National Library of Energy BETA

Sample records for boiling nuclear superheater

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

    E-Print Network [OSTI]

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

    2011-01-01

    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 ...

  2. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal Gas &SCE-SessionsSouthReport for the WeldonB10081278MaywoodWayne Site83 UMTRCA

  3. The role of surface conditions in nucleate boiling

    E-Print Network [OSTI]

    Griffith, P.

    1958-01-01

    Nucleation from a single cavity has been stuied indicating that cavity gemtry is aportant in two ways. The mouth diameter determines the superheat nmeded to initiate boiling and its shape determines its stability one boiling ...

  4. Superheat Control: A Hybrid Approach 

    E-Print Network [OSTI]

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

    2009-01-01

    , poorly tuned EEVs can still exhibit undesirable behavior, and frequent valve adjustments raise concerns about device longevity. In this work, we propose a cascaded control approach, which regulates evapo- rator pressure and superheat and is achieved... with a feedback control device that uses a hybrid of mechanical and electronic feedback. Analysis of the fundamental dynamic behavior of evapo- rator superheat motivates this approach, while experimental evaluation of two separate systems demonstrates...

  5. Alternate Materials for Recovery Boiler Superheater Tubes

    SciTech Connect (OSTI)

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

    2009-01-01

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1986-09-01

    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.

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

    SciTech Connect (OSTI)

    1995-08-01

    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.

  8. 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

    2010-01-01

    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.

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

    SciTech Connect (OSTI)

    Anh Bui; Nam Dinh; Brian Williams

    2013-09-01

    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.

  10. Saturated nucleate pool boiling of oxygen under magnetically-enhanced effective gravity

    E-Print Network [OSTI]

    T. A. Corcovilos; M. E. Turk; D. M. Strayer; N. N. Asplund; N. -C. Yeh

    2007-02-01

    We investigate the effect of enhancing gravity on saturated nucleate pool boiling of oxygen for effective gravities of 1g, 6.0g, and 16g (g=9.8 m/s^2) at a saturation pressure of 760 torr and for heat fluxes of 10 ~ 3000 W/m^2. The effective gravity on the oxygen is increased by applying a magnetic body force generated by a superconducting solenoid. We measure the heater temperature (expressed as a reduced superheat) as a function of heat flux and fit this data to a piecewise power-law/linear boiling curve. At low heat flux (<400 W/m^2) the superheat is proportional to the cube root of the heat flux. At higher heat fluxes, the superheat is a linear function of the heat flux. To within statistical uncertainties, which are limited by variations among experimental runs, we find no variation of the boiling curve over our applied gravity range.

  11. 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.

    1982-03-01

    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).

  12. EBR-II Superheater Duplex Tube Examination

    SciTech Connect (OSTI)

    Daniel M. Wachs; Dennis D. Keiser; Douglas L. Porter; Naoyuki Kisohara

    2008-12-01

    After 30 years of operation, the Experimental Breeder Reactor II (EBR-II) Superheater 710 at Argonne National Laboratory-West (now Idaho National Laboratory) was decommissioned. As part of its post-service examination, four duplex tube sections were removed and Charpy impact testing was performed to characterize the crack arresting ability of nickel-bonded tube interfaces. Scanning electron microscopy (SEM) examination was also performed to characterize and identify changes in bond material microstructure. From room temperature to 400 degrees C, all samples demonstrated ductility and crack-stopping ability similar to that exhibited by beginning-of-life samples. However, at low temperature (-5 degrees C), samples removed from the lower region of the superheater (near the sodium inlet) failed while those from the upper region (near the sodium outlet) did not. SEM analysis revealed that all the tube-tube interfaces showed evidence of iron diffusion into the nickel braze, which resulted in the formation of a multiphase diffusion structure. Yet, significant void formation was only observed in the bond layer of the tubes removed from the lower region. This may be due to a change in the crystal microstructure of one of the phases within the bond layer that occurs in the 350 to 450 degrees C temperature range, which results in a lower density and the formation of porosity. Apparently, only the samples from the higher temperature region were exposed to this transition temperature, and the resulting large voids that developed acted as stress concentrators that led to low-temperature embrittlement and failure of the Charpy impact specimens.

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

    SciTech Connect (OSTI)

    Liu, C.; Theofanous, T. G.

    2000-08-29

    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.

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

    SciTech Connect (OSTI)

    Sharp, William

    2011-12-01

    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

  15. Film boiling of R-11 on liquid metal surfaces

    SciTech Connect (OSTI)

    Greene, G.A.; Irvine, T.F. Jr.

    1986-01-01

    An interesting problem is the effect of an immiscible liquid heating surface on the process of film boiling. Such surfaces raise questions concerning interface stability to disturbances, effects of gas bubbling, and vapor explosions in layered systems. The specific motivation for this study was to investigate film boiling from a liquid surface with application to cooling of molten reactor core debris by an overlying pool of reactor coolant. To investigate this phenomenon, and apparatus consisting of a nominal six-inch diameter steel vessel to hold the liquid metal and boiling fluid was constructed; coolant reservoirs, heaters, controllers, and allied instrumentation were attached. A transient energy balance was performed on the liquid metal pool by a submerged assembly of microthermocouples in the liquid metal and an array of thermocouples on the wall of the test vessel. The thermocouple data were used to determine the boiling heat flux as well as the boiling superheat. On an average basis, the deviation between the prediction of the Berenson model and the experimental data was less than one percent when Berenson was corrected for thermal radiation effects. Evidence from visualization tests of R-11 in film boiling over molten metal pools to superheats in excess of 600 K supports this conclusion. 13 refs.

  16. 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)

    2010-01-12

    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.

  17. Void volumes in subcooled boiling systems

    E-Print Network [OSTI]

    Griffith, P.

    1958-01-01

    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 ...

  18. 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 Goodman Introduction In the papermaking industry, black liquor recovery boilers burn black liquor into the superheater region of the boiler, where the salt-deposit, or smelt, forms a scale on the superheater tubes.1

  19. Sulfur containing sediments influenced wear of superheater tubes

    SciTech Connect (OSTI)

    Yoffe, P. [Israel Electric Corp. Ltd., Haifa (Israel)

    1996-10-01

    The failed superheater tube was investigated at the point of failure and at a distance of 25 cm. At both of the cross sections two layers of sediment were detected. The external one contains Na2SO4 and Na3VO3 in concentration relation 6.4:1 approximately. The internal layer that situated on the tube surface consisted of V2O3 mainly (about 70%) and sodium pyrosulfates Na2SO3{center_dot}xSO3. The tube surface was covered with usual passive oxides such as Cr2O3, Fe2CrO4, and magnetite. In the failed section the last layer contained iron and chromium sulfides. Thinning of the wall was detected in both sections, but it was more evident in the failed section. Thermodynamic and kinetic estimations of the results were carried out. The following failure mechanism was proposed: (1) at 460 C the pyrosulfate dissociates with SO3 emanation, during SO3 interaction with steel both passive oxides and iron sulfide are produced simultaneously, sulfides damage the passive (protective) film; (2) in vicinity of partition the temperature increases and the structure of steel is changed, SO3 penetration along grain boundaries is possible, and it leads to embrittlement of the steel through sulfidization; (3) the embrittled layer is stripped off by falling ash deposit and collapsed passive film, thinning of the tube is accelerated at such a region. So, sulfur containing sediment accelerates the wear of the tube. This acceleration is strong especially in the vicinity of the boiler walls.

  20. 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

  1. Heat transport in boiling turbulent Rayleigh-B\\'{e}nard convection

    E-Print Network [OSTI]

    Lakkaraju, Rajaram; Oresta, Paolo; Verzicco, Roberto; Lohse, Detlef; Prosperetti, Andrea

    2014-01-01

    Boiling is an extremely effective way to promote heat transfer from a hot surface to a liquid due to several mechanisms many of which are not understood in quantitative detail. An important component of the overall process is that the buoyancy of the bubbles compounds with that of the liquid to give rise to a much enhanced natural convection. In this paper we focus specifically on this enhancement and present a numerical study of the resulting two-phase Rayleigh-B\\'enard convection process. We make no attempt to model other aspects of the boiling process such as bubble nucleation and detachment. We consider a cylindrical cell with a diameter equal to its height. The cell base and top are held at temperatures above and below the boiling point of the liquid, respectively. By keeping the temperature difference constant and changing the liquid pressure we study the effect of the liquid superheat in a Rayleigh number range that, in the absence of boiling, would be between $2\\times10^6$ and $5\\times10^9$. We find a...

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

    SciTech Connect (OSTI)

    Sabau, Adrian S; Wright, Ian G.; Shingledecker, John P.; Tortorelli, Peter F

    2014-01-01

    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.

  3. A study of electrowetting-assisted boiling

    E-Print Network [OSTI]

    Bralower, Harrison L. (Harrison Louis)

    2011-01-01

    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 ...

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

    SciTech Connect (OSTI)

    Terry C. Totemeier

    2006-12-01

    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.

  5. Bubble growth rates in boiling

    E-Print Network [OSTI]

    Griffith, P.

    1956-01-01

    The conditions determining the growth rate of a bubble on a surface in boiling are considered and a mathematical model framed in the light of these conditions. The growth rate is then calculated for bubbles growing under ...

  6. Subcooled flow boiling of fluorocarbons

    E-Print Network [OSTI]

    Murphy, Richard Walter

    1971-01-01

    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 ...

  7. Numerical Simulation of Boiling Heat Transfer by Transient Heating *@--i"OE`H@j@@@"`@Zi@Oi"OE`Hj@@@"`@SZR@vi"OE`Hj

    E-Print Network [OSTI]

    Maruyama, Shigeo

    with macrolayer model of Maruyama, we simulated the transient boiling curve for water and fluorinert FC-72(C6F14 boiling processes are very important in steel production and safety evaluations in nuclear reactors conduction within the heater. The transient boiling curves for water and FC-72 were predicted. The transient

  8. Correlating Radioactive Material to Sea Surface Temperature off the Coast of Japan: The Fukushima Daiichi Nuclear Disaster

    E-Print Network [OSTI]

    Gilbes, Fernando

    nuclear reactions to constantly boil water; the resulting steam is used to drive turbines and generate

  9. Evaluation of the economic simplified boiling water reactor human reliability analysis using the SHARP framework

    E-Print Network [OSTI]

    Dawson, Phillip Eng

    2007-01-01

    General Electric plans to complete a design certification document for the Economic Simplified Boiling Water Reactor to have the new reactor design certified by the United States Nuclear Regulatory Commission. As part of ...

  10. Dispersed flow film boiling

    E-Print Network [OSTI]

    Yoder, Graydon L.

    1980-01-01

    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 ...

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

    SciTech Connect (OSTI)

    Reichelt, J.; Pfrang-Stotz, G.; Bergfeldt, B.; Seifert, H.; Knapp, P.

    2013-01-15

    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.

  12. Re-evaluation of superheat conditions postulated in NRC Information Notice 84-90

    SciTech Connect (OSTI)

    Alsammarae, A.; Kruger, D.; Beutel, D.; Spisak, M.

    1994-08-01

    Information Notice 84-90, ''Main Steam Line Break Effect on Environmental Qualification of Equipment,'' describes a potential problem regarding existing plant analyses and Equipment Qualification (EQ) related to a postulated Main Steam Line Break (MSLB) with releases of superheated stream. This notice states that certain methodologies for computing mass and energy releases for a postulated MSLB did not account for heat transfer from the steam generator tube bundles if they were uncovered. Due to this potential change in the original environmental analysis, the EQ of various components may not consider the thermal environment which could result from superheated steam. Subsequent technical assessments may determine that the existing qualification basis for equipment/components does not envelop the postulated superheat condition. Corrective actions need to be taken to demonstrate that the affected equipment is qualified.

  13. Pool boiling on nano-finned surfaces 

    E-Print Network [OSTI]

    Sriraman, Sharan Ram

    2008-10-10

    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...

  14. Multifractal Analysis of Chaotic Flashing-Induced Instabilities in Boiling Channels in the Natural-Circulation CIRCUS Facility

    E-Print Network [OSTI]

    Demazière, Christophe

    . INTRODUCTION Instability of forced-circulation boiling water reac- tors ~BWRs!, which is manifested by selfMultifractal Analysis of Chaotic Flashing-Induced Instabilities in Boiling Channels in the Natural University of Technology, Department of Physics of Nuclear Reactors, Delft, 2629 BJ, The Netherlands Received

  15. Modeling acid-gas generation from boiling chloride brines

    SciTech Connect (OSTI)

    Zhang, Guoxiang; Spycher, Nicolas; Sonnenthal, Eric; Steefel, Carl

    2009-11-16

    This study investigates the generation of HCl and other acid gases from boiling calcium chloride dominated waters at atmospheric pressure, primarily using numerical modeling. The main focus of this investigation relates to the long-term geologic disposal of nuclear waste at Yucca Mountain, Nevada, where pore waters around waste-emplacement tunnels are expected to undergo boiling and evaporative concentration as a result of the heat released by spent nuclear fuel. Processes that are modeled include boiling of highly concentrated solutions, gas transport, and gas condensation accompanied by the dissociation of acid gases, causing low-pH condensate. Simple calculations are first carried out to evaluate condensate pH as a function of HCl gas fugacity and condensed water fraction for a vapor equilibrated with saturated calcium chloride brine at 50-150 C and 1 bar. The distillation of a calcium-chloride-dominated brine is then simulated with a reactive transport model using a brine composition representative of partially evaporated calcium-rich pore waters at Yucca Mountain. Results show a significant increase in boiling temperature from evaporative concentration, as well as low pH in condensates, particularly for dynamic systems where partial condensation takes place, which result in enrichment of HCl in condensates. These results are in qualitative agreement with experimental data from other studies. The combination of reactive transport with multicomponent brine chemistry to study evaporation, boiling, and the potential for acid gas generation at the proposed Yucca Mountain repository is seen as an improvement relative to previously applied simpler batch evaporation models. This approach allows the evaluation of thermal, hydrological, and chemical (THC) processes in a coupled manner, and modeling of settings much more relevant to actual field conditions than the distillation experiment considered. The actual and modeled distillation experiments do not represent expected conditions in an emplacement drift, but nevertheless illustrate the potential for acid-gas generation at moderate temperatures (<150 C).

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

    SciTech Connect (OSTI)

    T.C. Totemeier; D.M. Wachs; D.L. Porter

    2008-05-01

    Fatigue crack growth and impact 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 layer; the specimens incorporated this bond layer. Impact tests were performed at temperatures from –50 to 400°C; cracks propagating from the V-notch were arrested by delamination at the bond layer for all tests with one exception at –50°C. Fatigue crack growth 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. In all conditions the presence of the Ni bond layer was found to result in a net retardation of growth as the crack passed through the layer. The mechanism of retardation was identified as a disruption of crack planarity and uniformity after passing through the porous bond layer. 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-nickel interface.

  17. Evaluation of a superheater enhanced geothermal steam power plant in the Geysers area. Final report

    SciTech Connect (OSTI)

    Janes, J.

    1984-06-01

    This study was conducted to determine the attainable generation increase and to evaluate the economic merits of superheating the steam that could be used in future geothermal steam power plants in the Geyser-Calistoga Known Geothermal Resource Area (KGRA). It was determined that using a direct gas-fired superheater offers no economic advantages over the existing geothermal power plants. If the geothermal steam is heated to 900/sup 0/F by using the exhaust energy from a gas turbine of currently available performance, the net reference plant output would increase from 65 MW to 159 MW (net). Such hybrid plants are cost effective under certain conditions identified in this document. The power output from the residual Geyser area steam resource, now equivalent to 1437 MW, would be more than doubled by employing in the future gas turbine enhancement. The fossil fuel consumed in these plants would be used more efficiently than in any other fossil-fueled power plant in California. Due to an increase in evaporative losses in the cooling towers, the viability of the superheating concept is contingent on development of some of the water resources in the Geysers-Calistoga area to provide the necessary makeup water.

  18. 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

    1995-01-01

    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...

  19. 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...

  20. ADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING

    E-Print Network [OSTI]

    Mitchell, John E.

    ADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING WATER REACTOR AND THE HEAT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2. Advanced Boiling Water Reactor - General Description . . . . . . . . . . . 3 2.1 Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 ii #12;4. Advanced Boiling Water Reactor . . . . . . . . . . . . . . . . . . . . . . . 46 4

  1. Enhancement of Pool Boiling Heat Transfer in Confined Space 

    E-Print Network [OSTI]

    Hsu, Chia-Hsiang

    2014-05-05

    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...

  2. Boiling Water CanningProject Manual

    E-Print Network [OSTI]

    O'Laughlin, Jay

    Boiling Water CanningProject Manual pnW 652 a pacific northwest extension publication University preservation series contains four manuals: Freezing for ages 8­18 Drying for ages 8­18 Boiling water canning199/pnw199.pdf canning tomatoes and tomato products, pnW 300 http://extension.oregonstate.edu/catalog

  3. STEAM-SIDE OXIDE SCALE EXFOLIATION BEHAVIOR IN SUPERHEATERS AND REHEATERS

    SciTech Connect (OSTI)

    Sabau, Adrian S; Shingledecker, John P.; Wright, Ian G

    2011-01-01

    Advances in materials for power plants include not only new materials with higher-temperature capabilities, but also the use of current materials at increasingly higher temperatures. This latter activity builds on extensive experience of the performance of the various alloys, and provides a basis for identifying changes in alloy behavior with increasing temperature as well as understanding the factors that ultimately determine the maximum use temperatures of the different alloy classes. This paper presents results from an effort to model the exfoliation processes of steam-side oxide scales in a manner that describes as accurately as possible the evolution of strains in oxides growing inside small-diameter tubes subjected to large thermal gradients and to thermal transients typical of normal steam boiler operation. One way of portraying the results of such calculations is by plotting the evolving strains in a given oxide scale on an Exfoliation Diagram (of the type pioneered by Manning et al. of the British Central Electricity Research Laboratory) to determine the earliest time at which the trajectory of these strains intersects a criterion for scale failure. Understanding of how such strain trajectories differ among different alloys and are affected by the major variables associated with boiler operation has the potential to suggest boiler operating strategies to manage scale exfoliation, as well as to highlight the mode of scale failure and the limitations of each alloy. Preliminary results are presented of the strain trajectories calculated for alloys T22, T91, and TP347 subjected to the conditions experienced by superheaters under assumed boiler operating scenarios. For all three alloys the earliest predicted scale failures were associated with the increased strains developed during a boiler shut-down event; indeed, in the cases considered it appeared unlikely that scale failure would occur in any practically meaningful time due to strains accumulated during operation in a load-following mode in the absence of a shut down. The accuracy of the algorithms used for the kinetics of oxide growth appeared to be a very important consideration, especially for alloy TP347 for which large effects on oxide growth rate are known to occur with changes in alloy grain size and surface cold work.

