergency diesel generators (EDG) were automat­ically started to ensure continued cooling of the reactors and spent fuel pools. The situation was under control until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 meters, which was three times taller than the sea wall of 5m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to five of the six reactors. The flooding also resulted in the loss of instrumentation that would have other­ wise been used to monitor and control the emergency. The ugly aftermath included high radiation exposure to operators at the nuclear power plants and early contamina­tion of food supplies and water within several restricted areas in Japan, where high radiation levels have rendered them un­safe for human habitation. While the rest of the story will remain a tragic history, it is this part of the series of unfortunate events that has inspired our research. It has indubitably highlighted the need for a novel sensor and instrumentation system that can withstand similar or worse conditions to avoid future catastrophe and assume damage prevention as quickly as possible. This is the question which we are attempting to answer: Is it possible to implement a self-powered sensor that could transmit data independently of electronic networks while taking advantage of the harsh operating environment of the nuclear reactor?

  15. B53 Nuclear Bomb Dismantlement | National Nuclear Security Administrat...

    National Nuclear Security Administration (NNSA)

    Dismantlement | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear...

  16. Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay

    E-Print Network [OSTI]

    Quiter, Brian

    2012-01-01T23:59:59.000Z

    Spent Fuel Assay Using Nuclear Resonance Fluo- rescence,” Annual Meeting of the Institute of Nuclear Material Management,

  17. Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay

    E-Print Network [OSTI]

    Quiter, Brian

    2012-01-01T23:59:59.000Z

    of the Institute of Nuclear Material Management, Tucson, AZ,Assay, Institute of Nuclear Materials Management 51st Annual

  18. Nuclear disarmament verification

    SciTech Connect (OSTI)

    DeVolpi, A.

    1993-12-31T23:59:59.000Z

    Arms control treaties, unilateral actions, and cooperative activities -- reflecting the defusing of East-West tensions -- are causing nuclear weapons to be disarmed and dismantled worldwide. In order to provide for future reductions and to build confidence in the permanency of this disarmament, verification procedures and technologies would play an important role. This paper outlines arms-control objectives, treaty organization, and actions that could be undertaken. For the purposes of this Workshop on Verification, nuclear disarmament has been divided into five topical subareas: Converting nuclear-weapons production complexes, Eliminating and monitoring nuclear-weapons delivery systems, Disabling and destroying nuclear warheads, Demilitarizing or non-military utilization of special nuclear materials, and Inhibiting nuclear arms in non-nuclear-weapons states. This paper concludes with an overview of potential methods for verification.

  19. Nuclear radiation actuated valve

    DOE Patents [OSTI]

    Christiansen, David W. (Kennewick, WA); Schively, Dixon P. (Richland, WA)

    1985-01-01T23:59:59.000Z

    A nuclear radiation actuated valve for a nuclear reactor. The valve has a valve first part (such as a valve rod with piston) and a valve second part (such as a valve tube surrounding the valve rod, with the valve tube having side slots surrounding the piston). Both valve parts have known nuclear radiation swelling characteristics. The valve's first part is positioned to receive nuclear radiation from the nuclear reactor's fuel region. The valve's second part is positioned so that its nuclear radiation induced swelling is different from that of the valve's first part. The valve's second part also is positioned so that the valve's first and second parts create a valve orifice which changes in size due to the different nuclear radiation caused swelling of the valve's first part compared to the valve's second part. The valve may be used in a nuclear reactor's core coolant system.

  20. Triangle Universities Nuclear Laboratory

    SciTech Connect (OSTI)

    Not Available

    1991-01-01T23:59:59.000Z

    This report contains brief papers that discusses the following topics: Fundamental Symmetries in the Nucleus; Internucleon Interactions; Dynamics of Very Light Nuclei; Facets of the Nuclear Many-Body Problem; and Nuclear Instruments and Methods.

  1. Promulgating Nuclear Safety Requirements

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1996-05-15T23:59:59.000Z

    Applies to all Nuclear Safety Requirements Adopted by the Department to Govern the Conduct of its Nuclear Activities. Cancels DOE P 410.1. Canceled by DOE N 251.85.

  2. NUCLEAR PLANT AND CONTROL

    E-Print Network [OSTI]

    NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: software require- ments, safety analysis, formal, the missiles, and the digital protection systems embed- ded in nuclear power plants. Obviously, safety method SOFTWARE SAFETY ANALYSIS OF DIGITAL PROTECTION SYSTEM REQUIREMENTS USING A QUALITATIVE FORMAL

  3. Hegemony and nuclear proliferation

    E-Print Network [OSTI]

    Miller, Nicholas L. (Nicholas LeSuer)

    2014-01-01T23:59:59.000Z

    Contrary to longstanding of predictions of nuclear tipping points, the number of states interested in nuclear weapons has sharply declined in recent decades. In contrast to existing explanations, this dissertation argues ...

  4. Nuclear Explosive Safety Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2009-04-14T23:59:59.000Z

    This Manual provides supplemental details to support the requirements of DOE O 452.2D, Nuclear Explosive Safety.

  5. 3D NUCLEAR SEGMENTAT

    Energy Science and Technology Software Center (OSTI)

    003029WKSTN00 Delineation of nuclear structures in 3D multicellular systems  https://vision.lbl.gov/Software/3DMorphometry/ 

  6. Nuclear Nonproliferation Programs | ORNL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Initiatives Nonproliferation Technology Nonproliferation Systems Safeguards and Security Technology International Safeguards Nuclear Material Detection and Characterization For...

  7. Nuclear Multifragmentation Critical Exponents

    E-Print Network [OSTI]

    Wolfgang Bauer; William Friedman

    1994-11-14T23:59:59.000Z

    We show that the critical exponents of nuclear multi-fragmentation have not been determined conclusively yet.

  8. advanced nuclear reactor: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  9. advanced nuclear plant: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  10. advanced nuclear power: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  11. advanced nuclear plants: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  12. advanced nuclear reactors: Topics by E-print Network

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    . . . . 18 3.4.1 Heat Exchanger - Code description . . . . . . . . . . . . . . . 18 3.4.2 Simulation ResultsADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING...

  13. Nuclear Energy Research Brookhaven National

    E-Print Network [OSTI]

    Ohta, Shigemi

    Nuclear Energy Research Brookhaven National Laboratory William C. Horak, Chair Nuclear Science and Technology Department #12;BNL Nuclear Energy Research Brookhaven Graphite Research Reactor - 1948 National Nuclear Data Center - 1952* High Flux Beam Reactor - 1964 Technical Support for NRC - 1974

  14. NUCLEAR POWER in CALIFORNIA

    E-Print Network [OSTI]

    NUCLEAR POWER in CALIFORNIA: 2007 STATUS REPORT CALIFORNIA ENERGY COMMISSION October 2007 CEC-100, California Contract No. 700-05-002 Prepared For: California Energy Commission Barbara Byron, Senior Nuclear public workshops on nuclear power. The Integrated Energy Policy Report Committee, led by Commissioners

  15. Nuclear fact book

    SciTech Connect (OSTI)

    Hill, O.F.; Platt, A.M.; Robinson, J.V.

    1983-05-01T23:59:59.000Z

    This reference provides significant highlights and summary facts in the following areas: general energy; nuclear energy; nuclear fuel cycle; uranium supply and enrichment; nuclear reactors; spent fuel and advanced repacking concepts; reprocessing; high-level waste; gaseous waste; transuranic waste; low-level waste; remedial action; transportation; disposal; radiation information; environment; legislation; socio-political aspects; conversion factors; and a glossary. (GHT)

  16. NUCLEAR PLANT OPERATIONS AND

    E-Print Network [OSTI]

    Demazière, Christophe

    NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: moderator temper ature coefficient, reactivity co reactor Unit 4 of the Ringhals Nuclear Power Plant (Sweden) during fuel cycle 16 is analyzed absorption cross-section behavior. Consequently, if NUCLEAR TECHNOLOGY VOL. 140 NOV. 2002 147 #12;Demazière

  17. NUCLEAR PLANT OPERATIONS AND

    E-Print Network [OSTI]

    Pázsit, Imre

    NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: moderator temper- ature coefficient, reactivity co reactor Unit 4 of the Ringhals Nuclear Power Plant (Sweden) during fuel cycle 16 is analyzed. Consequently, if*E-mail: demaz@nephy.chalmers.se NUCLEAR TECHNOLOGY VOL. 140 NOV. 2002 147 #12;high-burnup fuel

  18. Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2009-04-14T23:59:59.000Z

    This Order establishes requirements to implement the nuclear explosive safety elements of DOE O 452.1D, Nuclear Explosive and Weapon Surety Program, for routine and planned nuclear explosive operations. Cancels DOE O 452.2C. Admin Chg 1, 7-10-13

  19. Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2009-04-14T23:59:59.000Z

    This Department of Energy (DOE) Order establishes requirements to implement the nuclear explosive safety (NES) elements of DOE O 452.1D, Nuclear Explosive and Weapon Surety Program, for routine and planned nuclear explosive operations (NEOs). Cancels DOE O 452.2C. Admin Chg 1, dated 7-10-13, cancels DOE O 452.2D.

  20. Nuclear & Particle Physics, Astrophysics, Cosmology

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Nuclear & Particle Physics science-innovationassetsimagesicon-science.jpg Nuclear & Particle Physics, Astrophysics, Cosmology National security depends on science and...

  1. Is Nuclear Energy the Solution?

    E-Print Network [OSTI]

    Saier, Milton H.; Trevors, Jack T.

    2010-01-01T23:59:59.000Z

    the potential of nuclear power to combat global warming havecompetitive today, and for nuclear power to succeed, it must

  2. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    SciTech Connect (OSTI)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01T23:59:59.000Z

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices.

  3. Nuclear spirals in galaxies

    E-Print Network [OSTI]

    Witold Maciejewski

    2006-11-08T23:59:59.000Z

    Recent high-resolution observations indicate that nuclear spirals are often present in the innermost few hundred parsecs of disc galaxies. My models show that nuclear spirals form naturally as a gas response to non-axisymmetry in the gravitational potential. Some nuclear spirals take the form of spiral shocks, resulting in streaming motions in the gas, and in inflow comparable to the accretion rates needed to power local Active Galactic Nuclei. Recently streaming motions of amplitude expected from the models have been observed in nuclear spirals, confirming the role of nuclear spirals in feeding of the central massive black holes.

  4. Commercial nuclear power 1990

    SciTech Connect (OSTI)

    Not Available

    1990-09-28T23:59:59.000Z

    This report presents the status at the end of 1989 and the outlook for commercial nuclear capacity and generation for all countries in the world with free market economies (FME). The report provides documentation of the US nuclear capacity and generation projections through 2030. The long-term projections of US nuclear capacity and generation are provided to the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) for use in estimating nuclear waste fund revenues and to aid in planning the disposal of nuclear waste. These projections also support the Energy Information Administration's annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment, and are provided to the Organization for Economic Cooperation and Development. The foreign nuclear capacity projections are used by the DOE uranium enrichment program in assessing potential markets for future enrichment contracts. The two major sections of this report discuss US and foreign commercial nuclear power. The US section (Chapters 2 and 3) deals with (1) the status of nuclear power as of the end of 1989; (2) projections of nuclear capacity and generation at 5-year intervals from 1990 through 2030; and (3) a discussion of institutional and technical issues that affect nuclear power. The nuclear capacity projections are discussed in terms of two projection periods: the intermediate term through 2010 and the long term through 2030. A No New Orders case is presented for each of the projection periods, as well as Lower Reference and Upper Reference cases. 5 figs., 30 tabs.

  5. Initiating Event Rates at U.S. Nuclear Power Plants 1988–2013

    SciTech Connect (OSTI)

    John A. Schroeder; Gordon R. Bower

    2014-02-01T23:59:59.000Z

    Analyzing initiating event rates is important because it indicates performance among plants and also provides inputs to several U.S. Nuclear Regulatory Commission (NRC) risk-informed regulatory activities. This report presents an analysis of initiating event frequencies at U.S. commercial nuclear power plants since each plant’s low-power license date. The evaluation is based on the operating experience from fiscal year 1988 through 2013 as reported in licensee event reports. Engineers with nuclear power plant experience staff reviewed each event report since the last update to this report for the presence of valid scrams or reactor trips at power. To be included in the study, an event had to meet all of the following criteria: includes an unplanned reactor trip (not a scheduled reactor trip on the daily operations schedule), sequence of events starts when reactor is critical and at or above the point of adding heat, occurs at a U.S. commercial nuclear power plant (excluding Fort St. Vrain and LaCrosse), and is reported by a licensee event report. This report displays occurrence rates (baseline frequencies) for the categories of initiating events that contribute to the NRC’s Industry Trends Program. Sixteen initiating event groupings are trended and displayed. Initiators are plotted separately for initiating events with different occurrence rates for boiling water reactors and pressurized water reactors. p-values are given for the possible presence of a trend over the most recent 10 years.

  6. Aging Management Guideline for commercial nuclear power plants: Motor control centers; Final report

    SciTech Connect (OSTI)

    Toman, G.; Gazdzinski, R.; O`Hearn, E. [Ogden Environmental and Energy Services Co., Inc., Blue Bell, PA (United States)

    1994-02-01T23:59:59.000Z

    This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) commercial nuclear power plant motor control centers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  7. Nuclear facilities: criteria for the design and operation of ventilation systems for nuclear installations other than nuclear reactors

    E-Print Network [OSTI]

    International Organization for Standardization. Geneva

    2004-01-01T23:59:59.000Z

    Nuclear facilities: criteria for the design and operation of ventilation systems for nuclear installations other than nuclear reactors

  8. Coherent Nuclear Radiation

    E-Print Network [OSTI]

    V. I. Yukalov; E. P. Yukalova

    2004-06-22T23:59:59.000Z

    The main part of this review is devoted to the comprehensive description of coherent radiation by nuclear spins. The theory of nuclear spin superradiance is developed and the experimental observations of this phenomenon are considered. The intriguing problem of how coherence develops from initially incoherent quantum fluctuations is analysed. All main types of coherent radiation by nuclear spins are discussed, which are: free nuclear induction, collective induction, maser generation, pure superradiance, triggered superradiance, pulsing superradiance, punctuated superradiance, and induced emission. The influence of electron-nuclear hyperfine interactions and the role of magnetic anisotropy are studied. Conditions for realizing spin superradiance by magnetic molecules are investigated. The possibility of nuclear matter lasing, accompanied by pion or dibaryon radiation, is briefly touched.

  9. Instrumentation for Nuclear Applications

    SciTech Connect (OSTI)

    NONE

    1998-09-18T23:59:59.000Z

    The objective of this project was to develop and coordinate nuclear instrumentation standards with resulting economies for the nuclear and radiation fields. There was particular emphasis on coordination and management of the Nuclear Instrument Module (NIM) System, U.S. activity involving the CAMAC international standard dataway system, the FASTBUS modular high-speed data acquisition and control system and processing and management of national nuclear instrumentation and detector standards, as well as a modest amount of assistance and consultation services to the Pollutant Characterization and Safety Research Division of the Office of Health and Environmental Research. The principal accomplishments were the development and maintenance of the NIM instrumentation system that is the predominant instrumentation system in the nuclear and radiation fields worldwide, the CAMAC digital interface system in coordination with the ESONE Committee of European Laboratories, the FASTBUS high-speed system and numerous national and international nuclear instrumentation standards.

  10. Nuclear Science References Database

    E-Print Network [OSTI]

    B. Pritychenko; E. B?ták; B. Singh; J. Totans

    2014-07-08T23:59:59.000Z

    The Nuclear Science References (NSR) database together with its associated Web interface, is the world's only comprehensive source of easily accessible low- and intermediate-energy nuclear physics bibliographic information for more than 210,000 articles since the beginning of nuclear science. The weekly-updated NSR database provides essential support for nuclear data evaluation, compilation and research activities. The principles of the database and Web application development and maintenance are described. Examples of nuclear structure, reaction and decay applications are specifically included. The complete NSR database is freely available at the websites of the National Nuclear Data Center http://www.nndc.bnl.gov/nsr and the International Atomic Energy Agency http://www-nds.iaea.org/nsr.

  11. INDEPENDENT PARTICLE ASPECTS OF NUCLEAR DYNAMICS

    E-Print Network [OSTI]

    Robel, M.C.