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

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

    Cooling Boiling in Head Region - PACCAR Integrated Underhood Thermal and External Aerodynamics- Cummins Cooling Boiling in Head Region - PACCAR Integrated Underhood Thermal and...

  5. Acoustically Enhanced Boiling Heat Transfer

    E-Print Network [OSTI]

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

    2008-01-07

    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.

  6. 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.

    2010-09-01

    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.

  7. Numerical Simulations of Boiling Jet Impingement Cooling in Power Electronics

    SciTech Connect (OSTI)

    Narumanchi, S.; Troshko, A.; Hassani, V.; Bharathan, D.

    2006-12-01

    This paper explores turbulent boiling jet impingement for cooling power electronic components in hybrid electric vehicles.

  8. Pool boiling heat transfer characteristics of nanofluids

    E-Print Network [OSTI]

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

    2007-01-01

    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 ...

  9. Pool boiling on nano-finned surfaces 

    E-Print Network [OSTI]

    Sriraman, Sharan Ram

    2009-05-15

    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 ...

  10. EEE 564 Interdisciplinary Nuclear Power Operations (3 hrs) Catalog Description: Nuclear power plant systems. Study of the interrelationship and propagation of

    E-Print Network [OSTI]

    Zhang, Junshan

    EEE 564 Interdisciplinary Nuclear Power Operations (3 hrs) Catalog Description: Nuclear power plant (Generation II) pressurized water reactors (PWRs) and boiling water reactors (BWRs) as well as the new Electric's advanced boiling water reactor (ABWR) and economic simplified boiling water reactor (ESBWR

  11. Boiling and condensation in a liquid-filled enclosure

    E-Print Network [OSTI]

    Bar-Cohen Avram

    1971-01-01

    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, ...

  12. 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

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

    E-Print Network [OSTI]

    Attinger, Daniel

    transport [1,2], prevents dropwise condensation or fogging [3,4], and facilitates boiling [5]. A surface

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

    E-Print Network [OSTI]

    Shai, Isaac

    1967-01-01

    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 ...

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

    E-Print Network [OSTI]

    Iloeje, Onwuamaeze C.

    1972-01-01

    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. ...

  16. 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 Xu, Jian-Qun; Wang, Xin-Meng; Tu, Yi-You

    2014-07-28

    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.

  17. 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

  18. 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.

    , e.g. nuclear power plant steam generators. Two main boiling regimes can be dis- tinguished: nucleate causes a recoil force anal- ogous to that created by the gas emitted by a rocket engine. It pushes expands while transforming from liquid to gas phase. Obviously, the stronger the evap- oration rate

  19. Pebble Bed Boiling Water Reactor Concept With Superheated Steam

    SciTech Connect (OSTI)

    Tsiklauri, G.; Newman, D.; Meriwether, G.; Korolev, V. [Pacific Northwest National Laboratory, P.O. Box 999 Richland, WA 99352 (United States)

    2002-07-01

    An Advanced Nuclear Reactor concept is presented which extends Boiling Water Reactor technology with micro-fuel elements (MFE) and produces superheated steam. A nuclear plant with MFE is highly efficient and safe, due to ceramic-clad nuclear fuel. Water is used as both moderator and coolant. The fuel consists of spheres of about 1.5 mm diameter of UO{sub 2} with several external coatings of different carbonaceous materials. The outer coating of the particles is SiC, manufactured with chemical vapor disposition (CVD) technology. Endurance of the integrity of the SiC coating in water, air and steam has been demonstrated experimentally in Germany, Russia and Japan. This paper describes a result of a preliminary design and analysis of 3750 MWt (1500 MWe) plant with standard pressure of 16 MPa, which is widely achieved in the vessel of pressurized-water type reactors. The superheated steam outlet temperature of 550 deg. C elevates the steam cycle to high thermal efficiency of 42%. (authors)

  20. 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.

    , and Compensation in Boiling Water Reactor Pressure Control Systems Abstract-An analysisis provided of the effeds THECONTROL of steam pressure inboiling water reactor(BWR)nuclear power stations is one of the critical plantIEEE TRANSACTIONS ON AUTOMATIC CONTROL, VOL. AC-20, NO. 6, DECEMBER 1975 127 Acoustics, Stability

  1. 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.)

    2001-01-01

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1992-11-01

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1992-11-01

    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.

  4. Metallurgical failure analysis of a propane tank boiling liquid...

    Office of Scientific and Technical Information (OSTI)

    Journal Article: Metallurgical failure analysis of a propane tank boiling liquid expanding vapor explosion (BLEVE). Citation Details In-Document Search Title: Metallurgical failure...

  5. A method of correlating heat transfer data for surface boiling of liquids

    E-Print Network [OSTI]

    Rohsenow, Warren M.

    1951-01-01

    A method based an a logical uxplanation of the meani of beat transfer associated with the boiling process is presented for correlating heat transfer data for nucleate boiling of liquids for the case of pool boiling. Tbe ...

  6. Exploring the Limits of Boiling and Evaporative Heat Transfer Using Micro/Nano Structures

    E-Print Network [OSTI]

    Lu, Ming-Chang

    2010-01-01

    transfer coefficient models in pool boiling In summary, highlength effect on nucleate pool boiling heat transfer AnnalsTheory of The Peak and Minimum Pool Boiling Heat Fluxes, CR-

  7. 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)

    2000-01-01

    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.

  8. Multi-cycle boiling water reactor fuel cycle optimization

    SciTech Connect (OSTI)

    Ottinger, K.; Maldonado, G.I.

    2013-07-01

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

  9. The correlation of nucleate boiling burn-out data

    E-Print Network [OSTI]

    Griffith, P.

    1957-01-01

    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 ...

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

    E-Print Network [OSTI]

    Fiori, Mario P.

    1966-01-01

    "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 ...

  11. Burnout in forced convection nucleate boiling of water

    E-Print Network [OSTI]

    Reynolds John Mitchell

    1957-01-01

    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 ...

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

    E-Print Network [OSTI]

    Dr?mer, Thomas

    1964-01-01

    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 ...

  13. Model of critical heat flux in subcooled flow boiling

    E-Print Network [OSTI]

    Fiori, Mario P.

    1968-01-01

    The physical phenomenon occurring before and at the critical heat flux (CHF) for subcooled flow boiling has been investigated. The first phase of this study established the basic nature of the flow structure at CHF. A ...

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

    E-Print Network [OSTI]

    Sathyamurthi, Vijaykumar

    2009-05-15

    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) ...

  15. AECU-4439 PHYSICS AND MATHEMATICS HYDRODYNAMIC ASPECTS OF BOILING...

    Office of Scientific and Technical Information (OSTI)

    of Bubble Formation in Heat Transfer to Subcoaled Liquids , ( ( Heat Trm sf er and Fluid Mechanics Inet. , Berkeley, 1949, 113, of Boiling Heat Phys Z e i tsch, 36 , 1935, 267 I ,...

  16. 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.%). ...

  17. 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

    2004-04-01

    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.

  18. Thermal boundary layer development in dispersed flow film boiling

    E-Print Network [OSTI]

    Hull, Lawrence M.

    1982-01-01

    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 ...

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

    E-Print Network [OSTI]

    Truong, Bao H. (Bao Hoai)

    2007-01-01

    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 ...

  20. Effects of surface parameters on boiling heat transfer phenomena

    E-Print Network [OSTI]

    Truong, Bao H. (Bao Hoai)

    2011-01-01

    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 ...

  1. Forced-convection, dispersed-flow film boiling

    E-Print Network [OSTI]

    Hynek, Scott Josef

    1969-01-01

    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. ...

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

    E-Print Network [OSTI]

    Barber, Jacqueline Claire

    2010-01-01

    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 ...

  3. 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.

    2006-10-20

    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.

  4. Numerical Simulations of Bubble Dynamics and Heat Transfer in Pool Boiling--Including the Effects of Conjugate Conduction, Level of Gravity, and Noncondensable Gas Dissolved in the Liquid

    E-Print Network [OSTI]

    Aktinol, Eduardo

    2014-01-01

    Numerical Simulation of Pool Boiling: A Review. ” Journal ofBooth, W. (2012). “Nucleate Pool Boiling Experiments (NPBX)and Booth, W. , “ ,Nucleate Pool Boiling Experiments (NPBX)

  5. Progress in the Development of Compressible, Multiphase Flow Modeling Capability for Nuclear Reactor Flow Applications

    SciTech Connect (OSTI)

    R. A. Berry; R. Saurel; F. Petitpas; E. Daniel; O. Le Metayer; S. Gavrilyuk; N. Dovetta

    2008-10-01

    In nuclear reactor safety and optimization there are key issues that rely on in-depth understanding of basic two-phase flow phenomena with heat and mass transfer. Within the context of multiphase flows, two bubble-dynamic phenomena – boiling (heterogeneous) and flashing or cavitation (homogeneous boiling), with bubble collapse, are technologically very important to nuclear reactor systems. The main difference between boiling and flashing is that bubble growth (and collapse) in boiling is inhibited by limitations on the heat transfer at the interface, whereas bubble growth (and collapse) in flashing is limited primarily by inertial effects in the surrounding liquid. The flashing process tends to be far more explosive (and implosive), and is more violent and damaging (at least in the near term) than the bubble dynamics of boiling. However, other problematic phenomena, such as crud deposition, appear to be intimately connecting with the boiling process. In reality, these two processes share many details.

  6. Pool boiling heat transfer enhancement over cylindrical tubes with water at atmospheric pressure, Part I: Experimental results

    E-Print Network [OSTI]

    Kandlikar, Satish

    Pool boiling heat transfer enhancement over cylindrical tubes with water at atmospheric pressure online 4 May 2013 Keywords: Pool boiling Heat transfer enhancement Open microchannels Cylindrical tube boiling heat transfer over enhanced cylindrical microchannel test surfaces with water at atmospheric

  7. On Boiling of Crude Oil under Elevated Pressure

    E-Print Network [OSTI]

    Pimenova, Anastasiya V

    2015-01-01

    We construct a thermodynamic model for theoretical calculation of the boiling process of multicomponent mixtures of hydrocarbons (e.g., crude oil). The model governs kinetics of the mixture composition in the course of the distillation process along with the boiling temperature increase. The model heavily relies on the theory of dilute solutions of gases in liquids. Importantly, our results are applicable for modelling the process under elevated pressure (while the empiric models for oil cracking are not scalable to the case of extreme pressure), such as in an oil field heated by lava intrusions.

  8. On Boiling of Crude Oil under Elevated Pressure

    E-Print Network [OSTI]

    Anastasiya V. Pimenova; Denis S. Goldobin

    2015-10-08

    We construct a thermodynamic model for theoretical calculation of the boiling process of multicomponent mixtures of hydrocarbons (e.g., crude oil). The model governs kinetics of the mixture composition in the course of the distillation process along with the boiling temperature increase. The model heavily relies on the theory of dilute solutions of gases in liquids. Importantly, our results are applicable for modelling the process under elevated pressure (while the empiric models for oil cracking are not scalable to the case of extreme pressure), such as in an oil field heated by lava intrusions.

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

    E-Print Network [OSTI]

    Reza, S.M. Mohsin

    2002-01-01

    at low-pressure conditions were underestimated. The same model was used to simulate high pressure subcooled boiling data. High pressure subcooled boiling experiments of Bartolomey and Sabotinov were simulated. The axial void fraction distribution results...

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

    E-Print Network [OSTI]

    Bergles A. E.

    1962-01-01

    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. ...

  11. Measurement of Nucleate Pool Boiling with Synchronized Particle Imaging Velocimetry and Infrared Thermometry

    E-Print Network [OSTI]

    Duan, X.

    Nucleate boiling is important in many energy systems including light water reactors. Currently significant efforts are underway to develop mechanisticmodels for nucleate boiling based on computational fluid dynamics (CFD). ...

  12. 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 ...

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

    E-Print Network [OSTI]

    Deane, Charles William

    1969-01-01

    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 ...

  14. Hydrophobic coatings for film boiling based drag reduction on a torpedo model

    E-Print Network [OSTI]

    Campbell, Ian J. (Ian James Kenneth)

    2015-01-01

    Previous research has shown that porous, hydrophobic surfaces exhibit a dramatic reduction in critical heat flux (CHF), the amount of heat over a surface area required to initiate film boiling. Film boiling is characterized ...

  15. 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

  16. Noise Decomposition in Boiling Water Reactors with Application to Stability Monitoring

    E-Print Network [OSTI]

    Pázsit, Imre

    Noise Decomposition in Boiling Water Reactors with Application to Stability Monitoring J. Karlsson in boiling water reactor (BWR) noise measure- ments, based on flux factorization techniques (i.e., using reactors4 or flux oscillations in boiling water reactors5,6 ~BWRs!. In these cases the different modes have

  17. Advanced Power Plant Modeling with Applications to an Advanced Boiling Water

    E-Print Network [OSTI]

    Mitchell, John E.

    Advanced Power Plant Modeling with Applications to an Advanced Boiling Water Reactor and a Heat and an Advanced Boiling Water Reactor (ABWR). The continuity wave equa- tions for single and two-phase flow advanced method, are shown. These both are applied to a simplified model of the Advanced Boil- ing Water

  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. Experimental investigation on the flow instability behavior of a multi-channel boiling natural circulation loop at low-pressures

    SciTech Connect (OSTI)

    Jain, Vikas; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2010-09-15

    Natural circulation as a mode of heat removal is being considered as a prominent passive feature in the innovative nuclear reactor designs, particularly in boiling-water-reactors, due to its simplicity and economy. However, boiling natural circulation system poses many challenges to designer due to occurrence of various kinds of instabilities such as excursive instability, density wave oscillations, flow pattern transition instability, geysering and metastable states in parallel channels. This problem assumes greater significance particularly at low-pressures i.e. during startup, where there is great difference in the properties of two phases. In light of this, a parallel channel loop has been designed and installed that has a geometrical resemblance to the pressure-tube-type boiling-water-reactor, to investigate into the behavior of boiling natural circulation. The loop comprises of four identical parallel channels connected between two common plenums i.e. steam drum and header. The recirculation path is provided by a single downcomer connected between steam drum and header. Experiments have been conducted over a wide range of power and pressures (1-10 bar). Two distinct unstable zones are observed with respect to power i.e. corresponding to low power (Type-I) and high power (Type-II) with a stable zone at intermediate powers. The nature of oscillations in terms of their amplitude and frequency and their evolution for Type-I and Type-II instabilities are studied with respect to the effect of heater power and pressure. This paper discusses the evolution of unstable and stable behavior along with the nature of flow oscillation in the channels and the effect of pressure on it. (author)

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

    E-Print Network [OSTI]

    Ervin, Elizabeth K.

    : Boiling Water Reactor Reactor Manufacturer: General Electric Turbine Generator Manufacturer: General a nuclear power plant. Plant was Entergy, a Boiling Water Reactor (BWR) type. Built in the 80's, it has from the reactor is stored under water. An alternative storage is the dry cask storage which

  2. NUCLEAR SCIENCE AND ENGINEERING: 158, 164193 ~2008! Multifractal Analysis of Chaotic Flashing-Induced Instabilities in

    E-Print Network [OSTI]

    Demazière, Christophe

    of forced-circulation boiling water reac- served experimentally in a number of cases since then.10 tors-Induced Instabilities in Boiling Channels in the Natural-Circulation CIRCUS Facility Christophe Demazière* Chalmers of Nuclear Reactors, Delft, 2629 BJ, The Netherlands Received October 30, 2006 Accepted May 12, 2007

  3. NUCLEAR ENERGY Annals of Nuclear Energy 32 (2005) 812842

    E-Print Network [OSTI]

    Demazière, Christophe

    2005-01-01

    or density wave oscillation (DWO) in boiling water reactors (BWRs) (Analytis et al., 2001; Lansa°ker, 1997 rods in pressurized water reactors (PWRs). Such noise sources are not considered in this paper.elsevier.com/locate/anucene Identification and localization of absorbers of variable strength in nuclear reactors C. Demazie`re a,*, G

  4. Efficiency of a solar collector with internal boiling

    SciTech Connect (OSTI)

    Neeper, D.A.

    1986-01-01

    The behavior of a solar collector with a boiling fluid is analyzed to provide a simple algebraic model for future systems simulations, and to provide guidance for testing. The efficiency equation is developed in a form linear in the difference between inlet and saturation (boiling) temperatures, whereas the expression upon which ASHRAE Standard 109P is based utilizes the difference between inlet and ambient temperatures. The coefficient of the revised linear term is a weak function of collector parameters, weather, and subcooling of the working fluid. For a glazed flat-plate collector with metal absorber, the coefficient is effectively constant. Therefore, testing at multiple values of insolation and subcooling, as specified by ASHRAE 109P, should not be necessary for most collectors. The influences of collector properties and operating conditions on efficiency are examined.

  5. Modeling acid-gas generation from boiling chloride brines

    E-Print Network [OSTI]

    Zhang, Guoxiang

    2010-01-01

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

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

    SciTech Connect (OSTI)

    Not Available

    1994-07-01

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1994-07-01

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    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.

  9. Evaluation of a severe accident management strategy for boiling water reactors -- Drywell flooding

    SciTech Connect (OSTI)

    Yu, D.; Xing, L.; Kastenberg, W.E.; Okrent, D. (Univ. of California, Los Angeles, CA (United States). Mechanical, Aerospace, and Nuclear Engineering Dept.)

    1994-05-01

    Flooding of the drywell has been suggested as a strategy to prevent reactor vessel and containment failure in boiling water reactors. To evaluate the candidate strategy, this study considers accident management as a decision problem ( drywell flooding'' versus do nothing'') and develops a decision-oriented framework, namely, the influence diagram approach. This analysis chooses the long-term station blackout sequence for a Mark 1 nuclear power plant (Peach Bottom), and an influence diagram with a single decision node is constructed. The node probabilities in the influence diagram are obtained from US Nuclear Regulatory Commission reports or estimated by probabilistic risk assessment methodology. In assessing potential benefits compared with adverse effects, this analysis uses two consequence measures, i.e., early and late fatalities, as decision criteria. The analysis concludes that even though potential adverse effects exist, such as ex-vessel steam explosions and containment isolation failure, the drywell flooding strategy is preferred to do nothing'' when evaluated in terms of these consequence measures.

  10. Union job fight boiling at DOE cleanup sites

    SciTech Connect (OSTI)

    Setzer, S.W.

    1993-11-15

    The US DOE is facing a growing jurisdictional dispute over which unions will perform the majority of clean-up work at its facilities. Unions affiliated with the AFL-CIO Metal Trades Council representing operations employees at the sites believe they have a fundamental right to work. Unions in the AFL-CIO's Building and Construction Trades Dept. insist that they have a clear mandate under federal labor law and the Davis-Bacon Act. The issue has heated up in recent weeks at the policy level and is boiling in a contentious dispute at DOE's Fernald site in Ohio.

  11. COMPOSITION OF VAPORS FROM BOILING NITRIC ACID SOLUTIONS B A

    Office of Scientific and Technical Information (OSTI)

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield MunicipalTechnical Report: Achievements ofCOMPOSITION OF VAPORS FROM BOILING NITRIC ACID SOLUTIONS B A

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

    DOE Patents [OSTI]

    Gluntz, Douglas M. (San Jose, CA)

    1996-01-01

    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.

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

    DOE Patents [OSTI]

    Forsberg, Charles W. (Kingston, TN)

    1987-01-01

    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.

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

    DOE Patents [OSTI]

    Gluntz, D.M.

    1996-03-12

    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, C.W.

    1985-02-19

    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.

  16. 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.

    2004-10-06

    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.

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

    E-Print Network [OSTI]

    Chung, J. N.

    2011-01-01

    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 ...

  18. Hypothetical Thermodynamic Properties: The Boiling and Critical Temperatures of Polyethylene and Polytetrafluoroethylene

    E-Print Network [OSTI]

    Chickos, James S.

    Hypothetical Thermodynamic Properties: The Boiling and Critical Temperatures of Polyethylene value asymptotically approaches TB() ) (1217 ( 246) K for series related to polyethylene by the melting temperature of polyethylene.4,5 In this article, the question of whether the normal boiling tem

  19. PHYSICAL REVIEW E 89, 013011 (2014) Unorthodox bubbles when boiling in cold water

    E-Print Network [OSTI]

    Granick, Steve

    2014-01-01

    -infrared laser beam heating water below the boiling point (60­70 °C) with heating powers spanning the range from the bubbles are conventional: They grow symmetrically through evaporation until buoyancy lifts them away produced by boiling [4,14,18­21] but while important, such studies show only by inference the life cycles

  20. 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.

    2011-08-08

    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...

  1. 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

    1995-12-31

    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).

  2. 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

    1995-12-31

    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.

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

    SciTech Connect (OSTI)

    PetrusTakaki, N.

    2012-07-01

    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)

  4. Statistical nature of boiling flows: an experimental approach

    SciTech Connect (OSTI)

    Jain, P.K.

    1981-01-01

    Two vertical, concentric annular test sections were used. Two flow-field variables, viz. static pressure fluctuations at the test section outer wall and instantaneous chordal-average vapor fraction, were studied. Matched piezo-electric pressure transducers were used for the pressure fluctuation measurements, and a linearized dual-beam x-ray system was used for the vapor fraction measurements. Steady state (mean) thermal-hydraulic condition in the last section was determined by an analytical model and verified to a certain extent by capacitance probe vapor volume fraction measurements. A wide range of local (measurement station) vapor fraction conditions with prevalent flow regimes ranging from subcooled bubbly to saturated churn turbulent-slug-annular was investigated. It is suggested that diagnosis of boiling flow regimes on the basis of the statistical properties of wall static pressure fluctuations and vapor fraction fluctuations may be possible.

  5. Analysis of scrams and forced outages at boiling water reactors

    SciTech Connect (OSTI)

    Earle, R. T.; Sullivan, W. P.; Miller, K. R.; Schwegman, W. J.

    1980-07-01

    This report documents the results of a study of scrams and forced outages at General Electric Boiling Water Reactors (BWRs) operating in the United States. This study was conducted for Sandia Laboratories under a Light Water Reactor Safety Program which it manages for the United States Department of Energy. Operating plant data were used to identify the causes of scrams and forced outages. Causes of scrams and forced outages have been summarized as a function of operating plant and plant age and also ranked according to the number of events per year, outage time per year, and outage time per event. From this ranking, identified potential improvement opportunities were evaluated to determine the associated benefits and impact on plant availability.

  6. Improvements of fuel failure detection in boiling water reactors using helium measurements

    SciTech Connect (OSTI)

    Larsson, I.; Sihver, L.; Grundin, A.; Helmersson, J. O.

    2012-07-01

    To certify a continuous and safe operation of a boiling water reactor, careful surveillance of fuel integrity is of high importance. The detection of fuel failures can be performed by off-line gamma spectroscopy of off-gas samples and/or by on-line nuclide specific monitoring of gamma emitting noble gases. To establish the location of a leaking fuel rod, power suppression testing can be used. The accuracy of power suppression testing is dependent on the information of the delay time and the spreading of the released fission gases through the systems before reaching the sampling point. This paper presents a method to improve the accuracy of power suppression testing by determining the delay time and gas spreading profile. To estimate the delay time and examine the spreading of the gas in case of a fuel failure, helium was injected in the feed water system at Forsmark 3 nuclear power plant. The measurements were performed by using a helium detector system based on a mass spectrometer installed in the off-gas system. The helium detection system and the results of the experiment are presented in this paper. (authors)

  7. Advanced fuel assembly characterization capabilities based on gamma tomography at the Halden boiling water reactor

    SciTech Connect (OSTI)

    Holcombe, S.; Eitrheim, K.; Svaerd, S. J.; Hallstadius, L.; Willman, C.

    2012-07-01

    Characterization of individual fuel rods using gamma spectroscopy is a standard part of the Post Irradiation Examinations performed on experimental fuel at the Halden Boiling Water Reactor. However, due to handling and radiological safety concerns, these measurements are presently carried out only at the end of life of the fuel, and not earlier than several days or weeks after its removal from the reactor core. In order to enhance the fuel characterization capabilities at the Halden facilities, a gamma tomography measurement system is now being constructed, capable of characterizing fuel assemblies on a rod-by-rod basis in a more timely and efficient manner. Gamma tomography for measuring nuclear fuel is based on gamma spectroscopy measurements and tomographic reconstruction techniques. The technique, previously demonstrated on irradiated commercial fuel assemblies, is capable of determining rod-by-rod information without the need to dismantle the fuel. The new gamma tomography system will be stationed close to the Halden reactor in order to limit the need for fuel transport, and it will significantly reduce the time required to perform fuel characterization measurements. Furthermore, it will allow rod-by-rod fuel characterization to occur between irradiation cycles, thus allowing for measurement of experimental fuel repeatedly during its irradiation lifetime. The development of the gamma tomography measurement system is a joint project between the Inst. for Energy Technology - OECD Halden Reactor Project, Westinghouse (Sweden), and Uppsala Univ.. (authors)

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

    SciTech Connect (OSTI)

    1994-04-30

    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.

  9. 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

    1995-10-01

    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.

  10. Critical heat flux for free convection boiling in thin rectangular channels

    SciTech Connect (OSTI)

    Cheng, Lap Y.; Tichler, P.R.

    1991-01-01

    A review of the experimental data on free convection boiling critical heat flux (CHF) in vertical rectangular channels reveals three mechanisms of burnout. They are the pool boiling limit, the circulation limit, and the flooding limit associated with a transition in flow regime from churn to annular flow. The dominance of a particular mechanism depends on the dimensions of the channel. Analytical models were developed for each free convection boiling limit. Limited agreement with data is observed. A CHF correlation, which is valid for a wide range of gap sizes, was constructed from the CHFs calculated according to the three mechanisms of burnout. 17 refs., 7 figs.

  11. Effect of surface roughness and polymeric additive on nucleate pool boiling at subatmospheric pressures

    SciTech Connect (OSTI)

    Tewari, P.K.; Verma, R.K.; Ramani, M.P.S.; Mahajan, S.P.

    1986-09-01

    This investigation pertains to boiling heat transfer from a submerged flat surface at subatmospheric and atmospheric pressures in the presence of hydroxy ethyl cellulose (HEC) as a polymeric additive in small doses. Boiling was carried out in presence of the additive on smooth and rough aluminium surfaces having effective cavity size within the range as predicted by Hsu model and the pressure was kept in the range of 8 - 100 KN/sq.m (abs). Effects of surface roughness, saturation pressure and polymer concentration on boiling heat transfer were studied and the results were compared with Rohsenow's correlation.

  12. 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.; Wang, S.

    2014-01-29

    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.

  13. Flow boiling and two-phase flow instabilities in silicon microchannel heat sinks for microsystems cooling 

    E-Print Network [OSTI]

    Bogojevi?, Dario

    2010-01-01

    Flow boiling in microchannels, while very promising as a cooling technology in electronics thermal management, is still a subject being explored that requires further investigation. Before applying this technology for ...

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

    E-Print Network [OSTI]

    Tow, Emily W

    2012-01-01

    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 ...

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

    E-Print Network [OSTI]

    Maulbetsch, John S.

    1965-01-01

    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 ...

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

    E-Print Network [OSTI]

    Phillips, Bren Andrew

    2011-01-01

    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 ...

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

    E-Print Network [OSTI]

    Levin, Alan Edward

    1980-01-01

    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 ...

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

    E-Print Network [OSTI]

    Lenci, Giancarlo

    2013-01-01

    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 ...

  19. Enhanced flow boiling heat transfer in microchannels with structured surfaces at varied mass flow rates

    E-Print Network [OSTI]

    Bian, David (David Wei)

    2015-01-01

    This thesis investigates the role of mass flux on flow boiling heat transfer in microchannels with surface micropillar arrays. The motivation for this investigation was to determine the general trends of the optimal ...

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

    E-Print Network [OSTI]

    Westheimer, David Thomas

    2000-01-01

    poorly with an Earth based flow regime map; (ii) predicting that the maximum two-phase heat transfer coefficient would occur near the location the boiling fluid reached a saturated state, which also corresponded with the location of maximum nucleate...

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

    E-Print Network [OSTI]

    Booker, Kayje

    2012-01-01

    is brought to a boil. 3. Simmer (low power): Immediatelywater is maintained at a simmer for 45 minutes. In this60% open during the low power (simmer phase) test. The same

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

    E-Print Network [OSTI]

    Di, Yuan 1988-

    2012-12-05

    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...

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

    E-Print Network [OSTI]

    Sathyamurthi, Vijaykumar

    2011-02-22

    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 ...

  4. Investigation of the effects of surfactant concentration on the boiling curve of water

    E-Print Network [OSTI]

    Reed, Darci Janelle

    2015-01-01

    Boiling is a widely used heat transfer process in industry that allows for high heat transfer with a small temperature gradient. In this study the effects of two homologous series of surfactants (trimethylammonium bromide ...

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

    E-Print Network [OSTI]

    Estrada Perez, Carlos E.

    2010-01-16

    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...

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

    E-Print Network [OSTI]

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

    2010-01-01

    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 ...

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

    E-Print Network [OSTI]

    Forslund, Robert Paul

    1966-01-01

    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 ...

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

    E-Print Network [OSTI]

    Laverty, W. F.

    1964-01-01

    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 ...

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

    SciTech Connect (OSTI)

    Duignan, M.R.; Greene, G.A. ); Irvine, T.F., Jr. . Dept. of Mechanical Engineering)

    1992-04-01

    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.

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

    SciTech Connect (OSTI)

    Duignan, M.R.; Greene, G.A.; Irvine, T.F., Jr.

    1992-04-01

    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.

  11. Development of 1000 MWe Advanced Boiling Water Reactor

    SciTech Connect (OSTI)

    Kazuo Hisajima; Ken Uchida; Keiji Matsumoto; Koichi Kondo; Shigeki Yokoyama; Takuya Miyagawa [Toshiba Corporation (Japan)

    2006-07-01

    1000 MWe Advanced Boiling Water Reactor has only two main steam lines and six reactor internal pumps, whereas 1350 MWe ABWR has four main steam lines and ten reactor internal pumps. In order to confirm how the differences affect hydrodynamic conditions in the dome and lower plenum of the reactor pressure vessel, fluid analyses have been performed. The results indicate that there is not substantial difference between 1000 MWe ABWR and 1350 MWe ABWR. The primary containment vessel of the ABWR consists of the drywell and suppression chamber. The suppression chamber stores water to suppress pressure increase in the primary containment vessel and to be used as the source of water for the emergency core cooling system following a loss-of-coolant accident. Because the reactor pressure vessel of 1000 MWe ABWR is smaller than that of 1350 MWe ABWR, there is room to reduce the size of the primary containment vessel. It has been confirmed feasible to reduce inner diameter of the primary containment vessel from 29 m of 1350 MWe ABWR to 26.5 m. From an economic viewpoint, a shorter outage that results in higher availability of the plant is preferable. In order to achieve 20-day outage that results in 97% of availability, improvement of the systems for removal of decay heat is introduced that enables to stop all the safety-related decay heat removal systems except at the beginning of an outage. (authors)

  12. Camera Inspection Arm for Boiling Water Reactors - 13330

    SciTech Connect (OSTI)

    Martin, Scott; Rood, Marc

    2013-07-01

    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)

  13. 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)

    1985-01-01

    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.

  14. Power extraction from a hot stream in the presence of phase change

    E-Print Network [OSTI]

    Ordonez, Juan C.

    divides itself into three sections: liquid preheating, boiling and vapor superheating. Numerical results which the m . w stream boils, and ®nally, the surface As that superheats the steam. The simplicity

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

    DOE Patents [OSTI]

    Fife, A.B.; Ballas, G.J.

    1998-02-24

    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. 5 figs.

  16. 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)

    1998-01-01

    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.

  17. Design and Testing of Vacuum Breaker Check Valve for Simplified Boiling Water Reactor

    SciTech Connect (OSTI)

    Ishii, M.; Xu, Y.; Revankar, S.T.

    2002-07-01

    A new design of the vacuum breaker check valve was developed to replace the mechanical valve in a simplified boiling water reactor. Scaling and design calculations were performed to obtain the geometry of new passive hydraulic vacuum breaker check valve. In order to check the valve performance, a RELAP5 model of the simplified boiling water reactor system with the new valve was developed. The valve was implemented in an integral facility, PUMA and was tested for large break loss of coolant accident. (authors)

  18. Microjet array single-phase and flow boiling heat transfer with R134a Eric A. Browne a

    E-Print Network [OSTI]

    Peles, Yoav

    :6 6 Nud 6 128. Boiling experiments were conducted with liquid subcoolings of 10, 20, and 30 °C at jet effective than forced air convection will be required. Single-phase flow and flow boiling in microchannels have been studied [4­10] with a variety of fluids and flow schemes and have been used to cool turbine

  19. 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

    2003-06-16

    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

  20. 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

    2009-01-01

    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 ...

  1. Nucleate Boiling Model Liping C, Y. Sung, and V. Kucukboyaci

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformationJesseworkSURVEY UNIVERSE The 2014 surveyNuclear and Particlesafety Subscribe

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

    SciTech Connect (OSTI)

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

    1994-06-01

    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.

  3. Analysis of the Simplified Boiling Water Reactor using the code Ramona-4B 

    E-Print Network [OSTI]

    Cuevas Vivas, Gabriel Francisco

    1995-01-01

    The analysis of the Simplified Boiling Water Reactor (SBVVR) is carried out through the use of the reactor analysis code RAMONA-4B in a scenario of an operational transient, a turbine trip with failure of all the bypass valves. This study is divided...

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

    E-Print Network [OSTI]

    Raghupathi, Sri Laxmi Priya

    1998-01-01

    The enhancement of heat transfer in the realm of pool boiling refrigerants, using the concepts of electrohydrodynamics(EHD), has been actively researched in the past decade. This research aims at studying the effect of EHD on new refrigerants (R-123...

  5. 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 surfaces in laboratories to obtain the heat transfer coefficient data. In many process applications however, a fluid stream is employed as the heating medium. The heat transfer data generated with the electrically

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

    E-Print Network [OSTI]

    Peles, Yoav

    Boiling heat transfer in a hydrofoil-based micro pin fin heat sink Ali Kosßar, Yoav Peles-based micro pin fin heat sink was investigated. Average two-phase heat transfer coefficients were obtained intermittent and spray-annular flows. Heat transfer coefficient trends and flow morphologies were used to infer

  7. 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, but both the human health and envi- ronmental impacts associated with reliance on this fuel structure of fossil-fuel energy systems. These scenarios are analysed for various environmental and health impacts

  8. Boiling water reactor fuel behavior at burnup of 26 GWd/tonne U under reactivity-initiated accident conditions

    SciTech Connect (OSTI)

    Nakamura, Takehiko; Yoshinaga, Makio . Dept. of Reactor Safety Research); Sobajima, Makoto ); Ishijima, Kiyomi; Fujishiro, Toshio . Dept. of Reactor Safety Research)

    1994-10-01

    Irradiated boiling water reactor (BWR) fuel behavior under reactivity-initiated accident (RIA) conditions was investigated in the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Research Institute. Short test fuel rods, refabricated from a commercial 7 x 7 type BWR fuel rod at a burnup of 26 GWd/ tonne U, were pulse irradiated in the NSRR under simulated cooled startup RIA conditions of the BWRs. Thermal energy from 230 J/g fuel (55 cal/g fuel) to 410 J/g fuel (98 cal/g fuel) was promptly subjected to the test fuel rods by pulse irradiation within [approximately] 10 ms. The peak fuel enthalpies are believed to be the same as the prompt energy depositions. The test fuel rods demonstrated characteristic behavior of the irradiated fuel rods under the accident conditions, such as enhanced pellet cladding mechanical interaction (PCMI) and fission gas release. However, all the fuel rods survived the accident conditions with considerable margins. Simulations by the FRAP-T6 code and fresh fuel rod tests under the same RIA conditions highlighted the burnup effects on the accident fuel performance. The tests and the simulation suggested that the BWR fuel would possibly fail by a cladding burst due to fission gas release during the cladding temperature escalation rather than the PCMI under the cold startup RIA conditions of a severe power burst.

  9. Boiling Water Reactor Fuel Behavior Under Reactivity-Initiated-Accident Conditions at Burnup of 41 to 45 GWd/tonne U

    SciTech Connect (OSTI)

    Nakamura, Takehiko; Yoshinaga, Makio; Takahashi, Masato; Okonogi, Kazunari; Ishijima, Kiyomi

    2000-02-15

    Boiling water reactor (BWR) fuel at burnup of 41 to 45 GWd/tonne U was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated-accident conditions. Current Japanese BWR fuel, 8 x 8BJ type (Step I), from Fukushima-Daiichi Unit 3 was refabricated into short segments, and the test rods were promptly subjected to thermal energy from 293 to 607 J/g (70 to 145 cal/g) within {approx}20 ms. The fuel cladding was ductile enough to survive the prompt deformation due to pellet cladding mechanical interaction, while the plastic hoop strain reached 1.5% at the peak location. Transient fission gas release by the pulse irradiation varied from 3.1 to 8.2%, depending on the peak fuel enthalpy and the steady-state operation conditions.

  10. Simultaneous boiling and spreading of liquefied petroleum gas on water. Final report, December 12, 1978-March 31, 1981

    SciTech Connect (OSTI)

    Chang, H.R.; Reid, R.C.