    2011-01-01T23:59:59.000Z

    situations: nuclear vibrations, fission, collisions, theformulae to nuclear vibrations, fission, collisions, thenuclear phenomena: nuclear vibrations, fission, collisions,

  12. Absolute nuclear material assay

    DOE Patents [OSTI]

    Prasad, Manoj K. (Pleasanton, CA); Snyderman, Neal J. (Berkeley, CA); Rowland, Mark S. (Alamo, CA)

    2012-05-15T23:59:59.000Z

    A method of absolute nuclear material assay of an unknown source comprising counting neutrons from the unknown source and providing an absolute nuclear material assay utilizing a model to optimally compare to the measured count distributions. In one embodiment, the step of providing an absolute nuclear material assay comprises utilizing a random sampling of analytically computed fission chain distributions to generate a continuous time-evolving sequence of event-counts by spreading the fission chain distribution in time.

  13. Absolute nuclear material assay

    DOE Patents [OSTI]

    Prasad, Manoj K. (Pleasanton, CA); Snyderman, Neal J. (Berkeley, CA); Rowland, Mark S. (Alamo, CA)

    2010-07-13T23:59:59.000Z

    A method of absolute nuclear material assay of an unknown source comprising counting neutrons from the unknown source and providing an absolute nuclear material assay utilizing a model to optimally compare to the measured count distributions. In one embodiment, the step of providing an absolute nuclear material assay comprises utilizing a random sampling of analytically computed fission chain distributions to generate a continuous time-evolving sequence of event-counts by spreading the fission chain distribution in time.

  14. Nuclear Fabrication Consortium

    SciTech Connect (OSTI)

    Levesque, Stephen

    2013-04-05T23:59:59.000Z

    This report summarizes the activities undertaken by EWI while under contract from the Department of Energy (DOE) � Office of Nuclear Energy (NE) for the management and operation of the Nuclear Fabrication Consortium (NFC). The NFC was established by EWI to independently develop, evaluate, and deploy fabrication approaches and data that support the re-establishment of the U.S. nuclear industry: ensuring that the supply chain will be competitive on a global stage, enabling more cost-effective and reliable nuclear power in a carbon constrained environment. The NFC provided a forum for member original equipment manufactures (OEM), fabricators, manufacturers, and materials suppliers to effectively engage with each other and rebuild the capacity of this supply chain by : � Identifying and removing impediments to the implementation of new construction and fabrication techniques and approaches for nuclear equipment, including system components and nuclear plants. � Providing and facilitating detailed scientific-based studies on new approaches and technologies that will have positive impacts on the cost of building of nuclear plants. � Analyzing and disseminating information about future nuclear fabrication technologies and how they could impact the North American and the International Nuclear Marketplace. � Facilitating dialog and initiate alignment among fabricators, owners, trade associations, and government agencies. � Supporting industry in helping to create a larger qualified nuclear supplier network. � Acting as an unbiased technology resource to evaluate, develop, and demonstrate new manufacturing technologies. � Creating welder and inspector training programs to help enable the necessary workforce for the upcoming construction work. � Serving as a focal point for technology, policy, and politically interested parties to share ideas and concepts associated with fabrication across the nuclear industry. The report the objectives and summaries of the Nuclear Fabrication Consortium projects. Full technical reports for each of the projects have been submitted as well.

  15. Assessing the nuclear age

    SciTech Connect (OSTI)

    Ackland, L.; McGuire, S.

    1986-01-01T23:59:59.000Z

    This book presents papers on nuclear weapons and arms control. Topics considered include historical aspects, the arms race, nuclear power, flaws in the non-proliferation treaty, North-South issues, East-West confrontation, Soviet decision making with regard to national defense, US and Soviet perspectives on national security, ballistic missile defense (''Star Wars''), political aspects, nuclear winter, stockpiles, US foreign policy, and military strategy.

  16. Nuclear Spectra from Skyrmions

    SciTech Connect (OSTI)

    Manton, N. S. [DAMTP, University of Cambridge, Wilberforce Road, Cambridge CB3 0WA (United Kingdom)

    2009-08-26T23:59:59.000Z

    The structures of Skyrmions, especially for baryon numbers 4, 8 and 12, are reviewed. The quantized Skyrmion states are compared with nuclear spectra.

  17. Nuclear Physics from QCD

    E-Print Network [OSTI]

    U. van Kolck

    2008-12-20T23:59:59.000Z

    Effective field theories provide a bridge between QCD and nuclear physics. I discuss light nuclei from this perspective, emphasizing the role of fine-tuning.

  18. Tag: nuclear deterrence

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    class"field-item even" property"content:encoded">

    The National Nuclear Security Administration has completed a major capital improvement project that has...

  19. Reference handbook: Nuclear criticality

    SciTech Connect (OSTI)

    Not Available

    1991-12-06T23:59:59.000Z

    The purpose for this handbook is to provide Rocky Flats personnel with the information necessary to understand the basic principles underlying a nuclear criticality.

  20. Nuclear Physics: Experiment Research

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    UserResearcher Information print version Research Highlights Public Interest Nuclear Physics Accelerator Free Electron Laser (FEL) Medical Imaging Physics Topics Campaigns...

  1. Nuclear Safety Regulatory Framework

    Broader source: Energy.gov (indexed) [DOE]

    overall Nuclear Safety Policy & ESH Goals Safety Basis Review and Approval In the DOE governance model, contractors responsible for the facility develop the safety basis and...

  2. Nuclear power attitude trends

    SciTech Connect (OSTI)

    Nealey, S.M.

    1981-11-01T23:59:59.000Z

    The increasing vulnerability of nuclear power to political pressures fueled by public concerns, particularly about nuclear plant safety and radioactive waste disposal, has become obvious. Since Eisenhower's Atoms-for-Peace program, utility and government plans have centered on expansion of nuclear power generating capability. While supporters have outnumbered opponents of nuclear power expansion for many years, in the wake of the Three Mile Island (TMI) accident the margin of support has narrowed. The purpose of this paper is to report and put in perspective these long-term attitude trends.

  3. Management of Nuclear Materials

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2009-08-17T23:59:59.000Z

    To establish requirements for the lifecycle management of DOE owned and/or managed accountable nuclear materials. Cancels DOE O 5660.1B.

  4. Unclassified Controlled Nuclear Information

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-09-25T23:59:59.000Z

    To prevent unauthorized dissemination of Unclassified Controlled Nuclear Information (UCNI). Cancels DOE 5635.4 and DOE 5650.3A

  5. National Nuclear Security Administration

    Broader source: Energy.gov (indexed) [DOE]

    and Related Structures within TA-3 at Los Alamos National Laboratory, Los Alamos, New Mexico U. S. Department of Energy National Nuclear Security Administration Los Alamos Area...

  6. Nuclear | Department of Energy

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItemResearch > TheNuclear Astrophysics One of the greatNuclearNuclear Nuclear An error

  7. National Nuclear Science Week 2012 - SRSCRO

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    to Know Nuclear National Nuclear Science Week January 23 - 27, 2012 Fostering a deeper public understanding Logos for: National Nuclear Science Week, Nuclear Workforce Initiative,...

  8. Nuclear Reactions and Reactor Safety

    E-Print Network [OSTI]

    Onuchic, José

    Nuclear Reactions and Reactor Safety DO NOT LICK We haven't entirely nailed down what element nuclear chain reaction, 1938 #12;Nuclear Chain Reactions Do nuclear chain reactions lead to runaway explosions? or ? -Controlled nuclear chain reactions possible: control energy release/sec -> More

  9. Nuclear fuel cycle information workshop

    SciTech Connect (OSTI)

    Not Available

    1983-01-01T23:59:59.000Z

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

  10. NUCLEAR ENERGY PERGAMON Annals of Nuclear Energy 27 (2000) 138551398

    E-Print Network [OSTI]

    Pázsit, Imre

    annafs of NUCLEAR ENERGY PERGAMON Annals of Nuclear Energy 27 (2000) 138551398 www-4549(00)00033-5 #12;1386 I. Phi!, V. Arzhanov. /Annals qf Nuclear Energy 27 (2000) 1385-1398 subcritical systems (ADS

  11. Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay

    E-Print Network [OSTI]

    Quiter, Brian

    2012-01-01T23:59:59.000Z

    W. Bertozzi and R.J. Ledoux, “Nuclear resonance ?uorescenceUrakawa, “Compton ring for nuclear waste management,” Nucl.and B.J. Quiter, “Using Nuclear Resonance Fluorscence for

  12. The Nuclear Revolution, Relative Gains, and International Nuclear Assistance

    E-Print Network [OSTI]

    Kroenig, Matthew

    2006-01-01T23:59:59.000Z

    204. Bhatia, Shyam. 1988. Nuclear rivals in the Middle East.of the merits of selective nuclear proliferation. Journal ofThe Case for a Ukranian nuclear deterrent. Foreign Affairs.

  13. We study the continuous extractive distillation of minimum and maximum boiling azeotropic mixtures A-B with a heavy or a light entrainer E, intending to assess its feasibility based on thermodynamic insights.

    E-Print Network [OSTI]

    Mailhes, Corinne

    #12;#12;Abstract We study the continuous extractive distillation of minimum and maximum boiling on operating parameters: distillate product purity and recovery, reflux ratio R and entrainer ­ feed flow rate. For the 1.0-2 class both A and B can be distillated. For one of them there exists a maximum entrainer - feed

  14. International Nuclear Safeguards | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated Codes | National NuclearInterlibrary LoanSafeguards | National Nuclear

  15. Nuclear Safety Information Agreement Between the U.S. Nuclear...

    Office of Environmental Management (EM)

    Information Agreement Between the U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, and the U.S. Department of Energy, Office of Environment,...

  16. Mr. John Kinneman, Chief Nuclear Materfals Branch Nuclear Regulatory...

    Office of Legacy Management (LM)

    111989 Mr. John Kinneman, Chief Nuclear Materfals Branch Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia. Pennsylvania 19406 Dear Mr. Kinneman: -;' .-. 'W...

  17. Western Interstate Nuclear Compact State Nuclear Policy (Multiple States)

    Broader source: Energy.gov [DOE]

    Legislation authorizes states' entrance into the Western Interstate Nuclear Compact, which aims to undertake the cooperation of participating states in deriving the optimum benefit from nuclear and...

  18. (U) Nuclear Data

    SciTech Connect (OSTI)

    White, Morgan C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-01-23T23:59:59.000Z

    Nuclear data comes from a variety of sources and in many flavors. Understanding where the data you use comes from and what flavor it is can be essential to understand and interpret your results. This talk will discuss the nuclear data pipeline with particular emphasis on providing links to additional resources that can be used to explore the issues you will encounter.

  19. Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2006-06-12T23:59:59.000Z

    The directive establishes specific nuclear explosive safety (NES) program requirements to implement the DOE NES standards and other NES criteria for routine and planned nuclear explosive operations. Cancels DOE O 452.2B. Canceled by DOE O 452.2D.

  20. Nuclear physics and cosmology

    SciTech Connect (OSTI)

    Coc, Alain [Centre de Sciences Nucléaires et de Sciences de la Matière (CSNSM), CNRS/IN2P3, Université Paris Sud 11, UMR 8609, Bâtiment 104, F-91405 Orsay Campus (France)

    2014-05-09T23:59:59.000Z

    There are important aspects of Cosmology, the scientific study of the large scale properties of the universe as a whole, for which nuclear physics can provide insights. Here, we will focus on Standard Big-Bang Nucleosynthesis and we refer to the previous edition of the School [1] for the aspects concerning the variations of constants in nuclear cosmo-physics.

  1. Nuclear Science & Engineering

    E-Print Network [OSTI]

    .90 76 Nuclear 19.9 1.68 25 Natural Gas 17.7 5.87 91 Hydroelectricity 6.6 Petroleum 3.0 5.39 88 Non Nuclear Science & Engineering Natural Gas Source: Sproule Associates Ltd. Generating costs are high. Gas shutdown: · Pickering 1 (515 MW(e), PHWR, Canada) reconnected 26 Sep 05 Final shutdowns: 3 reactors, Sweden

  2. China's Nuclear Industry After Fukushima

    E-Print Network [OSTI]

    YUAN, Jingdong

    2013-01-01T23:59:59.000Z

    Brief 2013-9 January 2013 China’s Nuclear Industry Aftera significant impact on the future of China’s nuclear power.the importance of safety as China builds more nuclear power

  3. NUCLEAR CHEMISTRY ANNUAL REPORT 1970

    E-Print Network [OSTI]

    Authors, Various

    2011-01-01T23:59:59.000Z

    Letters 24, 1507 (1970); Nuclear Data B4, 663 (1970). 5. R.S. Hager and E. C. Seltzer, Nuclear Data A4, 1 (1968). 6. H.J. Nijgh, and R. Van Lieshout, Nuclear Spectroscopy Tables (

  4. Is Nuclear Energy the Solution?

    E-Print Network [OSTI]

    Saier, Milton H.; Trevors, Jack T.

    2010-01-01T23:59:59.000Z

    10.1007/s11270-009-0270-y Is Nuclear Energy the Solution?MHS) attended a lecture on “Nuclear Responsibility” on theof the Alliance for Nuclear Responsibility. The information

  5. THz Dynamic Nuclear Polarization NMR

    E-Print Network [OSTI]

    Nanni, Emilio Alessandro

    Dynamic nuclear polarization (DNP) increases the sensitivity of nuclear magnetic resonance (NMR) spectroscopy by using high frequency microwaves to transfer the polarization of the electrons to the nuclear spins. The ...

  6. Supporting Our Nation's Nuclear Industry

    ScienceCinema (OSTI)

    Lyons, Peter

    2013-05-29T23:59:59.000Z

    On the 60th anniversary of the world's first nuclear power plant to produce electricity, Assistant Secretary for Nuclear Energy Peter Lyons discusses the Energy Department's and the Administration's commitment to promoting a nuclear renaissance in the United States.

  7. China's Nuclear Industry After Fukushima

    E-Print Network [OSTI]

    YUAN, Jingdong

    2013-01-01T23:59:59.000Z

    2013-9 January 2013 China’s Nuclear Industry After FukushimaMarch 2011 Fukushima nuclear accident has had a significanton the future of China’s nuclear power. First, it highlights

  8. World nuclear outlook 1995

    SciTech Connect (OSTI)

    NONE

    1995-09-29T23:59:59.000Z

    As part of the EIA program to provide energy information, this analysis report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the uranium market. Long-term projections of US nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed for the Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM). In turn, the OCRWM provides partial funding for preparation of this report. The projections of uranium requirements are provided to the Organization for Economic Cooperation and Development (OECD) for preparation of the Nuclear Energy Agency/OECD report, Summary of Nuclear Power and Fuel Cycle Data in OECD Member Countries.

  9. World nuclear outlook 1994

    SciTech Connect (OSTI)

    NONE

    1994-12-01T23:59:59.000Z

    As part of the EIA program to provide energy information, this analysis report presents the current status and projections through 2010 of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the uranium market. Long-term projections of US nuclear capacity, generation, and spent fuel discharges for three different scenarios through 2040 are developed for the Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM). In turn, the OCRWM provides partial funding for preparation of this report. The projections of uranium requirements are provided to the Organization for Economic Cooperation and Development (OECD) for preparation of the Nuclear Energy Agency/OECD report, Summary of Nuclear Power and Fuel Cycle Data in OECD Member Countries.

  10. US nuclear weapons policy

    SciTech Connect (OSTI)

    May, M.

    1990-12-05T23:59:59.000Z

    We are closing chapter one'' of the nuclear age. Whatever happens to the Soviet Union and to Europe, some of the major determinants of nuclear policy will not be what they have been for the last forty-five years. Part of the task for US nuclear weapons policy is to adapt its nuclear forces and the oganizations managing them to the present, highly uncertain, but not urgently competitive situation between the US and the Soviet Union. Containment is no longer the appropriate watchword. Stabilization in the face of uncertainty, a more complicated and politically less readily communicable goal, may come closer. A second and more difficult part of the task is to deal with what may be the greatest potential source of danger to come out of the end of the cold war: the breakup of some of the cooperative institutions that managed the nuclear threat and were created by the cold war. These cooperative institutions, principally the North Atlantic Treaty Organization (NATO), the Warsaw Pact, the US-Japan alliance, were not created specifically to manage the nuclear threat, but manage it they did. A third task for nuclear weapons policy is that of dealing with nuclear proliferation under modern conditions when the technologies needed to field effective nuclear weapons systems and their command and control apparatus are ever more widely available, and the leverage over some potential proliferators, which stemmed from superpower military support, is likely to be on the wane. This paper will make some suggestions regarding these tasks, bearing in mind that the unsettled nature of that part of the world most likely to become involved in nuclear weapons decisions today must make any suggestions tentative and the allowance for surprise more than usually important.

  11. Nuclear Power Generating Facilities (Maine)

    Broader source: Energy.gov [DOE]

    The first subchapter of the statute concerning Nuclear Power Generating Facilities provides for direct citizen participation in the decision to construct any nuclear power generating facility in...