    1981-04-01

    An experimental and theoretical investigation was carried out to study the boiling and spreading of liquid nitrogen, liquid methane and liquefied petroleum gas (LPG) on water in a one-dimensional configuration. Primary emphasis was placed on the LPG studies. Experimental work involved the design and construction of a spill/spread/boil apparatus which permitted the measurement of spreading and local boil-off rates. With the equations of continuity and momentum transfer, a mathematical model was developed to describe the boiling-spreading phenomena of cryogens spilled on water. The model accounted for a decrease in the density of the cryogenic liquid due to bubble formation. The boiling and spreading rates of LPG were found to be the same as those of pure propane. An LPG spill was characterized by the very rapid and violent boiling initially and highly irregular ice formation on the water surface. The measured local boil-off rates of LPG agreed reasonably well with theoretical predictions from a moving boundary heat transfer model. The spreading velocity of an LPG spill was found to be constant and determined by the size of the distributor opening. The maximum spreading distance was found to be unaffected by the spilling rate. These observations can be explained by assuming that the ice formation on the water surface controls the spreading of LPG spills. While the mathematical model did not predict the spreading front adequately, it predicted the maximum spreading distance reasonably well.

  11. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    SciTech Connect (OSTI)

    Durbin, Samuel; Lindgren, Eric R.

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below-ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on allowable heat load and the effect of simulated wind on a simplified below ground vent configuration. While incorporating the best available information, this test plan is subject to changes due to improved understanding from modeling or from as-built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

  12. 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.

    2011-01-19

    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.

  13. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01

    re- actor (PWR) and boiling-water reactor (BWR) designsin integral boiling water super heat reactors. Technical

  14. 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

    2014-01-01

    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 ...

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

    E-Print Network [OSTI]

    Phillips, Bren Andrew

    2014-01-01

    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 ...

  16. Modeling the Thermal Mechanical Behavior of a 300 K Vacuum Vessel that is Cooled by Liquid Hydrogen in Film Boiling

    E-Print Network [OSTI]

    Yang, S.Q.; Green, M.A.; Lau, W.

    2004-01-01

    VESSEL THAT IS COOLED BY LIQUID HYDROGEN IN FILM BOILING S.window that is part of a 20-liter liquid hydrogen vessel.This rupture will spill liquid hydrogen onto the walls and

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

    E-Print Network [OSTI]

    Schor, Andrei L.

    1982-01-01

    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 ...

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

    E-Print Network [OSTI]

    Zielinski, R. G.

    1981-01-01

    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, ...

  19. 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

    2009-01-01

    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 ...

  20. 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

    2009-01-01

    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 ...

  1. 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

    2011-09-02

    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.

  2. Critical Heat Flux for Downward-Facing Boiling on a Coated Hemispherical Vessel Surrounded by an Insulation Structure

    SciTech Connect (OSTI)

    J. Yang; F. B. Cheung; J. L. Rempe; K. Y. Suh; S. B. Kim

    2005-05-01

    An experimental study was performed to evaluate the effects of surface coating and an enhanced insulation structure on the downward facing boiling process and the critical heat flux on the outer surface of a hemispherical vessel. Steady-state boiling tests were conducted in the Subscale Boundary Layer Boiling (SBLB) facility using an enhanced vessel/insulation design for the cases with and without vessel coatings. Based on the boiling data, CHF correlations were obtained for both plain and coated vessels. It was found that the nucleate boiling rates and the local CHF limits for the case with micro-porous layer coating were consistently higher than those values for a plain vessel at the same angular location. The enhancement in the local CHF limits and nucleate boiling rates was mainly due to the micro-porous layer coating that increased the local liquid supply rate toward the vaporization sites on the vessel surface. For the case with thermal insulation, the local CHF limit tended to increase from the bottom center at first, then decrease toward the minimum gap location, and finally increase toward the equator. This nonmonotonic behavior, which differed significantly from the case without thermal insulation, was evidently due to the local variation of the two-phase motions in the annular channel between the test vessel and the insulation structure.

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

    DOE Patents [OSTI]

    Tokarz, R.D.

    1981-10-27

    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.

  4. Analysis of Nuclear Reconstitution, Nuclear

    E-Print Network [OSTI]

    Forbes, Douglass

    CHAPTER Analysis of Nuclear Reconstitution, Nuclear Envelope Assembly, and Nuclear Pore Assembly ....................................................................... 180 8.5 Assaying Assembly and Integrity of the Nuclear Envelope................................... 182 8.6 A Nuclear Pore Complex Assembly Assay Using pore-free Nuclear Intermediates

  5. BWR ATWS simulations for Browns Ferry Nuclear Plant Unit 1

    SciTech Connect (OSTI)

    Dallman, R.J.

    1984-01-01

    Under auspices of the US Nuclear Regulatory Commission, simulations of anticipated transients without scram (ATWS) in a boiling water reactor are being performed. A methodology has been developed to study the ATWS, and deterministic analyses have been conducted. Results are presented for one of the most probable (albeit hypothetical) sequences leading to core and containment damage. Areas presenting calculational uncertainties are identified, and requirements for their resolution are proposed.

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

    SciTech Connect (OSTI)

    Orlov, Andrey; Degueldre, Claude; Kaufmann, Wilfried

    2013-01-15

    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.

  7. Do surfaces with mixed hydrophilic and hydrophobic areas enhance pool boiling?

    E-Print Network [OSTI]

    Betz, Amy; Qiu, Huihe; Attinger, Daniel

    2010-01-01

    We demonstrate that smooth and flat surfaces combining hydrophilic and hydrophobic patterns improve pool boiling performance. Compared to a hydrophilic surface with 7^\\circ wetting angle, the measured critical heat flux and heat transfer coefficients of the enhanced surfaces are up to respectively 65 and 100% higher. Different networks combining hydrophilic and hydrophobic regions are characterized. While all tested networks enhance the heat transfer coefficient, large enhancements of critical heat flux are typically found for hydrophilic networks featuring hydrophobic islands. Hydrophilic networks indeed are shown to prevent the formation of an insulating vapor layer.

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

    E-Print Network [OSTI]

    Tu, Chau Qui

    1976-01-01

    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...

  9. Modeling Single-Phase and Boiling Liquid Jet Impingement Cooling in Power Electronics

    SciTech Connect (OSTI)

    Narumanchi, S. V. J.; Hassani, V.; Bharathan, D.

    2005-12-01

    Jet impingement has been an attractive cooling option in a number of industries over the past few decades. Over the past 15 years, jet impingement has been explored as a cooling option in microelectronics. Recently, interest has been expressed by the automotive industry in exploring jet impingement for cooling power electronics components. This technical report explores, from a modeling perspective, both single-phase and boiling jet impingement cooling in power electronics, primarily from a heat transfer viewpoint. The discussion is from the viewpoint of the cooling of IGBTs (insulated-gate bipolar transistors), which are found in hybrid automobile inverters.

  10. A study of out-of-phase power instabilities in boiling water reactors

    SciTech Connect (OSTI)

    March-Leuba, J.; Blakeman, E.D.

    1988-06-20

    This paper presents a study of the stability of subcritical neutronic modes in boiling water reactors that can result in out-of-phase power oscillations. A mechanism has been identified for this type of oscillation, and LAPUR code has been modified to account for it. Numerical results show that there is a region in the power-flow operating map where an out-or-phase stability mode is likely even if the core-wide mode is stable. 4 refs., 7 figs.

  11. 288 Int. J. Nuclear Energy Science and Technology, Vol. 7, No. 4, 2013 Copyright 2013 Inderscience Enterprises Ltd.

    E-Print Network [OSTI]

    Demazière, Christophe

    and boiling water reactor physics, reactor dynamics applied to both critical and sub-critical systems, power, Chalmers University of Technology, Sweden. His research interests include pressurised water reactor Inderscience Enterprises Ltd. Multi-physics modelling of nuclear reactors: current practices in a nutshell

  12. Int. J. Nuclear Energy Science and Technology, Vol. 4, No. 4, 2009 287 Development of computational methods and their

    E-Print Network [OSTI]

    Demazière, Christophe

    engineering. His research interests include pressurised water reactor and boiling water reactor physics, with typically 150­800 fuel assemblies constituting the core for Light Water Reactor (LWR) systems. Each@nephy.chalmers.se Abstract: A specificity of nuclear reactors is their multiphysics and multiscale character

  13. Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

    SciTech Connect (OSTI)

    J. Yang; F. B. Cheung; J. L. Rempe; K. Y. Suh; S. B. Kim

    2005-07-01

    Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods.

  14. 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)

    2013-12-04

    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}.

  15. Experimental investigation of micro-scale temperature transients in sub-cooled flow boiling on a horizontal heater

    E-Print Network [OSTI]

    Banerjee, Debjyoti

    systems and energy storage, materials processing and futuristic applications (e.g. ablation cooling on a silicon wafer during flow boiling on the silicon wafer which is heated from below. The silicon wafer") are fabricated on the surface of the silicon wafer. High speed data acquisition appa- ratus is used to record

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

    E-Print Network [OSTI]

    Torgersen, Christian

    Boiling Water at Hot Creek--The Dangerous and Dynamic Thermal Springs in California's Long Valley. Because of this danger, the U.S. Forest Service has had to close parts of the Hot Creek Geologic Site the region. The attractions of Hot Creek, however, also harbor danger. The locations, dis- charge rates

  17. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    SciTech Connect (OSTI)

    Schlereth, J.R.; Pennington, D.

    1996-12-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

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

    SciTech Connect (OSTI)

    PACE, M.E.

    2004-01-13

    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.

  19. Effect of nonuniformity of subcooled boiling flow on the onset of thermoacoustic vibrations

    SciTech Connect (OSTI)

    Gerliga, V.A.; Skalozubov, V.I.; Lesin, V.Y. )

    1991-01-01

    This paper develops the hypothesis that the factor responsible for the onset of thermoacoustic vibrations in two-phase bubble flow is positive work by bubbles condensing in the flow core. It is shown that the predicted threshold of generation of these vibrations depends strongly on the accuracy of description of the steady-state distribution of parameters of bubbles and the liquid. The results predicted on the basis of a two-zone nonequilibrium polydisperse model are compared with those given by the uniform-flow model and an equation representing the condition of applicability of one-dimensional models for predicting the steady-state parameters of nonequilibrium boiling flows is derived.

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

    DOE Patents [OSTI]

    Hill, Paul R. (Tucson, AZ)

    1994-01-01

    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.

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

    DOE Patents [OSTI]

    Hill, P.R.

    1994-12-27

    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.

  2. Policy Paper 37: Energy and Security in Northeast Asia: Proposals for Nuclear Cooperation

    E-Print Network [OSTI]

    Kaneko, Kumao; Suzuki, Atsuyuki; Choi, Jor-Shan; Fei, Edward

    1998-01-01

    power reactors (22 boiling water reactors (BWRs), 22 PWRs,two advanced boiling water reactors (ABWRs) at Kashiwazaki-

  3. Lattice Boltzmann Methods to Address Fundamental Boiling and Two-Phase Problems

    SciTech Connect (OSTI)

    Uddin, Rizwan

    2012-01-01

    This report presents the progress made during the fourth (no cost extension) year of this three-year grant aimed at the development of a consistent Lattice Boltzmann formulation for boiling and two-phase flows. During the first year, a consistent LBM formulation for the simulation of a two-phase water-steam system was developed. Results of initial model validation in a range of thermo-dynamic conditions typical for Boiling Water Reactors (BWRs) were shown. Progress was made on several fronts during the second year. Most important of these included the simulation of the coalescence of two bubbles including the surface tension effects. Work during the third year focused on the development of a new lattice Boltzmann model, called the artificial interface lattice Boltzmann model (AILB model) for the 3 simulation of two-phase dynamics. The model is based on the principle of free energy minimization and invokes the Gibbs-Duhem equation in the formulation of non-ideal forcing function. This was reported in detail in the last progress report. Part of the efforts during the last (no-cost extension) year were focused on developing a parallel capability for the 2D as well as for the 3D codes developed in this project. This will be reported in the final report. Here we report the work carried out on testing the AILB model for conditions including the thermal effects. A simplified thermal LB model, based on the thermal energy distribution approach, was developed. The simplifications are made after neglecting the viscous heat dissipation and the work done by pressure in the original thermal energy distribution model. Details of the model are presented here, followed by a discussion of the boundary conditions, and then results for some two-phase thermal problems.

  4. Nuclear Energy Governance and the Politics of Social Justice: Technology, Public Goods, and Redistribution in Russia and France

    E-Print Network [OSTI]

    Grigoriadis, Theocharis N

    2009-01-01

    Reactors (PWRs) CANDU Boiling Water Reactors (BWRs) AdvancedAGRs) Advanced Boiling Water Reactors (ABWRs) AP600 SystemEconomic Simplified Boiling Water Reactors (ESBWRs) APR-1400

  5. Nuclear Resonance Fluorescence for Nuclear Materials Assay

    E-Print Network [OSTI]

    Quiter, Brian Joseph

    2010-01-01

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

  6. United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    SciTech Connect (OSTI)

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

  7. 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

    2012-01-01

    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 ...

  8. 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

    2009-01-01

    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 ...

  9. 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

    1983-01-01

    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, ...

  10. Many-Group Cross-Section Adjustment Techniques for Boiling Water Reactor Adaptive Simulation

    SciTech Connect (OSTI)

    Jessee, Matthew Anderson

    2011-01-01

    Computational capability has been developed to adjust multigroup neutron cross sections, including self-shielding correction factors, to improve the fidelity of boiling water reactor (BWR) core modeling and simulation. The method involves propagating multigroup neutron cross-section uncertainties through various BWR computational models to evaluate uncertainties in key core attributes such as core k{sub eff}, nodal power distributions, thermal margins, and in-core detector readings. Uncertainty-based inverse theory methods are then employed to adjust multigroup cross sections to minimize the disagreement between BWR core modeling predictions and observed (i.e., measured) plant data. For this paper, observed plant data are virtually simulated in the form of perturbed three-dimensional nodal power distributions with the perturbations sized to represent actual discrepancies between predictions and real plant data. The major focus of this work is to efficiently propagate multigroup neutron cross-section uncertainty through BWR lattice physics and core simulator calculations. The data adjustment equations are developed using a subspace approach that exploits the ill-conditioning of the multigroup cross-section covariance matrix to minimize computation and storage burden. Tikhonov regularization is also employed to improve the conditioning of the data adjustment equations. Expressions are also provided for posterior covariance matrices of both the multigroup cross-section and core attributes uncertainties.

  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

    2005-01-01

    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. Heat removal characteristics of volume heated boiling pools with inclined boundaries

    SciTech Connect (OSTI)

    Greene, G.A.; Jones, O.C. Jr.; Schwarz, C.E.; Abuaf, N.

    1980-04-01

    The state-of-the-art of heat transfer from boiling liquids having internal heat generation is reviewed. Considerable scatter is found in the existing data. Attempts to correlate these data have relied on both natural and forced convection concepts. This report describes a new series of experiments wherein the data scatter appears to have been improved by a factor of four to six from previous experiments when compared on the basis of standard deviation in correlation coefficients. Local heat transfer data to both vertical and inclined surfaces (up to 30/sup 0/ from vertical) are reported having maximum to minimum heat transfer ratios of up to 5:1. It is shown that with surface vapor fluxes up to twice the free bubble rise velocities given by Harmathy there are two distinct flow regimes: bubbly and churn-turbulent. In bubble flows, the pool is generally quiescent and surface temperature fluctuations negligible. In churn-turbulent flows, the pool is generally chaotic and three dimensional. The surface temperatures showed large fluctuations up to the maximum pool-to-wall difference indicating intermittent destruction and renewal of boundary layer. Heat transfer coefficients were more uniform, and the maximum was observed to be in the range .25-.30 cal/cm/sup 2/ s /sup 0/C. 26 refs., 18 figs., 9 tabs.

  13. Nuclear Matter and Nuclear Dynamics

    E-Print Network [OSTI]

    M Colonna

    2009-02-26

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

  14. Fusion Nuclear Science | ORNL

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

    Fusion Nuclear Science Isotope Development and Production Nuclear Security Science & Technology Nuclear Systems Modeling, Simulation & Validation Nuclear Systems Technology...

  15. 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

    2013-09-01

    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.

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

    SciTech Connect (OSTI)

    M. Ishii; X. Sunn; S. Kuran

    2004-09-27

    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.

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

    E-Print Network [OSTI]

    Tetreault-Friend, Melanie

    2014-01-01

    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 ...

  18. Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles

    SciTech Connect (OSTI)

    Yin, Shengjun; Bass, Bennett Richard; Stevens, Gary; Kirk, Mark

    2011-01-01

    This paper describes stress analysis and fracture mechanics work performed to assess boiling water reactor (BWR) and pressurized water reactor (PWR) nozzles located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Various RPV nozzle geometries were investigated: 1. BWR recirculation outlet nozzle; 2. BWR core spray nozzle3 3. PWR inlet nozzle; ; 4. PWR outlet nozzle; and 5. BWR partial penetration instrument nozzle. The above nozzle designs were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-license (EOL) to require evaluation as part of establishing the allowed limits on heatup, cooldown, and hydrotest (leak test) conditions. These nozzles analyzed represent one each of the nozzle types potentially requiring evaluation. The purpose of the analyses performed on these nozzle designs was as follows: To model and understand differences in pressure and thermal stress results using a two-dimensional (2-D) axi-symmetric finite element model (FEM) versus a three-dimensional (3-D) FEM for all nozzle types. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated; To verify the accuracy of a selected linear elastic fracture mechanics (LEFM) hand solution for stress intensity factor for a postulated nozzle corner crack for both thermal and pressure loading for all nozzle types; To assess the significance of attached piping loads on the stresses in the nozzle corner region; and To assess the significance of applying pressure on the crack face with respect to the stress intensity factor for a postulated nozzle corner crack.

  19. Effect of rolling motion on critical heat flux for subcooled flow boiling in vertical tube

    SciTech Connect (OSTI)

    Hwang, J. S.; Park, I. U.; Park, M. Y.; Park, G. C.

    2012-07-01

    This paper presents defining characteristics of the critical heat flux (CHF) for the boiling of R-134a in vertical tube operation under rolling motion in marine reactor. It is important to predict CHF of marine reactor having the rolling motion in order to increase the safety of the reactor. Marine Reactor Moving Simulator (MARMS) tests are conducted to measure the critical heat flux using R-134a flowing upward in a uniformly heated vertical tube under rolling motion. MARMS was rotated by motor and mechanical power transmission gear. The CHF tests were performed in a 9.5 mm I.D. test section with heated length of 1 m. Mass fluxes range from 285 to 1300 kg m{sup -2}s{sup -1}, inlet subcooling from 3 to 38 deg. C and outlet pressures from 13 to 24 bar. Amplitudes of rolling range from 15 to 40 degrees and periods from 6 to 12 sec. To convert the test conditions of CHF test using R-134a in water, Katto's fluid-to-fluid modeling was used in present investigation. A CHF correlation is presented which accounts for the effects of pressure, mass flux, inlet subcooling and rolling angle over all conditions tested. Unlike existing transient CHF experiments, CHF ratio of certain mass flux and pressure are different in rolling motion. For the mass fluxes below 500 kg m{sup -2}s{sup -1} at 13, 16 (region of relative low mass flux), CHF ratio was decreased but was increased above that mass flux (region of relative high mass flux). Moreover, CHF tend to enhance in entire mass flux at 24 bar. (authors)

  20. METHODOLOGIES FOR REVIEW OF THE HEALTH AND SAFETY ASPECTS OF PROPOSED NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL SITES AND FACILITIES. VOLUME 9 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    pressurized-water or boiling- water reactor) geothermal (Vessel Code Boiling water reactor California Administrativepressurized-water and boiling-water reactor power plants.