  12. Pollux | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Pollux | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy...

  13. Virginia Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  14. Ohio Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Ohio nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  15. Arkansas Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  16. Michigan Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  17. Alabama Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  18. Texas Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  19. Tennessee Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Tennessee nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  20. Georgia Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  1. Nebraska Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Nebraska nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  2. Arizona Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  3. Maryland Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  4. Illinois Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Illinois nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  5. Florida Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Florida nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  6. Wisconsin Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Wisconsin nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  7. Minnesota Nuclear Profile - Power Plants

    U.S. Energy Information Administration (EIA) Indexed Site

    Minnesota nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  8. China's Nuclear Industry After Fukushima

    E-Print Network [OSTI]

    YUAN, Jingdong

    2013-01-01T23:59:59.000Z

    s Nuclear Industry After Fukushima Jingdong YUAN SummaryT he March 2011 Fukushima nuclear accident has had aand speedy responses to Fukushima-like and other unexpected

  9. Evaluated Nuclear Data

    SciTech Connect (OSTI)

    Oblozinsky, P.; Oblozinsky,P.; Herman,M.; Mughabghab,S.F.

    2010-10-01T23:59:59.000Z

    This chapter describes the current status of evaluated nuclear data for nuclear technology applications. We start with evaluation procedures for neutron-induced reactions focusing on incident energies from the thermal energy up to 20 MeV, though higher energies are also mentioned. This is followed by examining the status of evaluated neutron data for actinides that play dominant role in most of the applications, followed by coolants/moderators, structural materials and fission products. We then discuss neutron covariance data that characterize uncertainties and correlations. We explain how modern nuclear evaluated data libraries are validated against an extensive set of integral benchmark experiments. Afterwards, we briefly examine other data of importance for nuclear technology, including fission yields, thermal neutron scattering and decay data. A description of three major evaluated nuclear data libraries is provided, including the latest version of the US library ENDF/B-VII.0, European JEFF-3.1 and Japanese JENDL-3.3. A brief introduction is made to current web retrieval systems that allow easy access to a vast amount of up-to-date evaluated nuclear data for nuclear technology applications.

  10. Monthly/Annual Energy Review - nuclear section

    Reports and Publications (EIA)

    2015-01-01T23:59:59.000Z

    Monthly and latest annual statistics on nuclear electricity capacity, generation, and number of operable nuclear reactors.

  11. Nuclear Renaissance Requires Nuclear Enlightenment W J Nuttall

    E-Print Network [OSTI]

    Aickelin, Uwe

    Nuclear Renaissance Requires Nuclear Enlightenment W J Nuttall Judge Business School, Cambridge University, Trumpington Street Cambridge, CB2 1AG UK Abstract Nuclear energy was developed by technocratic as a result of global anthropogenic climate change, nuclear power might actually represent a means to preserve

  12. NUCLEAR ENERGY Annals of Nuclear Energy 32 (2005) 812842

    E-Print Network [OSTI]

    Demazière, Christophe

    annals of NUCLEAR ENERGY Annals of Nuclear Energy 32 (2005) 812­842 www.elsevier.com/locate/anucene Identification and localization of absorbers of variable strength in nuclear reactors C. Demazie`re a,*, G evenly distrib- uted throughout the core of a commercial nuclear reactor. The novelty

  13. Nuclear Regulatory Commission issuances

    SciTech Connect (OSTI)

    NONE

    1996-03-01T23:59:59.000Z

    This document is the March 1996 listing of NRC issuances. Included are: (1) NRC orders granting Cleveland Electric Illuminating Company`s petition for review of the ASLB order LBP-95-17, (2) NRC orders relating to the potential disqualification of two commissioners in the matter of the decommissioning of Yankee Nuclear Power Station, (3) ASLB orders pertaining to the Oncology Services Corporation, (4) ASLB orders pertaining to the Radiation Oncology Center, (5) ASLB orders pertaining to the Yankee Nuclear Power Station, and (6) Director`s decision pertaining to the Yankee Nuclear Power Station.

  14. Monitoring international nuclear activity

    SciTech Connect (OSTI)

    Firestone, R.B.

    2006-05-19T23:59:59.000Z

    The LBNL Table of Isotopes website provides primary nuclearinformation to>150,000 different users annually. We have developedthe covert technology to identify users by IP address and country todetermine the kinds of nuclear information they are retrieving. Wepropose to develop pattern recognition software to provide an earlywarning system to identify Unusual nuclear activity by country or regionSpecific nuclear/radioactive material interests We have monitored nuclearinformation for over two years and provide this information to the FBIand LLNL. Intelligence is gleaned from the website log files. Thisproposal would expand our reporting capabilities.

  15. Nuclear Emergency Search Team

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1991-09-20T23:59:59.000Z

    To establish Department of Energy (DOE) policy for Nuclear Emergency Search Team (NEST) operations to malevolent radiological incidents. This directive does not cancel another directive. Canceled by DOE O 153.1.

  16. State Nuclear Profiles 2010

    U.S. Energy Information Administration (EIA) Indexed Site

    60 Vermont Vermont total electric power industry, summer capacity and net generation, by energy source, 2010 Nuclear 620 55.0 4,782 72.2 Hydro and Pumped Storage 324 28.7 1,347...

  17. Pioneering the nuclear age

    SciTech Connect (OSTI)

    Seaborg, G.T.

    1988-09-01T23:59:59.000Z

    This paper reviews the historical aspects of nuclear physics. The scientific aspects of the early transuranium elements are discussed and arms control measures are reviewed. 11 refs., 14 figs. (LSP)

  18. Management of Nuclear Materials

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1994-05-26T23:59:59.000Z

    To establish requirements and procedures for the management of nuclear materials within the Department of Energy (DOE). Cancels DOE 5660.1A. Canceled by DOE O 410.2.

  19. Nuclear Material Packaging Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2008-03-07T23:59:59.000Z

    The manual provides detailed packaging requirements for protecting workers from exposure to nuclear materials stored outside of an approved engineered contamination barrier. No cancellation. Certified 11-18-10.

  20. Relativistic Nuclear Collisions

    E-Print Network [OSTI]

    Reinhard Stock

    2009-07-29T23:59:59.000Z

    A comprehensive introduction is given to the field of relativistic nuclear collisions, and the phase diagram of strongly interacting matter. The content of this complex of reviews is shown.

  1. JPRS report, nuclear developments

    SciTech Connect (OSTI)

    NONE

    1991-03-28T23:59:59.000Z

    This report contains articles concerning the nuclear developments of the following countries: (1) China; (2) Japan, North Korea, South Korea; (3) Bulgaria; (4) Argentina, Brazil, Honduras; (5) India, Iran, Pakistan, Syria; (6) Soviet Union; and (7) France, Germany, Turkey.

  2. Nuclear material operations manual

    SciTech Connect (OSTI)

    Tyler, R.P.

    1981-02-01T23:59:59.000Z

    This manual provides a concise and comprehensive documentation of the operating procedures currently practiced at Sandia National Laboratories with regard to the management, control, and accountability of nuclear materials. The manual is divided into chapters which are devoted to the separate functions performed in nuclear material operations-management, control, accountability, and safeguards, and the final two chapters comprise a document which is also issued separately to provide a summary of the information and operating procedures relevant to custodians and users of radioactive and nuclear materials. The manual also contains samples of the forms utilized in carrying out nuclear material activities. To enhance the clarity of presentation, operating procedures are presented in the form of playscripts in which the responsible organizations and necessary actions are clearly delineated in a chronological fashion from the initiation of a transaction to its completion.

  3. INL '@work' Nuclear Engineer

    ScienceCinema (OSTI)

    McLean, Heather

    2013-05-28T23:59:59.000Z

    Heather MacLean talks about her job as a Nuclear Engineer for Idaho National Laboratory. For more information about INL careers, visit http://www.facebook.com/idahonationallaboratory.

  4. Nuclear Fuel Reprocessing

    SciTech Connect (OSTI)

    Michael F. Simpson; Jack D. Law

    2010-02-01T23:59:59.000Z

    This is an a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. No formal abstract was required for the article. The full article will be attached.

  5. National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Nuclear Security Administration FY 2011 - FY 2015 Budget Outlook Managing the NNSA 4.0% Science, Technology & Engineering 14.5% Stockpile Support 17.9% Preventing the Spread of...

  6. Management of Nuclear Materials

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2009-08-17T23:59:59.000Z

    To establish requirements for the lifecycle management of DOE owned and/or managed accountable nuclear materials. Cancels DOE O 410.2. Admin Chg 1 dated 4-10-2014, cancels DOE O 410.2.

  7. PDFs for nuclear targets

    E-Print Network [OSTI]

    Karol Kovarik

    2010-06-25T23:59:59.000Z

    Understanding nuclear effects in parton distribution functions (PDF) is an essential component needed to determine the strange and anti-strange quark contributions in the proton. In addition Nuclear Parton Distribution Functions (NPDF) are critically important for any collider experiment with nuclei (e.g. RHIC, ALICE). Here two next-to-leading order chi^2-analyses of NPDF are presented. The first uses neutral current charged-lepton Deeply Inelastic Scattering (DIS) and Drell-Yan data for several nuclear targets and the second uses neutrino-nucleon DIS data. We compare the nuclear corrections factors (F_2^Fe/F_2^D) for the charged-lepton data with other results from the literature. In particular, we compare and contrast fits based upon the charged-lepton DIS data with those using neutrino-nucleon DIS data.

  8. International Nuclear Security

    SciTech Connect (OSTI)

    Doyle, James E. [Los Alamos National Laboratory

    2012-08-14T23:59:59.000Z

    This presentation discusses: (1) Definitions of international nuclear security; (2) What degree of security do we have now; (3) Limitations of a nuclear security strategy focused on national lock-downs of fissile materials and weapons; (4) What do current trends say about the future; and (5) How can nuclear security be strengthened? Nuclear security can be strengthened by: (1) More accurate baseline inventories; (2) Better physical protection, control and accounting; (3) Effective personnel reliability programs; (4) Minimize weapons-usable materials and consolidate to fewer locations; (5) Consider local threat environment when siting facilities; (6) Implement pledges made in the NSS process; and (7) More robust interdiction, emergency response and special operations capabilities. International cooperation is desirable, but not always possible.

  9. Western Nuclear Science Alliance

    SciTech Connect (OSTI)

    Steve Reese; George Miller; Stephen Frantz; Denis Beller; Denis Beller; Ed Morse; Melinda Krahenbuhl; Bob Flocchini; Jim Elliston

    2010-12-07T23:59:59.000Z

    The primary objective of the INIE program is to strengthen nuclear science and engineering programs at the member institutions and to address the long term goal of the University Reactor Infrastructure and Education Assistance Program.

  10. Nuclear Science Center - 4 

    E-Print Network [OSTI]

    Unknown

    2009-01-01T23:59:59.000Z

    How does the American public assess risk when it comes to national security issues? This paper addresses this question by analyzing variation in citizen probability assessments of the terrorism risk of nuclear power plants. Drawing on the literature...

  11. Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2006-06-12T23:59:59.000Z

    The directive provides supplemental details to support the requirements of DOE O 452.2C, Nuclear Explosive Safety, dated 6-12-06. Canceled by DOE M 452.2-1A.

  12. CONSTRUCTION OF NUCLEAR POWER PLANTS

    E-Print Network [OSTI]

    CONSTRUCTION OF NUCLEAR POWER PLANTS A Workshop on "NUCLEAR ENERGY RENAISSANCE" Addressing WAS DEEPLY INVOLVED IN ALMOST EVERY ASPECT OF BUILDING THE PLANTS THROUGH · Quality Assurance · Nuclear IN CONSTRUCTION OF ST. LUCIE-2 #12;LESSONS LEARNED FROM St. Lucie-2 NUCLEAR POWER PLANTS CAN BE BUILT

  13. Reactor- Nuclear Science Center 

    E-Print Network [OSTI]

    Unknown

    2011-08-17T23:59:59.000Z

    A COMPARISON OF NUCLEAR REACTOR CONTROL ROOM DISPLAY PANELS A Thesis by FRANCES RENAE BOWERS Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE May 1988... Major Subject: Industrial Engineering A COMPARISON OF NUCLEAR REACTOR CONTROL ROOM DISPLAY PANELS A Thesis by FRANCES RENAE BOWERS Approved as to style and content by: Rod er . oppa (Cha' of 'ttee) R. Quinn Brackett (Member) rome . Co gleton...

  14. Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel

    E-Print Network [OSTI]

    Ludewigt, Bernhard A

    2011-01-01T23:59:59.000Z

    10-01096) Journal of Nuclear Technology, in Press. [46] G.W.Library for Nuclear Science and Technology,” Nuclear Datacalculations,” Nuclear Data for Science and Technology

  15. After nuclear war - a nuclear winter

    SciTech Connect (OSTI)

    Tangley, L.

    1984-01-01T23:59:59.000Z

    The environmental and biological consequences of nuclear war were discussed by more than 100 eminent biologists, physicists and atmospheric scientists at the recent World after Nuclear War conference. The long-term effects were determined to be worse than the well-known immediate effects. They predicted that 225 million tons of smoke would be generated within a few days in their baseline scenario. As a result, the amount of sunlight reaching the earth would be reduced to a few percent of normal and temperatures would fall to -23/sup 0/C. About 30% of the northern middle latitudes would receive more than 250 rads radiation dose for several months and about 50% of the land area would receive more than 100 rads. Dangerous levels of solar ultraviolet light would burn through the atmosphere. It was also determined that these effects would be felt in the southern hemisphere. Those who survived the blast, fire and prompt radiation would face starvation from shutdown of plant photosynthesis and inhibition of phytoplankton photosynthesis. Huge wildfires and acid rains would stress any surviving plants and animals. Conference participants agreed that scientists had taken a new and significant step toward understanding the full consequences of nuclear war.

  16. Superpower nuclear minimalism

    SciTech Connect (OSTI)

    Graben, E.K.

    1992-01-01T23:59:59.000Z

    During the Cold War, the United States and the Soviet Union competed in building weapons -- now it seems like America and Russia are competing to get rid of them the fastest. The lengthy process of formal arms control has been replaced by exchanges of unilateral force reductions and proposals for reciprocal reductions not necessarily codified by treaty. Should superpower nuclear strategies change along with force postures President Bush has yet to make a formal pronouncement on post-Cold War American nuclear strategy, and it is uncertain if the Soviet/Russian doctrine of reasonable sufficiency formulated in the Gorbachev era actually heralds a change in strategy. Some of the provisions in the most recent round of unilateral proposals put forth by Presidents Bush and Yeltsin in January 1992 are compatible with a change in strategy. Whether such a change has actually occurred remains to be seen. With the end of the Cold War and the breakup of the Soviet Union, the strategic environment has fundamentally changed, so it would seem logical to reexamine strategy as well. There are two main schools of nuclear strategic thought: a maximalist school, mutual assured destruction (MAD) which emphasizes counterforce superiority and nuclear war- fighting capability, and a MAD-plus school, which emphasizes survivability of an assured destruction capability along with the ability to deliver small, limited nuclear attacks in the event that conflict occurs. The MAD-plus strategy is based on an attempt to conventionalize nuclear weapons which is unrealistic.

  17. The elements of nuclear power

    SciTech Connect (OSTI)

    Bennet, D.J.; Thomson, J.R.

    1989-01-01T23:59:59.000Z

    An introduction to the principles of nuclear fission power generation. Describes the physical processes which occur in a nuclear reactor and discusses the theory behind the calculations. Also covers heat transfer in reactors, thermodynamic power cycles, reactor operators, and radiation shielding. Material covered includes topics on the effects of nuclear radiation on humans, the safety of nuclear reactors and of those parts of the nuclear fuel cycle which deal with fuel element manufacture and the reprocessing of irradiated fuel.