  1. Nuclear Resonance Fluorescence for Nuclear Materials Assay

    E-Print Network [OSTI]

    Quiter, Brian Joseph

    2010-01-01

    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. 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.

    1983-03-01

    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.

  3. Nuclear Navy

    SciTech Connect (OSTI)

    Not Available

    1994-01-01

    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.

  4. Nuclear Navy

    SciTech Connect (OSTI)

    NONE

    1994-12-31

    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.

  5. Nuclear Power 

    E-Print Network [OSTI]

    2010-01-01

    be inherently safe and environmentally benign. These realities of today's world are among the reasons that lead to serious interest in deploying nuclear power as a sustainable energy source. Today's nuclear reactors are safe and highly efficient energy systems...

  6. 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...

  7. NUCLEAR ENERGY Annals of Nuclear Energy 32 (2005) 13051322

    E-Print Network [OSTI]

    Demazière, Christophe

    2005-01-01

    to quantify the stability properties of boiling water reactors (BWRs). The study was prompted-average (model) BWR boiling water reactor CEA Commissariat a` lÕEnergie Atomique DR decay ratio DWO density wave`re *, Imre Pa´zsit Department of Reactor Physics, Chalmers University of Technology, SE-412 96 Go

  8. Application of the Isotope Ratio Method to a Boiling Water Reactor

    SciTech Connect (OSTI)

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

    2010-08-11

    The isotope ratio method is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods. All reactor materials contain trace elemental impurities at parts per million levels, and the isotopes of these elements are transmuted by neutron irradiation in a predictable manner. While measuring the change in a particular isotope’s concentration is possible, it is difficult to correlate to energy production because the initial concentration of that element may not be accurately known. However, if the ratio of two isotopes of the same element can be measured, the energy production can then be determined without knowing the absolute concentration of that impurity since the initial natural ratio is known. This is the fundamental principle underlying the isotope ratio method. Extremely sensitive mass-spectrometric methods are currently available that allow accurate measurements of the impurity isotope ratios in samples. Additionally, “indicator” elements with stable activation products have been identified so that their post-irradiation isotope ratios remain constant. This method has been successfully demonstrated on graphite-moderated reactors. Graphite reactors are particularly well-suited to such analyses since the graphite moderator is resident in the fueled region of the core for the entire period of operation. Applying this method to other reactor types is more difficult since the resident portions of the reactor available for sampling are either outside the fueled region of the core or structural components of individual fuel assemblies. The goal of this research is to show that the isotope ratio method can produce meaningful results for light water-moderated power reactors. In this work, we use the isotope ratio method to estimate the energy production in a boiling water reactor fuel bundle based on measurements taken from the corresponding fuel assembly channel. Our preliminary results are in good agreement with the actual operating history of the reactor during the cycle that the fuel bundle was resident in the core.

  9. Comparing Simulation Results with Traditional PRA Model on a Boiling Water Reactor Station Blackout Case Study

    SciTech Connect (OSTI)

    Zhegang Ma; Diego Mandelli; Curtis Smith

    2011-07-01

    A previous study used RELAP and RAVEN to conduct a boiling water reactor station black-out (SBO) case study in a simulation based environment to show the capabilities of the risk-informed safety margin characterization methodology. This report compares the RELAP/RAVEN simulation results with traditional PRA model results. The RELAP/RAVEN simulation run results were reviewed for their input parameters and output results. The input parameters for each simulation run include various timing information such as diesel generator or offsite power recovery time, Safety Relief Valve stuck open time, High Pressure Core Injection or Reactor Core Isolation Cooling fail to run time, extended core cooling operation time, depressurization delay time, and firewater injection time. The output results include the maximum fuel clad temperature, the outcome, and the simulation end time. A traditional SBO PRA model in this report contains four event trees that are linked together with the transferring feature in SAPHIRE software. Unlike the usual Level 1 PRA quantification process in which only core damage sequences are quantified, this report quantifies all SBO sequences, whether they are core damage sequences or success (i.e., non core damage) sequences, in order to provide a full comparison with the simulation results. Three different approaches were used to solve event tree top events and quantify the SBO sequences: “W” process flag, default process flag without proper adjustment, and default process flag with adjustment to account for the success branch probabilities. Without post-processing, the first two approaches yield incorrect results with a total conditional probability greater than 1.0. The last approach accounts for the success branch probabilities and provides correct conditional sequence probabilities that are to be used for comparison. To better compare the results from the PRA model and the simulation runs, a simplified SBO event tree was developed with only four top events and eighteen SBO sequences (versus fifty-four SBO sequences in the original SBO model). The estimated SBO sequence conditional probabilities from the original SBO model were integrated to the corresponding sequences in the simplified SBO event tree. These results were then compared with the simulation run results.

  10. RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    the BEIR Committee - boiling water reactor - Code of FederalACCIDENT FOR BOILING WATER REACTORS (Revision 2, 6/74) 1.4ACCIDENT FOR BOILING WATER REACTORS (3/71) 1.24 ASSUMPTIONS

  11. PROBING DENSE NUCLEAR MATTER VIA NUCLEAR COLLISIONS

    E-Print Network [OSTI]

    Stocker, H.

    2012-01-01

    shocked nuclear matter during the compression and expansionand isentropic expansion were valid in nuclear collisions.

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

    E-Print Network [OSTI]

    Demazière, Christophe

    parameters that are essential for reactor safety, like reactivity coefficients and control rod worths reactors Klara Insulander Björk a,b,*, Valentin Fhager a , Christophe Demazière b a Thor Energy, Sommerrogaten 13e15, NO-0255 Oslo, Norway b Chalmers University of Technology, Department of Nuclear Engineering

  13. 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.

    1982-03-01

    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.

  14. 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

  15. 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)

    1995-01-01

    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.

  16. 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)

    1996-01-01

    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.

  17. 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 ...

  18. Nuclear Counterterrorism

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

    2013-08-26

    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. Supersedes DOE O 457.1 and DOE M 457.1-1.

  19. Nuclear fuel elements made from nanophase materials

    DOE Patents [OSTI]

    Heubeck, Norman B. (Schenectady, NY)

    1998-01-01

    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.

  20. Nuclear fuel elements made from nanophase materials

    DOE Patents [OSTI]

    Heubeck, N.B.

    1998-09-08

    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.

  1. Nuclear shadowing

    E-Print Network [OSTI]

    N. Armesto

    2006-07-05

    The phenomenon of shadowing of nuclear structure functions at small values of Bjorken-$x$ is analyzed. First, multiple scattering is discussed as the underlying physical mechanism. In this context three different but related approaches are presented: Glauber-like rescatterings, Gribov inelastic shadowing and ideas based on high-density Quantum Chromodynamics. Next, different parametrizations of nuclear partonic distributions based on fit analysis to existing data combined with Dokshitzer-Gribov-Lipatov-Altarelli-Parisi evolution, are reviewed. Finally, a comparison of the different approaches is shown, and a few phenomenological consequences of nuclear shadowing in high-energy nuclear collisions are presented.

  2. Alcohol-free alkoxide process for containing nuclear waste

    DOE Patents [OSTI]

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

    1984-01-01

    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.

  3. Modeling and Thermal Performance Evaluation of Porous Curd Layers in Sub-Cooled Boiling Region of PWRs and Effects of Sub-Cooled Nucleate Boiling on Anomalous Porous Crud Deposition on Fuel Pin Surfaces

    SciTech Connect (OSTI)

    Barclay Jones

    2005-06-27

    A significant number of current PWRs around the world are experiencing anomalous crud deposition in the sub-cooled region of the core, resulting in an axial power shift or Axial Offset Anomaly (AOA), a condition that continues to elude prediction of occurrence and thermal/neutronic performance. This creates an operational difficulty of not being able to accurately determine power safety margin. In some cases this condition has required power ''down rating'' by as much as thirty percent and the concomitant considerable loss of revenue for the utility. This study examines two aspects of the issue: thermal performance of crud layer and effect of sub-cooled nucleate boiling on the solute concentration and its influence on initiation of crud deposition/formation on fuel pin surface.

  4. PROBING DENSE NUCLEAR MATTER VIA NUCLEAR COLLISIONS

    E-Print Network [OSTI]

    Stocker, H.

    2012-01-01

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

  5. Nuclear Systems Modeling, Simulation & Validation | Nuclear Science...

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

    Research Areas Fuel Cycle Science & Technology Fusion Nuclear Science Isotope Development and Production Nuclear Security Science & Technology Nuclear Systems Modeling, Simulation...

  6. Nuclear Resonance Fluorescence for Nuclear Materials Assay

    E-Print Network [OSTI]

    Quiter, Brian Joseph

    2010-01-01

    130] International Nuclear Safety Center, Available onlinefrom Inter- national Nuclear Safety Center (INSC) website(from International Nuclear Safety Center (INSC) website(

  7. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would likeUniverseIMPACT EVALUATION PLANIs gravity aOverview ARM ClimateEnergyInterim-

  8. 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

    , College Station, TX 77843, United States a r t i c l e i n f o Article history: Received 2 August 2008) Chemical vapor deposition Multi phase flows a b s t r a c t In this study we measure the flow boiling heat). In addition, reducing the hydraulic diameter (Dh) can enhance the heat transfer coefficient (h) in single

  9. 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)

    1993-09-15

    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.

  10. Influence of lubricant oil on heat transfer performance of refrigerant flow boiling inside small diameter tubes. Part II: Correlations

    SciTech Connect (OSTI)

    Wei, Wenjian; Ding, Guoliang; Hu, Haitao; Wang, Kaijian

    2007-10-15

    The predictive ability of the available state-of-the-art heat transfer correlations of refrigerant-oil mixture is evaluated with the present experiment data of small tubes with inside diameter of 6.34 mm and 2.50 mm. Most of these correlations can be used to predict the heat transfer coefficient of 6.34 mm tube, but none of them can predict heat transfer coefficient of 2.50 mm tube satisfactorily. A new correlation of two-phase heat transfer multiplier with local properties of refrigerant-oil mixture is developed. This correlation approaches the actual physical mechanism of flow boiling heat transfer of refrigerant-oil mixture and can reflect the actual co-existing conditions of refrigerant and lubricant oil. More than 90% of the experiment data of both test tubes have less than {+-}20% deviation from the prediction values of the new correlations. (author)

  11. 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

    2008-07-08

    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.

  12. RADIOCHEMISTRY, AND NUCLEAR CHEMISTRY; NUCLEAR MEDICINE; HISTORICAL

    Office of Scientific and Technical Information (OSTI)

    The early days Richards, P. 38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEAR CHEMISTRY; NUCLEAR MEDICINE; HISTORICAL ASPECTS; TECHNETIUM 99; COLLOIDS; MOLYBDENUM...

  13. 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...

  14. Climate Change, Nuclear Power and Nuclear

    E-Print Network [OSTI]

    Climate Change, Nuclear Power and Nuclear Proliferation: Magnitude Matters Rob Goldston MIT IAP biomass wind hydro coal CCS coal nat gas CCS nat gas nuclear Gen IV nuclear Gen III nuclear Gen II 5-1 Electricity Generation: CCS and Nuclear Power Technology Options Available Global Electricity Generation WRE

  15. Nuclear Counterterrorism

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

    2006-02-07

    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

  16. Nuclear Science and Engineering

    E-Print Network [OSTI]

    Bahler, Dennis R.

    Nuclear Science and Engineering Education Sourcebook 2014 American Nuclear Society US Department of Energy #12;Nuclear Science & Engineering Education Sourcebook 2014 North American Edition American Nuclear Society Education, Training, and Workforce Division US Department of Energy Office of Nuclear

  17. nuclear navy

    National Nuclear Security Administration (NNSA)

    7%2A en Powering the Nuclear Navy http:www.nnsa.energy.govourmissionpoweringnavy

    Page...

  18. Nuclear Celebrations

    E-Print Network [OSTI]

    Hacker, Randi; Tsutsui, William

    2006-11-01

    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...

  19. Nuclear Nonproliferation

    SciTech Connect (OSTI)

    Atkins-Duffin, C E

    2008-12-10

    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.

  20. Nuclear Data

    SciTech Connect (OSTI)

    White, Morgan C.

    2014-01-23

    PowerPoint presentation targeted for educational use. 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.

  1. Supporting Organizations | Nuclear Science | ORNL

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

    Nuclear Science Engineering Fusion & Materials for Nuclear Systems Nuclear Science Home | Science & Discovery | Nuclear Science | Supporting Organizations SHARE Supporting...

  2. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTechtail.Theory ofDid you notHeatMaRIEdioxideUser Work FeaturedNuclearNP Home NuclearNuclear

  3. RADIOLOGICAL EMERGENCY RESPONSE PLANNING FOR NUCLEAR POWER PLANTS IN CALIFORNIA. VOLUME 4 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Yen, W.W.S.

    2010-01-01

    of Coolant Accident for Boiling Water Reactors,n and 1. 6,.Cool- ant for Boiling Water Reactors." U. S. Atomic EnergyCoolant Accident for Boiling Water Reactors," and Regulatory

  4. Supercritical Water Nuclear Steam Supply System: Innovations In Materials, Neutronics & Thermal-Hydraulics

    SciTech Connect (OSTI)

    Mark Anderson; M.L. Corradini; K. Sridharan; P. WIlson; D. Cho; T.K. Kim; S. Lomperski

    2004-09-02

    In the 1990's supercritical light-water reactors were considered in conceptual designs. A nuclear reactor cooled by supercritical waster would have a much higher thermal efficiency with a once-through direct power cycle, and could be based on standardized water reactor components (light water or heavy water). The theoretical efficiency could be improved by more than 33% over that of other water reactors and could be simplified with higher reliability; e.g., a boiling water reactor without steam separators or dryers.

  5. Advanced Modeling and Evaluation of the Response of Base-Isolated Nuclear Facility Structures to Vertical Earthquake Excitation

    E-Print Network [OSTI]

    Keldrauk, Eric Scott

    2012-01-01

    an advanced boiling water reactor. The ground motionsPWRs) and 35 boiling water reactors (BWRs), that operate atA - houses five boiling water reactors (BWRs) and the worlds

  6. Steam Line Break and Station Blackout Transients for Proliferation-Resistant Hexagonal Tight Lattice Boiling Water Reactor

    SciTech Connect (OSTI)

    Rohatgi, Upendra S. [Brookhaven National Laboratory (United States); Jo, Jae H. [Brookhaven National Laboratory (United States); Chung, Bub Dong [Brookhaven National Laboratory (United States); Takahashi, Hiroshi [Brookhaven National Laboratory (United States); Downar, Thomas J. [Purdue University (United States)

    2004-01-15

    Safety analyses of a proliferation-resistant, economically competitive, high-conversion boiling water reactor (HCBWR) fueled with fissile plutonium and fertile thorium oxide fuel elements, and with passive safety systems, are presented here. The HCBWR developed here is characterized by a very tight lattice with a relatively small water volume fraction in the core that therefore operates with a fast reactor neutron spectrum and a considerably improved neutron economy compared to the current generation of light water reactors. The tight lattice core has a very narrow flow channel with a hydraulic diameter less than half of the regular boiling water reactor (BWR) core and, thus, presents a special challenge to core cooling because of reduced water inventory and high friction in the core. The primary safety concern when reducing the moderator-to-fuel ratio and when using a tightly packed lattice arrangement is to maintain adequate cooling of the core during both normal operation and accident scenarios.In the preliminary HCBWR design, the core is placed in a vessel with a large chimney section, and the vessel is connected to the isolation condenser system (ICS). The vessel is placed in containment with the gravity driven cooling system (GDCS) and passive containment cooling system (PCCS) in a configuration similar to General Electric's simplified BWR (SBWR). The safety systems are similar to those of the SBWR; the ICS and PCCS are scaled with power. An internal recirculation pump is placed in the downcomer to augment the buoyancy head provided by the chimney since the buoyancy provided by the chimney alone could not generate sufficient recirculation in the vessel as the tight lattice configuration results in much larger friction in the core than with the SBWR.The constitutive relationships for RELAP5 are assessed for narrow channels, and as a result the heat transfer package is modified. The modified RELAP5 is used to simulate and analyze two of the most limiting events for a tight pitch lattice core: the station blackout and the main-steam-line-break events. The results of the analyses indicate that the HCBWR system will be safely brought to the shutdown condition for these transients.

  7. Explosive boiling of Ge{sub 35}Sb{sub 10}S{sub 55} glass induced by a CW laser

    SciTech Connect (OSTI)

    Knotek, P.; Tichy, L.

    2013-09-01

    Graphical abstract: - Highlights: • Interaction of the CW 785 nm laser with chalcogenide GeSbS glass. • First demonstration of the explosive boiling induced by CW laser in glass. • Different processes as photo-induced oxidation, expansion, and viscosity-flow observed. • Applied diagnostics SEM, DHM, AFM, force spectroscopy, and micro-Raman spectroscopy. • Damage threshold determined at 1.2 × 10{sup 24}s{sup ?1} cm{sup ?3} of absorbed photons. - Abstract: The response of bulk Ge{sub 35}Sb{sub 10}S{sub 55} glass to illumination by a continuous wave (CW) laser, sub-band-gap photons, was studied specifically with an atomic force microscopy including a force spectroscopy, with a digital holographic microscopy and with a scanning electron microscopy. Depending on the number of photons absorbed, photo-expansion, photo-oxidation and explosive boiling were observed.

  8. NUCLEAR PROXIMITY FORCES

    E-Print Network [OSTI]

    Randrup, J.

    2011-01-01

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

  9. Nuclear Waste: Forever Contaminated?

    E-Print Network [OSTI]

    Wang, Andrew

    2015-01-01

    Impacts of the Fukushima nuclear power plants on marineAccident at the Chernobyl Nuclear Power Plant. Epidemiologicand projected nuclear power. Environ. Sci. Technol. , 47,

  10. Nuclear Science | ORNL

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

    and Reports News and Awards Supporting Organizations Home | Science & Discovery | Nuclear Science Nuclear Science | Nuclear Science SHARE In World War II's Manhattan Project,...

  11. Nuclear Waste: Forever Contaminated?

    E-Print Network [OSTI]

    Wang, Andrew

    2015-01-01

    Impacts of the Fukushima nuclear power plants on marineBeyond Fukushima: Disasters, nuclear energy, and energy law.Nuclear Energy, and Energy Law (December 20, 2011). Brigham

  12. Nuclear Science & Technology

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

    Nuclear Science & Technology Nuclear Science & Technology1354608000000Nuclear Science & TechnologySome of these resources are LANL-only and will require Remote Access. No...

  13. Nuclear Reactions

    E-Print Network [OSTI]

    C. A. Bertulani

    2010-07-14

    Nuclear reactions generate energy in nuclear reactors, in stars, and are responsible for the existence of all elements heavier than hydrogen in the universe. Nuclear reactions denote reactions between nuclei, and between nuclei and other fundamental particles, such as electrons and photons. A short description of the conservation laws and the definition of basic physical quantities is presented, followed by a more detailed account of specific cases: (a) formation and decay of compound nuclei; (b)direct reactions; (c) photon and electron scattering; (d) heavy ion collisions; (e) formation of a quark-gluon plasma; (f) thermonuclear reactions; (g) and reactions with radioactive beams. Whenever necessary, basic equations are introduced to help understand general properties of these reactions. Published in Wiley Encyclopedia of Physics, ISBN-13: 978-3-527-40691-3 - Wiley-VCH, Berlin, 2009.

  14. Nuclear Models

    SciTech Connect (OSTI)

    Fossion, Ruben [Instituto de Ciencias Nucleares, Universidad Nacional Autonoma de Mexico, Apartado Postal 70-543, Mexico D. F., C.P. 04510 (Mexico)

    2010-09-10

    The atomic nucleus is a typical example of a many-body problem. On the one hand, the number of nucleons (protons and neutrons) that constitute the nucleus is too large to allow for exact calculations. On the other hand, the number of constituent particles is too small for the individual nuclear excitation states to be explained by statistical methods. Another problem, particular for the atomic nucleus, is that the nucleon-nucleon (n-n) interaction is not one of the fundamental forces of Nature, and is hard to put in a single closed equation. The nucleon-nucleon interaction also behaves differently between two free nucleons (bare interaction) and between two nucleons in the nuclear medium (dressed interaction).Because of the above reasons, specific nuclear many-body models have been devised of which each one sheds light on some selected aspects of nuclear structure. Only combining the viewpoints of different models, a global insight of the atomic nucleus can be gained. In this chapter, we revise the the Nuclear Shell Model as an example of the microscopic approach, and the Collective Model as an example of the geometric approach. Finally, we study the statistical properties of nuclear spectra, basing on symmetry principles, to find out whether there is quantum chaos in the atomic nucleus. All three major approaches have been rewarded with the Nobel Prize of Physics. In the text, we will stress how each approach introduces its own series of approximations to reduce the prohibitingly large number of degrees of freedom of the full many-body problem to a smaller manageable number of effective degrees of freedom.

  15. 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.

    1980-06-01

    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.

  16. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTechtail.Theory ofDid you notHeatMaRIEdioxideUser Work FeaturedNuclear & ParticleNuclear

  17. Nuclear Science

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTechtail.Theory ofDid you notHeatMaRIEdioxideUser Work FeaturedNuclearNP Home Nuclear

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

    SciTech Connect (OSTI)

    Dykin, V.; Pazsit, I.

    2012-07-01

    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)

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

    SciTech Connect (OSTI)

    Trianti, Nuri E-mail: szaki@fi.itba.c.id; Su'ud, Zaki E-mail: szaki@fi.itba.c.id; Arif, Idam E-mail: szaki@fi.itba.c.id; Riyana, EkaSapta

    2014-09-30

    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.

  20. Influence of lubricant oil on heat transfer performance of refrigerant flow boiling inside small diameter tubes. Part I: Experimental study

    SciTech Connect (OSTI)

    Wei, Wenjian; Ding, Guoliang; Hu, Haitao; Wang, Kaijian

    2007-10-15

    Two-phase flow pattern and heat transfer characteristics of refrigerant-oil mixture flow boiling inside small tubes with inside diameters of 6.34 mm and 2.50 mm are investigated experimentally. The test condition of nominal oil concentration is from 0% to 5%, mass flux from 200 to 400 kg m{sup -2} s{sup -1}, heat flux from 3.2 to 14 kW m{sup -2}, evaporation temperature of 5 C, inlet quality from 0.1 to 0.8, and quality change from 0.1 to 0.2. Wavy, wavy-annular, annular and mist-annular flow pattern in 6.34 mm tube are observed, while only slug-annular and annular flow pattern are observed in 2.50 mm tube. Oil presence can make annular flow to form early and to retard to diminish in quality direction at nominal oil concentration {>=}3%. Augmentation effect of oil on heat transfer coefficient becomes weakened or even diminishes for small diameter tube while detrimental effect of oil on small tube performance becomes more significant than large tube. For both test tubes, variation of heat transfer coefficient and enhanced factor with oil concentration is irregular. Two-phase heat transfer multiplier with refrigerant-oil mixture properties increases consistently and monotonically with local oil concentration at different vapor quality. (author)

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

    SciTech Connect (OSTI)

    Not Available

    1993-05-13

    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.

  2. Neutrino nuclear response and photo nuclear reaction

    E-Print Network [OSTI]

    H. Ejiri; A. I. Titov; M. Boswell; A. Young

    2013-11-10

    Photo nuclear reactions are shown to be used for studying neutrino/weak nuclear responses involved in astro-neutrino nuclear interactions and double beta decays. Charged current weak responses for ground and excited states are studied by using photo nuclear reactions through isobaric analog states of those states, while neutral current weak responses for excited states are studied by using photo nuclear reactions through the excited states. The weak interaction strengths are studied by measuring the cross sections of the photo nuclear reactions, and the spin and parity of the state are studied by measuring angular correlations of particles emitted from the photo nuclear reactions. Medium-energy polarized photons obtained from laser photons scattered off GeV electrons are very useful. Nuclear responses studied by photo nuclear reactions are used to evaluate neutrino/weak nuclear responses, i.e. nuclear beta and double beta matrix elements and neutrino nuclear interactions, and to verify theoretical calculations for them.

  3. Nuclear Golf

    E-Print Network [OSTI]

    Hacker, Randi; Tsutsui, William

    2006-12-06

    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...

  4. Collecting and recirculating condensate in a nuclear reactor containment

    DOE Patents [OSTI]

    Schultz, T.L.

    1993-10-19

    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.

  5. Collecting and recirculating condensate in a nuclear reactor containment

    DOE Patents [OSTI]

    Schultz, Terry L. (Murrysville Boro, PA)

    1993-01-01

    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.

  6. Nuclear Security Summit | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

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

  7. 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...

  8. Nuclear Waste: Forever Contaminated?

    E-Print Network [OSTI]

    Wang, Andrew

    2015-01-01

    Beyond Fukushima: Disasters, nuclear energy, and energy law.Nuclear Energy, and Energy Law (December 20, 2011). Brigham

  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.

    2000-08-01

    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. NRC (Nuclear Regulatory Commission) staff evaluation of the General Electric Company Nuclear Reactor Study (''Reed Report'')

    SciTech Connect (OSTI)

    1987-07-01

    In 1975, the General Electric Company (GE) published a Nuclear Reactor Study, also referred to as ''the Reed Report,'' an internal product-improvement study. GE considered the document ''proprietary'' and thus, under the regulations of the Nuclear Regulatory Commission (NRC), exempt from mandatory public disclosure. Nonetheless, members of the NRC staff reviewed the document in 1976 and determined that it did not raise any significant new safety issues. The staff also reached the same conclusion in subsequent reviews. However, in response to recent inquiries about the report, the staff reevaluated the Reed Report from a 1987 perspective. This re-evaluation, documented in this staff report, concluded that: (1) there are no issues raised in the Reed Report that support a need to curtail the operation of any GE boiling water reactor (BWR); (2) there are no new safety issues raised in the Reed Report of which the staff was unaware; and (3) although certain issues addressed by the Reed Report are still being studied by the NRC and the industry, there is no basis for suspending licensing and operation of GE BWR plants while these issues are being resolved.

  11. OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

    SciTech Connect (OSTI)

    Gauld, Ian C; Sly, Nicholas C; Michel-Sendis, Franco

    2014-01-01

    Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

  12. 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

    2012-01-01

    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 ...

  13. Nuclear lamins: building blocks of nuclear architecture

    E-Print Network [OSTI]

    Goldman, Robert D.

    REVIEW Nuclear lamins: building blocks of nuclear architecture Robert D. Goldman,1,3,4 Yosef Biological Laboratory, Woods Hole, Massachusetts 02543, USA Nuclear lamins were initially identified as the major components of the nuclear lamina, a proteinaceous layer found at the interface between chromatin

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

    E-Print Network [OSTI]

    Kroenig, Matthew

    2006-01-01

    of selective nuclear proliferation. Journal of Conflictmissile and nuclear proliferation: Issues for Congress. CRSSpector, L. 1988. Nuclear proliferation today. Cambridge,

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

    SciTech Connect (OSTI)

    Dallman, R J; Gottula, R C; Holcomb, E E; Jouse, W C; Wagoner, S R; Wheatley, P D

    1987-05-01

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

  16. Towards consistent nuclear models and comprehensive nuclear data...

    Office of Scientific and Technical Information (OSTI)

    Towards consistent nuclear models and comprehensive nuclear data evaluations Citation Details In-Document Search Title: Towards consistent nuclear models and comprehensive nuclear...

  17. Research Areas | Nuclear Science | ORNL

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

    Fuel Cycle Science & Technology Fusion Nuclear Science Isotope Development and Production Nuclear Security Science & Technology Nuclear Systems Modeling, Simulation & Validation...

  18. 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.

    2014-03-31

    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.

  19. Nuclear Photonics

    E-Print Network [OSTI]

    D. Habs; M. M. Guenther; M. Jentschel; P. G. Thirolf

    2012-01-21

    With new gamma-beam facilities like MEGa-ray at LLNL (USA) or ELI-NP at Bucharest with 10^13 g/s and a bandwidth of Delta E_g/E_g ~10^-3, a new era of g-beams with energies <=20 MeV comes into operation, compared to the present world-leading HIGS facility (Duke Univ., USA) with 10^8 g/s and Delta E_g/E_g~0.03. Even a seeded quantum FEL for g-beams may become possible, with much higher brilliance and spectral flux. At the same time new exciting possibilities open up for focused g-beams. We describe a new experiment at the g-beam of the ILL reactor (Grenoble), where we observed for the first time that the index of refraction for g-beams is determined by virtual pair creation. Using a combination of refractive and reflective optics, efficient monochromators for g-beams are being developed. Thus we have to optimize the system of the g-beam facility, the g-beam optics and g-detectors. We can trade g-intensity for band width, going down to Delta E_g/E_g ~ 10^-6 and address individual nuclear levels. 'Nuclear photonics' stresses the importance of nuclear applications. We can address with g-beams individual nuclear isotopes and not just elements like with X-ray beams. Compared to X rays, g-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 micron resolution using Nucl. Reson. Fluorescence for detection with eV resolution and high spatial resolution. We discuss the dominating M1 and E1 excitations like 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.

  20. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTechtail.Theory ofDid you notHeatMaRIEdioxideUser Work FeaturedNuclear

  1. NUCLEAR ENERGY

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

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergyTher i nAand DOEDepartment of Energy009At26-2009NSRC_MOU.pdffactsNUCLEAR ENERGY

  2. Reconversion of nuclear weapons

    E-Print Network [OSTI]

    Kapitza, Sergei P

    1993-01-01

    The nuclear predicament or nuclear option. Synopsis of three lectures : 1- The physical basis of nuclear technology. Physics of fission. Chain reaction in reactors and weapons. Fission fragments. Separration of isotopes. Radiochemistry.2- Nuclear reactors with slow and fast neutrons. Power, size, fuel and waste. Plutonium production. Dose rate, shielding and health hazard. The lessons of Chernobyl3- Nuclear weapons. Types, energy, blast and fallout. Fusion and hydrogen bombs. What to do with nuclear weapons when you cannot use them? Testing. Nonmilittary use. Can we get rid of the nuclear weapon? Nuclear proliferation. Is there a nuclear future?

  3. Nuclear Safety | Department of Energy

    Office of Environmental Management (EM)

    Nuclear Safety Nuclear Safety The Office of Nuclear Safety establishes and maintains nuclear safety policy, requirements, and guidance including policy and requirements relating to...

  4. Nuclear Data | More Science | ORNL

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

    Data SHARE Nuclear Data Nuclear Data ORNL is a recognized, international leader in nuclear data research and development (R&D) to support nuclear applications analyses. For more...

  5. Nuclear Sciences | More Science | ORNL

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

    Nuclear Sciences SHARE Nuclear Sciences In World War II's Manhattan Project, ORNL helped usher in the nuclear age. Today, laboratory scientists are leaders in using nuclear...

  6. Nuclear Explosive Safety

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

    2014-07-10

    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).

  7. NUCLEAR DEFORMATION ENERGIES

    E-Print Network [OSTI]

    Blocki, J.

    2009-01-01

    J.R. Nix, Theory of Nuclear Fission and Superheavy Nuclei,energy maps relevant for nuclear fission and nucleus-nucleusof macroscopic aspects of nuclear fission and of collisions

  8. Nuclear Waste: Forever Contaminated?

    E-Print Network [OSTI]

    Wang, Andrew

    2015-01-01

    into when undergoing nuclear fission. 175-3000 times higheranother byproduct of nuclear fission, but that will receiveNuclear Energy, and Energy Law (December 20, 2011). Brigham Young University Law Review, Fission

  9. NUCLEAR STRUCTURE DATABASE

    E-Print Network [OSTI]

    Firestone, R.B.

    2010-01-01

    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

  10. RELATIVISTIC NUCLEAR COLLISIONS: THEORY

    E-Print Network [OSTI]

    Gyulassy, M.

    2010-01-01

    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.

  11. Nuclear Waste: Forever Contaminated?

    E-Print Network [OSTI]

    Wang, Andrew

    2015-01-01

    Went Wrong in Japan’s Nuclear Reactors. Retrieved March 28,went-wrong-in-japans-nuclear-reactors World Statistics. (nuclear disaster since Chernobyl. Chernobyl happened on April 26, 1986, when a reactor

  12. 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.

  13. Nuclear Fuel Cycle & Vulnerabilities

    SciTech Connect (OSTI)

    Boyer, Brian D.

    2012-06-18

    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.

  14. Hollow nuclear matter

    E-Print Network [OSTI]

    Gao-Chan Yong

    2015-12-18

    It is generally considered that an atomic nucleus is always compact. Based on the isospin-dependent Boltzmann nuclear transport model, here I show that large block nuclear matter or excited nuclear matter may both be hollow. And the size of inner bubble in these matter is affected by the charge number of nuclear matter. Existence of hollow nuclear matter may have many implications in nuclear or atomic physics or astrophysics as well as some practical applications.

  15. Hollow nuclear matter

    E-Print Network [OSTI]

    Yong, Gao-Chan

    2015-01-01

    It is generally considered that an atomic nucleus is always compact. Based on the isospin-dependent Boltzmann nuclear transport model, here I show that large block nuclear matter or excited nuclear matter may both be hollow. And the size of inner bubble in these matter is affected by the charge number of nuclear matter. Existence of hollow nuclear matter may have many implications in nuclear or atomic physics or astrophysics as well as some practical applications.

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

    E-Print Network [OSTI]

    Quiter, Brian

    2012-01-01

    Security of the National Nuclear Security Administration, USof Energys National Nuclear Security Administration (NNSA)

  17. Nuclear weapons modernizations

    SciTech Connect (OSTI)

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

    2014-05-09

    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.

  18. 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...

  19. Nuclear Waste: Forever Contaminated?

    E-Print Network [OSTI]

    Wang, Andrew

    2015-01-01

    Fallout that spread across Fukushima R eferences Buesseler,M. (2011). Impacts of the Fukushima nuclear power plants onL. L. (2011). Beyond Fukushima: Disasters, nuclear energy,

  20. NUCLEAR DEFORMATION ENERGIES

    E-Print Network [OSTI]

    Blocki, J.

    2009-01-01

    nuclear energies in the absence of a proximity contribution.contributions represent the major part of the potential energy of a nuclear

  1. 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.

  2. Nuclear Energy Advisory Committee

    Broader source: Energy.gov [DOE]

    The Nuclear Energy Advisory Committee (NEAC), formerly the Nuclear Energy Research Advisory Committee (NERAC), was established on October 1, 1998, to provide independent advice to the Office of...

  3. 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...

  4. Nuclear Forensics | 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...

  5. Nuclear Nonproliferation Treaty | National Nuclear Security Administra...

    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...

  6. 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...

  7. Nuclear & Radiological Material Removal | National Nuclear Security...

    National Nuclear Security Administration (NNSA)

    & Radiological Material Removal | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation...

  8. 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...

  9. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

    Douglas M. Gerstner

    2009-05-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 “flux traps” (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop’s temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation.

  10. Financing the Nuclear Renaissance

    E-Print Network [OSTI]

    Nuttall, William J.; Taylor, S.

    is the decision made in 2002 to restart Browns Ferry unit 1 in Alabama USA [5]. The background to that story is: Browns Ferry Unit 1 is a Boiling Water Reactor operational from 1973. It was shutdown by fire for one year in 1975. The fire prompted much general... pellets and carefully canned and prepared in fuel assemblies appropriate to the reactor type. As such, for a Pressurised Water Reactor operating a “once-through” fuel cycle only roughly one quarter of the fuel costs relate to the yellowcake precursor [3...

  11. The Joys of Nuclear Engineering

    ScienceCinema (OSTI)

    Jon Carmack

    2010-01-08

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

  12. 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

  13. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B. (Severna Park, MD)

    2004-03-16

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

  14. 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 the United States and the Soviet Union, smoke from the fires started by nuclear weapons, especially the black

  15. INSTRUCTIONS FOR SUBMITTING NUCLEAR

    E-Print Network [OSTI]

    CALIFORNIA ENERGY COMMISSION INSTRUCTIONS FOR SUBMITTING NUCLEAR POWER PLANT-RELATED 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

  16. Advanced nuclear fuel

    ScienceCinema (OSTI)

    Terrani, Kurt

    2014-07-15

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

  17. Nuclear Engineering Program Ranking

    E-Print Network [OSTI]

    Evans, Paul G.