  18. Panel report: nuclear physics

    SciTech Connect (OSTI)

    Carlson, Joseph A [Los Alamos National Laboratory; Hartouni, Edward P [LLNL

    2010-01-01T23:59:59.000Z

    Nuclear science is at the very heart of the NNSA program. The energy produced by nuclear processes is central to the NNSA mission, and nuclear reactions are critical in many applications, including National Ignition Facility (NIF) capsules, energy production, weapons, and in global threat reduction. Nuclear reactions are the source of energy in all these applications, and they can also be crucial in understanding and diagnosing the complex high-energy environments integral to the work of the NNSA. Nuclear processes are complex quantum many-body problems. Modeling and simulation of nuclear reactions and their role in applications, coupled tightly with experiments, have played a key role in NNSA's mission. The science input to NNSA program applications has been heavily reliant on experiment combined with extrapolations and physical models 'just good enough' to provide a starting point to extensive engineering that generated a body of empirical information. This body of information lacks the basic science underpinnings necessary to provide reliable extrapolations beyond the domain in which it was produced and for providing quantifiable error bars. Further, the ability to perform additional engineering tests is no longer possible, especially those tests that produce data in the extreme environments that uniquely characterize these applications. The end of testing has required improvements to the predictive capabilities of codes simulating the reactions and associated applications for both well known and well characterized cases as well as incompletely known cases. Developments in high performance computing, computational physics, applied mathematics and nuclear theory have combined to make spectacular advances in the theory of fission, fusion and nuclear reactions. Current research exploits these developments in a number of Office of Science and NNSA programs, and in joint programs such as the SciDAC (Science Discovery through Advanced Computing) that supports the project Building a Universal Nuclear Energy Density Fuctional whose goals are to provide the unified approach to calculating the properties of nuclei. The successful outcome of this, and similar projects is a first steps toward a predictive nuclear theory based on fundamental interactions between constituent nucleons. The application of this theory to the domain of nuclei important for national security missions will require computational resources at the extreme scale, beyond what will be available in the near term future.

  19. Model Action Plan for Nuclear Forensics and Nuclear Attribution

    SciTech Connect (OSTI)

    Dudder, G B; Niemeyer, S; Smith, D K; Kristo, M J

    2004-03-01T23:59:59.000Z

    Nuclear forensics and nuclear attribution have become increasingly important tools in the fight against illegal trafficking in nuclear and radiological materials. This technical report documents the field of nuclear forensics and nuclear attribution in a comprehensive manner, summarizing tools and procedures that have heretofore been described independently in the scientific literature. This report also provides national policy-makers, decision-makers, and technical managers with guidance for responding to incidents involving the interdiction of nuclear and radiological materials. However, due to the significant capital costs of the equipment and the specialized expertise of the personnel, work in the field of nuclear forensics has been restricted so far to a handful of national and international laboratories. In fact, there are a limited number of specialists who have experience working with interdicted nuclear materials and affiliated evidence. Most of the laboratories that have the requisite equipment, personnel, and experience to perform nuclear forensic analysis are participants in the Nuclear Smuggling International Technical Working Group or ITWG (see Section 1.8). Consequently, there is a need to disseminate information on an appropriate response to incidents of nuclear smuggling, including a comprehensive approach to gathering evidence that meets appropriate legal standards and to developing insights into the source and routes of nuclear and radiological contraband. Appendix A presents a ''Menu of Options'' for other Member States to request assistance from the ITWG Nuclear Forensics Laboratories (INFL) on nuclear forensic cases.

  20. Nuclear explosive safety study process

    SciTech Connect (OSTI)

    NONE

    1997-01-01T23:59:59.000Z

    Nuclear explosives by their design and intended use require collocation of high explosives and fissile material. The design agencies are responsible for designing safety into the nuclear explosive and processes involving the nuclear explosive. The methodology for ensuring safety consists of independent review processes that include the national laboratories, Operations Offices, Headquarters, and responsible Area Offices and operating contractors with expertise in nuclear explosive safety. A NES Study is an evaluation of the adequacy of positive measures to minimize the possibility of an inadvertent or deliberate unauthorized nuclear detonation, high explosive detonation or deflagration, fire, or fissile material dispersal from the pit. The Nuclear Explosive Safety Study Group (NESSG) evaluates nuclear explosive operations against the Nuclear Explosive Safety Standards specified in DOE O 452.2 using systematic evaluation techniques. These Safety Standards must be satisfied for nuclear explosive operations.

  1. Tennessee Nuclear Profile - Watts Bar Nuclear Plant

    U.S. Energy Information Administration (EIA) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5 Tables July 1996 Energy Information Administration Office ofthroughYear JanYear Jan Feb Mar Apr May Jun Jul AugSameWatts Bar Nuclear

  2. Chernobyl Nuclear Accident | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTubahq.na.gov Office of theNuclear

  3. International Nuclear Security | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated Codes | National NuclearInterlibrary LoanSafeguards | National

  4. Nuclear Controls | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOE Directives,838 NovemberNuclearControls

  5. Nuclear Detonation Detection | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOE Directives,838Nuclear Detection

  6. Nuclear Forensics | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOE Directives,838NuclearForensics |

  7. Nuclear Incident Team | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOE Directives,838NuclearForensics

  8. Nuclear Material Removal | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOERecovery Nuclear Material

  9. Nuclear Operations | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOERecovery NuclearSensor Nodes

  10. Nuclear Security 101 | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOERecovery NuclearSensorAdvisors New

  11. Nuclear Security Enterprise | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOERecovery NuclearSensorAdvisors

  12. Nuclear Verification | National Nuclear Security Administration

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOERecoveryNuclear

  13. defense nuclear security | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA Approved: 5-13-14Russian NuclearNational5/%2A en Office ofcontractingcyber

  14. Sandia National Laboratories: Nuclear Energy Publications

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Publications Nuclear Energy Publications Nuclear Energy Safety Fact Sheets Assuring Safe Transportation of Nuclear and Hazardous Materials Human Reliability Assessment (HRA)...

  15. National Nuclear Security Administration Supplemental Listing...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    National Nuclear Security Administration Supplemental Listing of Directives Affecting Nuclear Safety Requirements Last Updated 062014 U.S. DEPARTMENT OF ENERGY National Nuclear...

  16. Nuclear Resonance Fluorescence for Materials Assay

    E-Print Network [OSTI]

    Quiter, Brian J.

    2010-01-01T23:59:59.000Z

    et al. “Investigation of Nuclear Structure by Resonance1996, pp. G. Warren et al. “Nuclear Resonance Fluorescenceof 235U” IEEE Nuclear Science Symposium 2006, pp. 914. W.

  17. Nuclear Resonance Fluorescence for Materials Assay

    E-Print Network [OSTI]

    Quiter, Brian

    2010-01-01T23:59:59.000Z

    clandestine material with nuclear resonance fluorescence”.E. Norman, UC Berkeley Dept. of Nuclear Engineering, privatepp. 349. G. Warren et al. “Nuclear Resonance Fluorescence of

  18. NUCLEAR SCIENCE ANNUAL REPORT 1977-1978

    E-Print Network [OSTI]

    Schroeder, L.S.

    2011-01-01T23:59:59.000Z

    A Relation Between Nuclear Dynamics and the RenormalizationMultiplicity Distributions in Nuclear Collision M. GyulassyHigh Energy Nuclear Collisions in the Resonance Dominated

  19. Office of Nuclear Facility Safety Programs

    Broader source: Energy.gov [DOE]

    The Office of Nuclear Facility Safety Programs establishes nuclear safety requirements related to safety management programs that are essential to the safety of DOE nuclear facilities.

  20. Nuclear Safety Information Dashboard | Department of Energy

    Office of Environmental Management (EM)

    Nuclear Safety Information Dashboard Nuclear Safety Information Dashboard The Nuclear Safety Information (NSI) Dashboard provides a new user interface to the Occurrence Reporting...

  1. Nuclear Science and Engineering | ornl.gov

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Nuclear Science Engineering SHARE Nuclear Science and Engineering The Nuclear Science and Engineering Directorate (NSED) at Oak Ridge National Laboratory (ORNL) is committed to...

  2. Nuclear Explosive Safety Manual - DOE Directives, Delegations...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    1A Admin Chg 1, Nuclear Explosive Safety Manual by Carl Sykes Functional areas: Administrative Change, Defense Nuclear Facility Safety and Health Requirement, Nuclear Safety,...

  3. CRAD, Nuclear Facility Construction - Structural Concrete, May...

    Broader source: Energy.gov (indexed) [DOE]

    CRAD, Nuclear Facility Construction - Structural Concrete, May 29, 2009 CRAD, Nuclear Facility Construction - Structural Concrete, May 29, 2009 May 29, 2009 Nuclear Facility...