    Nuclear Engineering Program Ranking 2 Enrollment Approximately 200 undergraduate students and 120 in Nuclear Engineering (BS) · Bachelor of Science in Engineering Physics (BS) · Master of Science in Nuclear Engineering and Engineering Physics (MS) · Doctor of Philosophy in Nuclear Engineering and Engineering Physics

  18. Advanced nuclear fuel

    SciTech Connect (OSTI)

    Terrani, Kurt

    2014-07-14

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

  19. Nuclear Structure Thomas Neff

    E-Print Network [OSTI]

    Neff, Thomas

    Nuclear Structure in FMD Thomas Neff #12;FMD attributes Fermionic ^Q = C A q1 ··· qA Unitary = ^Q H ^Q ^Q ^Q Nuclear Structure in FMD Thomas Neff­ September 2, 1998 #12;Nuclear Interactions-interactions Nuclear Structure in FMD Thomas Neff­ September 2, 1998 #12;Unitary Correlator How to address the hard

  20. Nuclear Structure Thomas Neff

    E-Print Network [OSTI]

    Neff, Thomas

    Nuclear Structure in FMD Thomas Neff #12; FMD attributes Fermionic #12; #12; â?? Q #11; =C #24; A #24 #11; Nuclear Structure in FMD Thomas Neff-- September 2, 1998 #12; Nuclear Interactions Effective­interactions Nuclear Structure in FMD Thomas Neff-- September 2, 1998 #12; Unitary Correlator How to address the hard

  1. Nuclear Reactions Some Basics

    E-Print Network [OSTI]

    Gilfoyle, Jerry

    Nuclear Reactions Some Basics I. Reaction Cross Sections #12;Common Units in Nuclear Physics sphere: = (4r2)/r2 = 4 (sr)steradians r A 2 = (r)radians r s = r A O s r O #12;Types of Nuclear Reactions · When a particle strikes a nucleus, the resulting interaction is referred to as a "nuclear

  2. 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)

    2007-07-01

    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)

  3. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    this man- (PWRs) and boiling-water reactors to compare thesepressur- ized and boiling water reactors is presented. ItAmerican Physical reactor cooling - boiling-water ECC(S) -

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

    SciTech Connect (OSTI)

    NONE

    2000-08-01

    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.

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

    E-Print Network [OSTI]

    Kroenig, Matthew

    2006-01-01

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

  6. Dynamics of nuclear envelope and nuclear pore complex formation

    E-Print Network [OSTI]

    Anderson, Daniel J.

    2008-01-01

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

  7. Dynamics of nuclear envelope and nuclear pore complex formation

    E-Print Network [OSTI]

    Anderson, Daniel J.

    2008-01-01

    instability due to nuclear fusions. The universal action ofnegatively regulates nuclear membrane fusion and nuclearrequired for vesicle fusion during nuclear envelope assembly

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

    E-Print Network [OSTI]

    Kroenig, Matthew

    2006-01-01

    nature of the nuclear recipient’s security environment. ThisKeywords: Nuclear weapons proliferation; security; securitynature of the nuclear recipient’s security environment. This

  9. NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 05 NUCLEAR FUELS...

    Office of Scientific and Technical Information (OSTI)

    Title list of documents made publicly available, January 1-31, 1998 NONE 21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 05 NUCLEAR FUELS; BIBLIOGRAPHIES; NUCLEAR POWER PLANTS;...

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

    E-Print Network [OSTI]

    Kroenig, Matthew

    2006-01-01

    T. 1993. The Nuclear Suppliers Group. Nonproliferationeds. 1985. The nuclear suppliers and nonproliferation:of the emerging nuclear suppliers. Lexington, MA: Lexington

  11. Thermoacoustic Thermometry for Nuclear Reactor Monitoring

    SciTech Connect (OSTI)

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

    2013-06-01

    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?

  12. Nuclear Detonation Detection | National Nuclear Security Administratio...

    National Nuclear Security Administration (NNSA)

    the entire planet from space to detect and report surface, atmospheric, or space nuclear detonations; produces and updates the regional geophysical datasets enabling...

  13. 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...

  14. 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...

  15. 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...

  16. 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"...

  17. Nuclear Suppliers Group & Regimes | National Nuclear Security...

    National Nuclear Security Administration (NNSA)

    Suppliers Group & Regimes | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the...

  18. Nuclear Structure and Nuclear Reactions | Argonne Leadership...

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

    x2 - triaxiality, and x3 - pairing correlations. Calculations were carried out using nuclear density functional theory. The collective action was minimized using the dynamical...

  19. Neutronic evaluation of a non-fertile fuel for the disposition of weapons-grade plutonium in a boiling water reactor

    SciTech Connect (OSTI)

    Sterbentz, J.W.

    1994-10-01

    A new non-fertile, weapons-grade plutonium oxide fuel concept is developed and evaluated for deep burn applications in a boiling water reactor environment using the General Electric 8x8 Advanced Boiling Water Reactor (ABWR) fuel assembly dimensions and pitch. Detailed infinite lattice fuel burnup results and neutronic performance characteristics are given and although preliminary in nature, clearly demonstrate the fuel`s potential as an effective means to expedite the disposition of plutonium in existing light water reactors. The new non-fertile fuel concept is an all oxide composition containing plutonia, zirconia, calcia, and erbia having the following design weight percentages: 8.3; 80.4; 9.7; and 1.6. This fuel composition in an infinite fuel lattice operating at linear heat generation rates of 6.0 or 12.0 kW/ft per rod can remain critical for up to 1,200 and 600 Effective Full Power Days (EFPD), respectively, and achieve a burnup of 7.45 {times} 10{sup 20} f/cc. These burnups correspond to a 71--73% total plutonium isotope destruction and a 91--94% destruction of the {sup 239}Pu isotope for the 0--40% moderator steam void condition. Total plutonium destruction greater than 73% is possible with a fuel management scheme that allows subcritical fuel assemblies to be driven by adjacent high reactivity assemblies. The fuel exhibits very favorable neutron characteristics from beginning-of-life (BOL) to end-of-life (EOL). Prompt fuel Doppler coefficient of reactivity are negative, with values ranging between {minus}0.4 to {minus}2.0 pcm/K over the temperature range of 900 to 2,200 K. The ABWR fuel lattice remains in an undermoderated condition for both hot operational and cold startup conditions over the entire fuel burnup lifetime.

  20. China's Nuclear Industry After Fukushima

    E-Print Network [OSTI]

    YUAN, Jingdong

    2013-01-01

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

  1. China's Nuclear Industry After Fukushima

    E-Print Network [OSTI]

    YUAN, Jingdong

    2013-01-01

    generation of Chinese nuclear submarines continues to sufferalready) benefit its nuclear submarine propulsion. Forwas based on the naval submarine nuclear reactor. There have

  2. Is Nuclear Energy the Solution?

    E-Print Network [OSTI]

    Saier, Milton H.; Trevors, Jack T.

    2010-01-01

    first, investments in nuclear power are risky as indicatedto stay clear; second, nuclear power plants are statedrisks of their own; third, nuclear power will not reduce our

  3. Is Nuclear Energy the Solution?

    E-Print Network [OSTI]

    Saier, Milton H.; Trevors, Jack T.

    2010-01-01

    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)

  4. Nuclear Sciences | More Science | ORNL

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

    Sciences SHARE Nuclear Sciences In World War II's Manhattan Project, ORNL helped usher in the nuclear age. Today, laboratory scientists are leaders in using nuclear technologies...

  5. Nuclear Science | More Science | ORNL

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

    Science SHARE Nuclear Science In World War II's Manhattan Project, ORNL helped usher in the nuclear age. Today, laboratory scientists are leaders in using nuclear technologies and...

  6. NUCLEAR SCIENCE ANNUAL REPORT 1975

    E-Print Network [OSTI]

    Authors, Various

    2010-01-01

    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

    2013-01-01

    power plants must meet nuclear safety standards and adoptapplications; review of nuclear safety regula- tions; andpower development plans. Nuclear safety was placed front and

  8. Sandia Energy - Nuclear Fuel Cycle Options Catalog

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

    Nuclear Fuel Cycle Options Catalog Home Stationary Power Nuclear Fuel Cycle Nuclear Energy Workshops Nuclear Fuel Cycle Options Catalog Nuclear Fuel Cycle Options CatalogAshley...

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

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

    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...

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

    SciTech Connect (OSTI)

    John A. Schroeder; Gordon R. Bower

    2014-02-01

    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.

  11. Nuclear reactions and nuclear ssion March 19, 2002

    E-Print Network [OSTI]

    Landstreet, John D.

    Nuclear reactions and nuclear ¯ssion March 19, 2002 1 Cross sections and reaction rates ² Consider

  12. Hegemony and nuclear proliferation

    E-Print Network [OSTI]

    Miller, Nicholas L. (Nicholas LeSuer)

    2014-01-01

    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 ...

  13. Nuclear information systems

    SciTech Connect (OSTI)

    Cowley, R.L. (Enercon Services, Inc., Broken Arrow, OK (US))

    1989-01-01

    This paper reviews options, challenges, and opportunities facing the nuclear industry as it strives to achieve the most effective use of the available resources in support of managing the nuclear power plant business.

  14. Promulgating Nuclear Safety Requirements

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

    1996-05-15

    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.

  15. American Nuclear Society Awards

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

    Nuclear Society Awards Established in 1999 by the Fusion Energy Division of the American Nuclear Society (ANS) and named after LLNL's co-founder, the Edward Teller Medal recognizes...

  16. NUCLEAR PROXIMITY FORCES

    E-Print Network [OSTI]

    Randrup, J.

    2011-01-01

    usual series expansions of the nuclear energy. additionalnuclear energies than the calcula- tion of the above series expansions

  17. 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...

  18. Nuclear Explosive Safety Manual

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

    2009-04-14

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

  19. 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/ 

  20. Abnormal event identification in nuclear power plants using a neural network and knowledge processing

    SciTech Connect (OSTI)

    Ohga, Yukiharu; Seki, Hiroshi (Hitachi, Ltd. Energy Research Lab., Ibarakiken (Japan))

    1993-02-01

    The combination of a neural network and knowledge processing have been used to identify abnormal events that cause a reactor to scram in a nuclear power plant. The neural network recognizes the abnormal event from the change pattern of analog data for state variables, and this result is confirmed from digital data using a knowledge base of plant status when each event occurs. The event identification method is tested using test data based on simulated results of a transient analysis program for boiling water reactors. It is confirmed that a neural network can identify an event in which it has been trained even when the plant conditions, such as fuel burnup, differ from those used in the training and when the analog data contain white noise. The network does not mistakenly identify the nontrained event as a trained one. The method is feasible for event identification, and knowledge processing improves the reliability of the identification.

  1. Nuclear Explosive Safety

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

    2009-04-14

    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.

  2. Nuclear Explosive Safety

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

    2015-01-26

    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). Supersedes DOE O 452.2D and DOE M 452.2-1A.

  3. Nuclear Explosive Safety

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

    2009-04-14

    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

  4. NUCLEAR POWER in CALIFORNIA

    E-Print Network [OSTI]

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

  5. OPPORTUNITIES NUCLEAR ASTROPHYSICS

    E-Print Network [OSTI]

    Pfrender, Michael

    i OPPORTUNITIES IN NUCLEAR ASTROPHYSICS ORIGIN OF THE ELEMENTS #12;ii 30 SEPTEMBER, 1999 OPPORTUNITIES IN NUCLEAR ASTROPHYSICS Conclusions of a Town Meeting held at the University of Notre Dame 7-8 June 1999 #12;iii PREFACE A Town Meeting on Opportunities in Nuclear Astrophysics was held

  6. NUCLEAR PLANT AND CONTROL

    E-Print Network [OSTI]

    NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: software require- ments, safety analysis, formal for the digital protection systems of a nuclear power plant. When spec- ifying requirements for software and CRSA processes are described using shutdown system 2 of the Wolsong nuclear power plants as the digital

  7. 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...

  8. China's Nuclear Industry After Fukushima

    E-Print Network [OSTI]

    YUAN, Jingdong

    2013-01-01

    continue to pursue nuclear expansion as part of an energythe rapid expansion of China’s nuclear industry requires a

  9. Is Nuclear Energy the Solution?

    E-Print Network [OSTI]

    Saier, Milton H.; Trevors, Jack T.

    2010-01-01

    last 50 years, nuclear energy subsidies have totaled nearlyof subsidies. Never- theless, claims that nuclear power is a

  10. IMPROVISED NUCLEAR DEVICE An Improvised Nuclear Device (IND) is a type of nuclear weapon.

    E-Print Network [OSTI]

    IMPROVISED NUCLEAR DEVICE An Improvised Nuclear Device (IND) is a type of nuclear weapon. When of an Improvised Nuclear Device? An IND would cause great destruction, death, and injury and have a wide area

  11. Nuclear spirals in galaxies

    E-Print Network [OSTI]

    Witold Maciejewski

    2006-11-08

    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.

  12. Commercial nuclear power 1990

    SciTech Connect (OSTI)

    Not Available

    1990-09-28

    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.

  13. Use of an influence diagram and fuzzy probability for evaluating accident management in a boiling water reactor

    SciTech Connect (OSTI)

    Yu, D.; Kastenberg, W.E.; Okrent, D. (Univ. of California, Los Angeles, CA (United States). Mechanical, Aerospace, and Nuclear Engineering Dept.)

    1994-06-01

    A new approach is presented for evaluating the uncertainties inherent in severe accident management strategies. At first, this analysis considers accident management as a decision problem (i.e., applying a strategy compared with do nothing) and uses an influence diagram. To evaluate imprecise node probabilities in the influence diagram, the analysis introduces the concept of a fuzzy probability. When fuzzy logic is applied, fuzzy probabilities are easily propagated to obtain results. In addition, the results obtained provide not only information similar to the classical approach, which uses point-estimate values, but also additional information regarding the impact of using imprecise input data. As an illustrative example, the proposed methodology is applied to the evaluation of the drywell flooding strategy for a long-term station blackout sequence at the Peach Bottom nuclear power plant. The results show that the drywell flooding strategy is beneficial for preventing reactor vessel breach. It is also effective for reducing the probability of containment failure for both liner melt-through and late overpressurization. Even though uncertainty exists in the results, flooding is preferred to do nothing when evaluated in terms of two risk measures: early and late fatalities.

  14. Office of Defense Nuclear Nonproliferation | National Nuclear...

    National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration (DOENNSA) and the U.K. Department of Energy and Climate Change concluded a workshop at Wilton Park, About This Site Budget IG Web Policy...

  15. 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

    2004-01-01

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

  16. 24. Nuclear? Nuclear power comes in two flavours

    E-Print Network [OSTI]

    Ernst, Damien

    24. Nuclear? Nuclear power comes in two flavours: 1. Nuclear fission. Split up heavy nuclei into medium-sized nu- clei, thereby releasing energy. 2. Nuclear fusion. Fuse light nuclei into medium-sized nuclei, thereby releasing energy. 1 #12;Nuclear energy available per atom is roughly one mil- lion times

  17. FUSION NUCLEAR SCIENCE PROGRAM & SUPPORTING FUSION NUCLEAR SCIENCE FACILITY (FNSF)

    E-Print Network [OSTI]

    FUSION NUCLEAR SCIENCE PROGRAM & SUPPORTING FUSION NUCLEAR SCIENCE FACILITY (FNSF): UPDATE in order to apply the knowledge we gained about burning plasma state #12;FUSION NUCLEAR SCIENCE PROGRAM #12;FUSION NUCLEAR SCIENCE PROGRAM DEVELOPMENT OF ELEMENTS OF THE FUSION NUCLEAR SCIENCE PROGRAM

  18. Instrumentation for Nuclear Applications

    SciTech Connect (OSTI)

    NONE

    1998-09-18

    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.

  19. Nuclear Science References Database

    E-Print Network [OSTI]

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

    2014-07-08

    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.

  20. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    SciTech Connect (OSTI)

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  1. Security Science & Technology | Nuclear Science | ORNL

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

    Security Treaty Verification Nuclear Systems Modeling, Simulation & Validation Nuclear Systems Technology Reactor Technology Nuclear Science Home | Science & Discovery |...

  2. Absolute nuclear material assay

    DOE Patents [OSTI]

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

    2012-05-15

    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.

  3. Absolute nuclear material assay

    DOE Patents [OSTI]

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

    2010-07-13

    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.

  4. Nuclear Fabrication Consortium

    SciTech Connect (OSTI)

    Levesque, Stephen

    2013-04-05

    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.

  5. Application of PSA to review and define technical specifications for advanced nuclear power plants

    SciTech Connect (OSTI)

    Kim, I.S.; Samanta, P.K.; Reinhart, F.M.; Wohl, M.L.

    1995-11-01

    As part of the design certification process, probabilistic safety assessments (PSAS) are performed at the design stage for each advanced nuclear power plant. Among other usages, these PSAs are important inputs in defining the Technical Specifications (TSs) for these plants. Knowledge gained from their use in improving the TSs for operating nuclear power plants is providing methods and insights for using PSAs at this early stage. Evaluating the safety or the risk significance of the TSs to be defined for an advanced plant encompasses diverse aspects: (a) determining the basic limiting condition for operation (LCO); (b) structuring conditions associated with the LCO; (c) defining completion times (equivalent to allowed outage times in the TS for conventional plants); and, (d) prescribing required actions to be taken within the specified completion times. In this paper, we consider the use of PSA in defining the TSs for an advanced nuclear plant, namely General Electric`s Advanced Boiling Water Reactor (ABWR). Similar approaches are being taken for ABB-CE`s System 80+ and Westinghouse`s AP-600. We discuss the general features of an advanced reactor`s TS, how PSA is being used in reviewing the TSs, and we give an example where the TS submittal was reviewed using a PSA-based analysis to arrive at the requirements for the plant.

  6. Nuclear Materials Science

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

    comprises the core actinide materials science and metallurgical capability within the nuclear weapons production and surveillance communities. Contact Us Group Leader David...

  7. Office Of Nuclear Energy

    Energy Savers [EERE]

    irradiation effect * Demonstrate the TEG-powered WSN prototype 3 Background and motivation * TEG is very compact and reliable * Heat sources are very abundant in nuclear power...

  8. Nuclear reactor apparatus

    DOE Patents [OSTI]

    Wade, Elman E. (Ruffs Dale, PA)

    1978-01-01

    A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

  9. Management of Nuclear Materials

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

    2009-08-17

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

  10. Nuclear Physics: Experiment Research

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

    search Nuclear Physics Program Please upgrade your browser. This site's design is only visible in a graphical browser that supports web standards, but its content is accessible to...

  11. Nuclear Physics Program

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

    Experimental Halls Hall A Hall B Hall C Hall D Physics Departments Administrative Office Data Acquisition Group Detector & Imaging Group Electronics Group User Liaison Nuclear...

  12. Office Of Nuclear Energy

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

    Attributes of Software-Based Safety Critical Instrumentation and Control Systems in Nuclear Power Plants) (Carol Smidts) (The Ohio State University) (NEET 2) October 28-29, 2015...

  13. 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...

  14. defense nuclear security

    National Nuclear Security Administration (NNSA)

    3%2A en Defense Nuclear Security http:www.nnsa.energy.govaboutusourprogramsnuclearsecurity

  15. Natlonal Nuclear Security Admlnlstratlon

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

    Natlonal Nuclear Security Admlnlstratlon Washington, DC 20585 January 18, 2007 OFFICE OF THE ADMINISTRATOR MEMORANDUM FOR The Deputy Secretary FROM: SUBJECT: Linton F. Brooks...

  16. National Nuclear Security Administration

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

    Washington. DC 20585 DEC 2 0 2010 MEMORANDUM FOR DENNIS M. MIOTLA THRU: FROM: SUBJECT: DEPUTY ASSISTANT SECRETARY FOR NUCLEAR FACILITY OPERATIONS PETE B. LYONS fJ ' ... ,-)-...

  17. Office Of Nuclear Energy

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

    Office Of Nuclear Energy Sensors and Instrumentation Annual Review Meeting Enhanced Micro-Pocket Fission Detector (MPFD) for High Temperature Reactors Troy Unruh Idaho National...

  18. Unclassified Controlled Nuclear Information

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

    1995-09-25

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

  19. Nuclear Physics: Experiment Research

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

    Research Highlights Public Interest Nuclear Physics Accelerator Free Electron Laser (FEL) Medical Imaging Physics Topics Campaigns Meetings Recent Talks Archived Talks...

  20. Nuclear Physics from QCD

    E-Print Network [OSTI]

    U. van Kolck

    2008-12-20

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

  1. Reference handbook: Nuclear criticality

    SciTech Connect (OSTI)

    Not Available

    1991-12-06

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

  2. Nuclear Spectra from Skyrmions

    SciTech Connect (OSTI)

    Manton, N. S.

    2009-08-26

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

  3. 2013 Nuclear Workforce Development ...

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

    all describe Suzy Hobbs Baker. Hear and ask questions about her experience traveling Europe as a nuclear tourist. The Babcock & Wilcox Company Suzy Hobbs Baker Founder of...

  4. Nuclear Regulatory Commission | Department of Energy

    Office of Environmental Management (EM)

    Commission Nuclear Regulatory Commission Nuclear Regulatory Commission More Documents & Publications What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear...

  5. Sandia Energy - Nuclear Fuel Cycle Options Catalog

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

    Fuel Cycle Options Catalog Home Stationary Power Nuclear Fuel Cycle Advanced Nuclear Energy Nuclear Fuel Cycle Options Catalog Nuclear Fuel Cycle Options CatalogAshley...

  6. Generation IV Nuclear Energy Systems ...

    E-Print Network [OSTI]

    Kemner, Ken

    Generation IV Nuclear Energy Systems ... The U.S. Department of Energy's Office of Nuclear Energy enhance safety and security, and develop nuclear power as an energy source for industrial applications Information ... U.S. Department of Energy www.energy.gov DOE Office of Nuclear Energy www.nuclear

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

    E-Print Network [OSTI]

    Kroenig, Matthew

    2006-01-01

    ambitions: The spread of nuclear weapons 1989-1990. Boulder:Determinants of nuclear weapons proliferation. UnpublishedWhy nations forgo nuclear weapons. Montreal: McGill-Queen’s

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

    E-Print Network [OSTI]

    Quiter, Brian

    2012-01-01

    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

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

    E-Print Network [OSTI]

    Pázsit, Imre

    2000-01-01

    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

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

    E-Print Network [OSTI]

    Kroenig, Matthew

    2006-01-01

    1982. Yugoslavia. In Nuclear power in developing countries,Potter, W.C. 1982. Nuclear power and nonproliferation: AnNorth Korea is coded as nuclear power beginning in 1993.

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

    E-Print Network [OSTI]

    Kroenig, Matthew

    2006-01-01

    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.

  12. Advanced Review PSF: nuclear busy-body or nuclear

    E-Print Network [OSTI]

    Lynch, Kristen W.

    Advanced Review PSF: nuclear busy-body or nuclear facilitator? Christopher A. Yarosh,1 Joseph R of multifunc- tional nuclear factors termed DBHS (Drosophila behavior human splicing) proteins, which also

  13. Int. J. Nuclear Energy Science and Technology, Vol. 2, No. 3, 2006 167 Copyright 2006 Inderscience Enterprises Ltd.

    E-Print Network [OSTI]

    Demazière, Christophe

    -based methods, the purpose of which is to determine whether the core of a Boiling Water Reactor (BWR) contains coexisting with a regional oscillation. Keywords: neutron noise analysis; Boiling Water Reactor (BWR include pressurised water reactor and boiling water reactor physics, mixed oxide fuel, power reactor noise

  14. NUCLEAR SCIENCE AND ENGINEERING: 72, 304-321 ( 1 9 7 9 ) Linear Response of the Neutron Field to a

    E-Print Network [OSTI]

    Pázsit, Imre

    to a Propagating Perturbation of Moderator Density (Two-Group Theory of Boiling Water Reactor Noise) K. Behringer,1979 Accepted July 2,1979 According to recent investigations, the neutron noise field in a boiling water reactor for a two-phase flow of the coolant in boiling water reactor (BWR) systems for frequencies higher than 2 Hz

  15. 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

    of a Boiling Water Reactor (BWR) contains unseated fuel assembly(ies) or not. The characteristic feature. Keywords: neutron noise analysis; Boiling Water Reactor (BWR); stability; Decay Ratio (DR); space and boiling water reactor physics, mixed oxide fuel, power reactor noise and noise diagnostics, signal

  16. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal Gas &SCE-SessionsSouthReporteeo | National NuclearaCSGF MagazineNuclearChernobyl Nuclear

  17. National Nuclear Science Week Day 2: NNSA Showcases Nuclear Science...

    National Nuclear Security Administration (NNSA)

    2: NNSA Showcases Nuclear Science Careers | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing...

  18. NNSA Celebrates National Nuclear Science Week | National Nuclear...

    National Nuclear Security Administration (NNSA)

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

  19. National Nuclear Science Week - Jan. 24-28 | National Nuclear...

    National Nuclear Security Administration (NNSA)

    - Jan. 24-28 | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy...

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

    E-Print Network [OSTI]

    Kroenig, Matthew

    2006-01-01

    Without the bomb: The politics of nuclear nonproliferation.impact of nuclear weapons on international politics. Statesnuclear/missile ties and balance of power politics. The

  1. Naval Nuclear Propulsion Plants | National Nuclear Security Administra...

    National Nuclear Security Administration (NNSA)

    Naval Nuclear Propulsion Plants In naval nuclear propulsion plants, fissioning of uranium atoms in the reactor core produces heat. Because the fission process also produces...

  2. 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...

  3. JPRS report supplement: Nuclear developments. Iraq -- Nuclear and missile proliferation

    SciTech Connect (OSTI)

    NONE

    1990-09-14

    This document contains articles from foreign periodicals from throughout the world, translated into English, that concern nuclear developments, specifically nuclear and missile proliferation in Iraq.

  4. Physics of Nuclear Antishadowing

    E-Print Network [OSTI]

    Ivan Schmidt

    2005-11-14

    Shadowing and antishadowing of the electromagnetic nuclear structure functions are produced by the coherence of multiscattering quark nuclear processes. This picture leads to substantially different antishadowing for charged and neutral current processes, particularly in anti-neutrino reactions, thus affecting the extraction of the weak-mixing angle $\\sin^2\\theta_W$.

  5. Whither Nuclear Physics ?

    E-Print Network [OSTI]

    Syed Afsar Abbas

    2008-01-07

    Nuclear Physics has had its ups and downs. However in recent years, bucked up by some new and often puzzling data, it has become a potentially very rich field. We review some of these exciting developments in a few important sectors of nuclear physics. Emphasis shall be on the study of exotic nuclei and the new physics that these nuclei are teaching us.

  6. 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 of Reactor Physics SE-41296 Gothenburg, Sweden GABOR PÓR Budapest University of Technology and Economics H

  7. 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 of Reactor Physics SE-41296 Gothenburg, Sweden GABOR PÓR Budapest University of Technology and Economics H

  8. Nuclear Explosive Safety

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

    2006-06-12

    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.

  9. United States Nuclear Regulatory

    E-Print Network [OSTI]

    United States Nuclear Regulatory Commission United States Department of Energy United States.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: DISTRIBUTION@nrc.gov Facsimile: 301; and Commission papers and their attachments. NRC publications in the NUREG series, NRC regulations, and Title 10

  10. Nuclear Science & Engineering

    E-Print Network [OSTI]

    % of world electricity. #12;. 3 Nuclear Science & Engineering New Plants are coming on line Nuclear Power, Germany, Spain -1GW Construction initiation: · Olkiluoto 3 (1600 MW(e), EPR, Finland) - construction China: 2nd largest energy consumer 2004 Electricity growth 9% generation 16% demand so still... 30GW

  11. Sensing remote nuclear spins

    E-Print Network [OSTI]

    Nan Zhao; Jan Honert; Berhard Schmid; Junichi Isoya; Mathew Markham; Daniel Twitchen; Fedor Jelezko; Ren-Bao Liu; Helmut Fedder; Jörg Wrachtrup

    2012-04-29

    Sensing single nuclear spins is a central challenge in magnetic resonance based imaging techniques. Although different methods and especially diamond defect based sensing and imaging techniques in principle have shown sufficient sensitivity, signals from single nuclear spins are usually too weak to be distinguished from background noise. Here, we present the detection and identification of remote single C-13 nuclear spins embedded in nuclear spin baths surrounding a single electron spins of a nitrogen-vacancy centre in diamond. With dynamical decoupling control of the centre electron spin, the weak magnetic field ~10 nT from a single nuclear spin located ~3 nm from the centre with hyperfine coupling as weak as ~500 Hz is amplified and detected. The quantum nature of the coupling is confirmed and precise position and the vector components of the nuclear field are determined. Given the distance over which nuclear magnetic fields can be detected the technique marks a firm step towards imaging, detecting and controlling nuclear spin species external to the diamond sensor.

  12. 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)

    2014-05-09

    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.

  13. Supporting Our Nation's Nuclear Industry

    ScienceCinema (OSTI)

    Lyons, Peter

    2013-05-29

    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.

  14. China's Nuclear Industry After Fukushima

    E-Print Network [OSTI]

    YUAN, Jingdong

    2013-01-01

    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

  15. NUCLEAR SCIENCE ANNUAL REPORT 1975

    E-Print Network [OSTI]

    Authors, Various

    2010-01-01

    Graham, and K. P. Jackson, Chalk River Nuclear LaboratoriesFrance. *Present address: Chalk River Nuclear Laboratories,g Permanent address: Chalk River Nuclear Labora­ t o r i e

  16. China's Nuclear Industry After Fukushima

    E-Print Network [OSTI]

    YUAN, Jingdong

    2013-01-01

    on the future of China’s nuclear power. First, it highlightsas China builds more nuclear power plants. The challengesto manage, run, and inspect nuclear power plants across the

  17. Nuclear Energy Research Brookhaven National

    E-Print Network [OSTI]

    Ohta, Shigemi

    & Screening of Fuel Cycle Options Advanced Fuel Development · Thorium Fuel Cycles · Silicon Carbide - 1996* Advanced Nuclear Fuels* Materials in Radiation Environments* * Continuing program within NS Nuclear Safety Advanced Nuclear Systems · Radiation Resistant Materials · Accident Tolerant Fuels

  18. Is Nuclear Energy the Solution?

    E-Print Network [OSTI]

    Saier, Milton H.; Trevors, Jack T.

    2010-01-01

    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

  19. NUCLEAR CHEMISTRY ANNUAL REPORT 1970

    E-Print Network [OSTI]

    Authors, Various

    2011-01-01

    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 (

  20. Full Core, Heterogeneous, Time Dependent Neutron Transport Calculations with the 3D Code DeCART

    E-Print Network [OSTI]

    Hursin, Mathieu

    2010-01-01

    3-D Full Core Boiling Water Reactor Applications. NuclearPWR) rod-ejection and Boiling Water Reactor (BWR) rod-drop

  1. World nuclear outlook 1995

    SciTech Connect (OSTI)

    1995-09-29

    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.

  2. World nuclear outlook 1994

    SciTech Connect (OSTI)

    NONE

    1994-12-01

    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.

  3. dnn | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration (DOENNSA) and the U.K. Department of Energy and Climate Change concluded a workshop at Wilton Park, Shaping the future of nuclear...

  4. testmenu | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

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

  5. Iowa Nuclear Profile - Power Plants

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

    Iowa nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

  6. 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...

  7. 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...

  8. 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...

  9. 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...

  10. Missouri 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...

  11. 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...

  12. 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...

  13. 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...

  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. 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...

  16. 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...

  17. 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...

  18. Massachusetts 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. 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...

  20. Vermont 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. Kansas Nuclear Profile - Power Plants

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

    Kansas nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

  2. 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...

  3. 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...

  4. 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...

  5. China's Nuclear Industry After Fukushima

    E-Print Network [OSTI]

    YUAN, Jingdong

    2013-01-01

    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

  6. NUCLEAR CHEMISTRY ANNUAL REPORT 1970

    E-Print Network [OSTI]

    Authors, Various

    2011-01-01

    Phys. A162, 552 ):'Present address: Chalk River NuclearLaboratories, AECL, Chalk River, Ontario, Canada. NATODenmark. +Present address: Chalk River Nuclear Laboratories,

  7. taiwan | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

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

  8. hrp | National Nuclear Security Administration

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

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

  9. sliderphotos | National Nuclear Security Administration

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

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

  10. An Outlook on Nuclear Physics

    E-Print Network [OSTI]

    Balantekin, A B

    2013-01-01

    A brief outlook on low-energy nuclear physics is presented. Selected recent developments in nuclear structure theory are highlighted and a few open questions are discussed.

  11. An Outlook on Nuclear Physics

    E-Print Network [OSTI]

    A. B. Balantekin

    2013-01-05

    A brief outlook on low-energy nuclear physics is presented. Selected recent developments in nuclear structure theory are highlighted and a few open questions are discussed.

  12. Facilities & Capabilities | Nuclear Science | ORNL

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

    R&D in radioisotopes and stable isotopes; radioisotope production; radiochemistry; global nuclear security technologies; fusion energy science; nuclear criticality safety; and all...

  13. Trident | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

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

  14. Nike | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

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

  15. Importance of Nuclear DataImportance of Nuclear Data to the Naval Nuclear Propulsion Program

    E-Print Network [OSTI]

    Danon, Yaron

    Importance of Nuclear DataImportance of Nuclear Data to the Naval Nuclear Propulsion Program Don of Evaluated Criticality Safety Benchmark Experiments," NEA/NSC/DOC(95)03, 2009. #12;MC21 Analysis of LEU COMP

  16. Perspectives of Nuclear Physics

    E-Print Network [OSTI]

    Amand Faessler

    2002-12-06

    The organizers of this meeting have asked me to present perspectives of nuclear physics. This means to identify the areas where nuclear physics will be expanding in the next future. In six chapters a short overview of these areas will be given, where I expect that nuclear physics willdevelop quite fast: A. Quantum Chromodynamics and effective field theories in the confinement region; B. Nuclear structure at the limits; C. High energy heavy ion collisions; D. Nuclear astrophysics; E. Neutrino physics; F. Test of physics beyond the standard model by rare processes. After a survey over these six points I will pick out a few topics where I will go more in details. There is no time to give for all six points detailed examples. I shall discuss the following examples of the six topics mentionned above: 1. The perturbative chiral quark model and the nucleon $\\Sigma$-term, 2. VAMPIR (Variation After Mean field Projection In Realistic model spaces and with realistic forces) as an example of the nuclear structure renaissance, 3. Measurement of important astrophysical nuclear reactions in the Gamow peak, 4. The solar neutrino problem. As examples for testing new physics beyond the standard model by rare processes I had prepared to speak about the measurement of the electric neutron dipole moment and of the neutrinoless double beta decay. But the time is limited and so I have to skip these points, although they are extremely interesting.

  17. Evaluated Nuclear Data

    SciTech Connect (OSTI)

    Oblozinsky, P.; Oblozinsky,P.; Herman,M.; Mughabghab,S.F.

    2010-10-01

    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.

  18. Nuclear forensics: Soil content

    SciTech Connect (OSTI)

    Beebe, Merilyn Amy

    2015-08-31

    Nuclear Forensics is a growing field that is concerned with all stages of the process of creating and detonating a nuclear weapon. The main goal is to prevent nuclear attack by locating and securing nuclear material before it can be used in an aggressive manner. This stage of the process is mostly paperwork; laws, regulations, treaties, and declarations made by individual countries or by the UN Security Council. There is some preliminary leg work done in the form of field testing detection equipment and tracking down orphan materials; however, none of these have yielded any spectacular or useful results. In the event of a nuclear attack, the first step is to analyze the post detonation debris to aid in the identification of the responsible party. This aspect of the nuclear forensics process, while reactive in nature, is more scientific. A rock sample taken from the detonation site can be dissolved into liquid form and analyzed to determine its chemical composition. The chemical analysis of spent nuclear material can provide valuable information if properly processed and analyzed. In order to accurately evaluate the results, scientists require information on the natural occurring elements in the detonation zone. From this information, scientists can determine what percentage of the element originated in the bomb itself rather than the environment. To this end, element concentrations in soils from sixty-nine different cities are given, along with activity concentrations for uranium, thorium, potassium, and radium in various building materials. These data are used in the analysis program Python.

  19. The Decline and Death of Nuclear Power

    E-Print Network [OSTI]

    Melville, Jonathan

    2013-01-01

    Nuclear- Wastes/Appendices/Radioactive-Waste-Management-Nuclear Association (2010). Radioactive Waste Management:

  20. Virtual nuclear weapons

    SciTech Connect (OSTI)

    Pilat, J.F.

    1997-08-01

    The term virtual nuclear weapons proliferation and arsenals, as opposed to actual weapons and arsenals, has entered in recent years the American lexicon of nuclear strategy, arms control, and nonproliferation. While the term seems to have an intuitive appeal, largely due to its cyberspace imagery, its current use is still vague and loose. The author believes, however, that if the term is clearly delineated, it might offer a promising approach to conceptualizing certain current problems of proliferation. The first use is in a reference to an old problem that has resurfaced recently: the problem of growing availability of weapon-usable nuclear materials in civilian nuclear programs along with materials made `excess` to defense needs by current arms reduction and dismantlement. It is argued that the availability of these vast materials, either by declared nuclear-weapon states or by technologically advanced nonweapon states, makes it possible for those states to rapidly assemble and deploy nuclear weapons. The second use has quite a different set of connotations. It is derived conceptually from the imagery of computer-generated reality. In this use, one thinks of virtual proliferation and arsenals not in terms of the physical hardware required to make the bomb but rather in terms of the knowledge/experience required to design, assemble, and deploy the arsenal. Virtual weapons are a physics reality and cannot be ignored in a world where knowledge, experience, materials, and other requirements to make nuclear weapons are widespread, and where dramatic army reductions and, in some cases, disarmament are realities. These concepts are useful in defining a continuum of virtual capabilities, ranging from those at the low end that derive from general technology diffusion and the existence of nuclear energy programs to those at the high end that involve conscious decisions to develop or maintain militarily significant nuclear-weapon capabilities.