  4. Nuclear fuel element

    DOE Patents [OSTI]

    Armijo, Joseph S. (Saratoga, CA); Coffin, Jr., Louis F. (Schenectady, NY)

    1983-01-01T23:59:59.000Z

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

  5. Nuclear fuel element

    DOE Patents [OSTI]

    Armijo, Joseph S. (Saratoga, CA); Coffin, Jr., Louis F. (Schenectady, NY)

    1980-04-29T23:59:59.000Z

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

  6. Pediatric nuclear medicine

    SciTech Connect (OSTI)

    Not Available

    1986-01-01T23:59:59.000Z

    This symposium presented the latest techniques and approaches to the proper medical application of radionuclides in pediatrics. An expert faculty, comprised of specialists in the field of pediatric nuclear medicine, discussed the major indications as well as the advantages and potential hazards of nuclear medicine procedures compared to other diagnostic modalities. In recent years, newer radiopharmaceuticals labeled with technetium-99m and other short-lived radionuclides with relatively favorable radiation characteristics have permitted a variety of diagnostic studies that are very useful clinically and carry a substantially lower radiation burden then many comparable X-ray studies. This new battery of nuclear medicine procedures is now widely available for diagnosis and management of pediatric patients. Many recent research studies in children have yielded data concerning the effacacy of these procedures, and current recommendations will be presented by those involved in conducting such studies. Individual papers are processed separately for the Energy Data Base.

  7. Nanofluid heat transfer enhancement for nuclear reactor applications

    E-Print Network [OSTI]

    Buongiorno, Jacopo

    Colloidal dispersions of nanoparticles are known as `nanofluids'. Such engineered fluids offer the potential for enhancing heat transfer, particularly boiling heat transfer, while avoiding the drawbacks (i.e., erosion, ...

  8. Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

    SciTech Connect (OSTI)

    Su'ud, Zaki; Anshari, Rio [Nuclear and Biophysics Research Group, Dept. of Physics, Bandung Institute of Technology, Jl.Ganesha 10, Bandung, 40132 (Indonesia)

    2012-06-06T23:59:59.000Z

    Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

  9. Security and Control of Nuclear Explosives and Nuclear Weapons

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2001-12-17T23:59:59.000Z

    This directive establishes requirements and responsibilities to prevent the deliberate unauthorized use of U.S. nuclear explosives and U.S. nuclear weapons. Cancels DOE O 452.4.

  10. Determining Factors Influencing Nuclear Envelope and Nuclear Pore Complex Structure.

    E-Print Network [OSTI]

    Gouni, Sushanth

    2013-02-04T23:59:59.000Z

    The cell’s nuclear envelope (NE) has pores that are stabilized by nuclear pore complexes (NPC), large proteinaceous structures whose function is to mediate transport between the nucleus and cytoplasm. Although the transport process is well studied...

  11. NCSRNCSR ""DEMOKRITOSDEMOKRITOS"" INSTITUTE OF NUCLEAR TECHNOLOGY & RADIATION PROTECTIONINSTITUTE OF NUCLEAR TECHNOLOGY & RADIATION PROTECTION

    E-Print Network [OSTI]

    PROTECTIONINSTITUTE OF NUCLEAR TECHNOLOGY & RADIATION PROTECTION ·· ENVIRONMENTAL RESEARCH LABORATORYENVIRONMENTAL·· NCSRNCSR ""DEMOKRITOSDEMOKRITOS"" ·· INSTITUTE OF NUCLEAR TECHNOLOGY & RADIATION

  12. Nuclear Masses in Astrophysics

    E-Print Network [OSTI]

    Christine Weber; Klaus Blaum; Hendrik Schatz

    2008-12-09T23:59:59.000Z

    Among all nuclear ground-state properties, atomic masses are highly specific for each particular combination of N and Z and the data obtained apply to a variety of physics topics. One of the most crucial questions to be addressed in mass spectrometry of unstable radionuclides is the one of understanding the processes of element formation in the Universe. To this end, accurate atomic mass values of a large number of exotic nuclei participating in nucleosynthesis are among the key input data in large-scale reaction network calculations. In this paper, a review on the latest achievements in mass spectrometry for nuclear astrophysics is given.

  13. Neutrinos in Nuclear Physics

    E-Print Network [OSTI]

    R. D. McKeown

    2014-12-03T23:59:59.000Z

    Since the discovery of nuclear beta decay, nuclear physicists have studied the weak interaction and the nature of neutrinos. Many recent and current experiments have been focused on the elucidation of neutrino oscillations and neutrino mass. The quest for the absolute value of neutrino mass continues with higher precision studies of the tritium beta decay spectrum near the endpoint. Neutrino oscillations are studied through measurements of reactor neutrinos as a function of baseline and energy. And experiments searching for neutrinoless double beta decay seek to discover violation of lepton number and establish the Majorana nature of neutrino masses.

  14. Neutrinos in Nuclear Physics

    E-Print Network [OSTI]

    McKeown, R D

    2014-01-01T23:59:59.000Z

    Since the discovery of nuclear beta decay, nuclear physicists have studied the weak interaction and the nature of neutrinos. Many recent and current experiments have been focused on the elucidation of neutrino oscillations and neutrino mass. The quest for the absolute value of neutrino mass continues with higher precision studies of the tritium beta decay spectrum near the endpoint. Neutrino oscillations are studied through measurements of reactor neutrinos as a function of baseline and energy. And experiments searching for neutrinoless double beta decay seek to discover violation of lepton number and establish the Majorana nature of neutrino masses.

  15. Clustering in nuclear environment

    E-Print Network [OSTI]

    G. Röpke

    2014-08-12T23:59:59.000Z

    The properties of few-body clusters (mass number $A \\le 4$) are modified if they are immersed in a nuclear medium. In particular, Pauli blocking that reflects the antisymmetrization of the many-body wave function is responsible for the medium modification of light clusters and the dissolution with increasing density. A more consistent description is given with takes also the contribution of correlations in the continuum into account. The relation between cluster formation in warm dense matter and in nuclear structure is discussed.

  16. Nuclear medicine imaging system

    DOE Patents [OSTI]

    Bennett, G.W.; Brill, A.B.; Bizais, Y.J.C.; Rowe, R.W.; Zubal, I.G.

    1983-03-11T23:59:59.000Z

    It is an object of this invention to provide a nuclear imaging system having the versatility to do positron annihilation studies, rotating single or opposed camera gamma emission studies, and orthogonal gamma emission studies. It is a further object of this invention to provide an imaging system having the capability for orthogonal dual multipinhole tomography. It is another object of this invention to provide a nuclear imaging system in which all available energy data, as well as patient physiological data, are acquired simultaneously in list mode.

  17. Nuclear Science at NERSC

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItemResearch > TheNuclear Astrophysics One of the great scientificResearchNuclear

  18. Nuclear Workforce Initiative - SRSCRO

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItemResearch > TheNuclear Astrophysics One of the greatNuclear

  19. Importance of Delayed Neutrons on the Coupled Neutronic-Thermohydraulic Stability of a Natural Circulation Heavy Water-Moderated Boiling Light Water-Cooled Reactor

    SciTech Connect (OSTI)

    Nayak, A.K. [Bhaha Atomic Research Centre (India); Aritomi, M. [Tokyo Institute of Technology (Japan); Raj, V. Venkat [Bhaha Atomic Research Centre (India)

    2001-07-15T23:59:59.000Z

    The coupled neutronic-thermohydraulic stability characteristics of a natural circulation heavy water-moderated boiling light water-cooled reactor was investigated analytically considering the effects of prompt and delayed neutrons. For this purpose, the reactor considered is the Indian Advanced Heavy Water Reactor. The analytical model considers a point kinetics model for the neutron dynamics, a homogeneous two-phase flow model for the coolant thermal hydraulics, and a lumped heat transfer model for the fuel thermal dynamics. A higher mode of oscillation having a frequency much greater than the density-wave oscillation frequency was observed if prompt neutrons alone were considered. The occurrence of a higher mode of oscillation was found to be dependent on the concentration of delayed neutrons, the void reactivity coefficient, and the fuel time constant. The core inlet subcooling is found to have different effects on the decay ratio of the fundamental and higher modes of oscillations. The influences of void reactivity coefficient and fuel time constant on the fundamental and higher modes of oscillations were also found to be opposite in nature.

  20. Nuclear effects in Neutrino Nuclear Cross-sections

    E-Print Network [OSTI]

    S. K. Singh; M. Sajjad Athar

    2007-10-24T23:59:59.000Z

    Nuclear effects in the quasielastic and inelastic scattering of neutrinos(antineutrinos) from nuclear targets have been studied. The calculations are done in the local density approximation which take into account the effect of nucleon motion as well as renormalisation of weak transition strengths in the nuclear medium. The inelastic reaction leading to production of pions is calculated in a $\\Delta$ dominance model taking into account the renormalization of $\\Delta$ properties in the nuclear medium.