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1

2010 Inspection and Status Report for the Boiling Nuclear Superheater...  

Office of Legacy Management (LM)

3 Annual Inspection - Boiling Nuclear Superheat (BONUS) Site, Rincn, Puerto Rico October 2013 Page 1 2013 Inspection and Status Report for the Former Boiling Nuclear Superheater...

2

BOILING NUCLEAR SUPERHEATER (BONUS) POWER STATION. Final Summary Design Report  

SciTech Connect

The design and construction of the Boiling Nuclear Superheater (BONUS) Power Station at Punta Higuera on the seacoast at the westernmost tip of Puerto Rico are described. The reactor has an output of 17.5 Mw(e). This report will serve as a source of information for personnel engaged in management, evaluation, and training. (N.W.R.)

1962-05-01T23:59:59.000Z

3

BOILING NUCLEAR SUPERHEATER (BONUS) POWER STATION. Supplementary Study. Extrapolation to Large Central Station Integral Nuclear Superheat Plant  

SciTech Connect

An evaluation was made of the maximum size plant for which the BONUS reactor plant could serve as a realistic prototype and the design changes required to increase the size and characteristics for the present BONUS design such that it could serve as a realistic prototype for the largest feasible integral-superheat reactor power plant. (M.C.G.)

1962-10-31T23:59:59.000Z

4

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

Science Conference Proceedings (OSTI)

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

N /A

2003-02-24T23:59:59.000Z

5

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

394: Public Access to the BONUS Facility January 2003 394: Public Access to the BONUS Facility January 2003 i DOE/EA-1394 ENVIRONMENTAL ASSESSMENT FOR AUTHORIZING THE PUERTO RICO ELECTRIC POWER AUTHORITY (PREPA) TO ALLOW PUBLIC ACCESS TO THE BOILING NUCLEAR SUPERHEAT (BONUS) REACTOR BUILDING, RINCÓN, PUERTO RICO January 2003 U.S. Department of Energy Oak Ridge Operations Office Oak Ridge, Tennessee DOE/EA-1394: Public Access to the BONUS Facility January 2003 ii TABLE OF CONTENTS LIST OF FIGURES V LIST OF TABLES V ACRONYMS VI UNIT ABBREVIATIONS VII SUMMARY VIII 1. INTRODUCTION 10 1.1 Purpose and Need for Action 10 1.2 Operational and Decommissioning History 15 1.3 Summary of Radiological Conditions at the BONUS Facility 19 2. DESCRIPTION OF THE PROPOSED ACTION AND ALTERNATIVES 25

6

FEASIBILITY STUDY FOR THE CONCEPTUAL DESIGN OF ADUAL-CORE BOILING SUPERHEAT REACTOR.  

E-Print Network (OSTI)

??For research concerning economical applications of high temperature reactortechnology, a novel approach for creating a Boiling Superheat Reactor (BSR) byaugmenting an Advanced Boiling Water Reactor… (more)

Ross, Jacob

2009-01-01T23:59:59.000Z

7

Conceptual design of an annular-fueled superheat boiling water reactor  

E-Print Network (OSTI)

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 ...

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

2011-01-01T23:59:59.000Z

8

Boiling of nuclear liquid in core-collapse supernova explosions  

E-Print Network (OSTI)

We investigate the possibility of boiling instability of nuclear liquid in the inner core of the proto-neutron star formed in the core collapse of a type II supernova. We derive a simple criterion for boiling to occur. Using this criterion for one of best described equations of state of supernova matter, we find that boiling is quite possible under the conditions realized inside the proto-neutron star. We discuss consequences of this process such as the increase of heat transfer rate and pressure in the boiling region. We expect that taking this effect into account in the conventional neutrino-driven delayed-shock mechanism of type II supernova explosions can increase the explosion energy and reduce the mass of the neutron-star remnant.

Peter Fomin; Dmytro Iakubovskyi; Yuri Shtanov

2007-08-31T23:59:59.000Z

9

Boiling of nuclear liquid in core-collapse supernova explosions  

E-Print Network (OSTI)

We investigate the possibility of boiling instability of nuclear liquid in the inner core of the proto-neutron star formed in the core collapse of a type II supernova. We derive a simple criterion for boiling to occur. Using this criterion for one of best described equations of state of supernova matter, we find that boiling is quite possible under the conditions realized inside the proto-neutron star. We discuss consequences of this process such as the increase of heat transfer rate and pressure in the boiling region. We expect that taking this effect into account in the conventional neutrino-driven delayed-shock mechanism of type II supernova explosions can increase the explosion energy and reduce the mass of the neutron-star remnant.

Fomin, Peter; Shtanov, Yuri

2007-01-01T23:59:59.000Z

10

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1962-08-14T23:59:59.000Z

11

THE DETECTION OF BOILING IN A WATER-COOLED NUCLEAR REACTOR  

SciTech Connect

Measurements made at ORNL to study the feasibility of boiling detection in a water-cooled nuclear reactor are described. The methods selected for the detection of boiling include measurement of the acoustical noise produced by the generation of bubbles and measurement of changes in the reactor-power spectral density produced by bubbles. Preliminary results indicating that both methods could detect boiling are shown. (auth)

Colomb, A.L.; Binford, F.T.

1962-08-17T23:59:59.000Z

12

The role of surface conditions in nucleate boiling  

E-Print Network (OSTI)

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 ...

Griffith, P.

1958-01-01T23:59:59.000Z

13

SELF-REGULATING BOILING-WATER NUCLEAR REACTORS  

DOE Patents (OSTI)

A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

Ransohoff, J.A.; Plawchan, J.D.

1960-08-16T23:59:59.000Z

14

CHLORIDE DEPOSITION FROM STEAM ONTO SUPERHEATER FUEL CLAD MATERIALS  

SciTech Connect

Experimemts using Cl/sup 36/ in a steam test loop were conducted to study the deposition behavior of chlorides on BONUS superheater fuel assembly materials. The moisture content of the steam was varied between 0 and 0.5 wt%, and superheat was added up to 15 deg F before the steam passed over the test cartridge heater. The effects of vaiiables on the chloride deposition on the heater were studied in detail. Chloride deposition from moist steam was found to result in heavy, adherent deposits which are conducive to severe chloride stress corrosion of austenitic steels, while removal of all moisture from the incoming steam reduces the chloride deposition and minimizes the chloride stress corrosion. The heater surface condition was found to be a very important variable; deposition is increased by surface defects and pits. Neither the temperature of steam or heater nor the amount of superheat had an appreciable effect on the deposition, when no moisture existed in the steam. However, low steam velocities and spacer protoberances increase the deposition. Different clad materials (Inconel and Type 304 and 347 stainless steel) with similar surface conditions did not affect the deposition, although subsequent corrosion effects do modify the deposition behavior. Recommendations are given for the control of chloride deposition in nuclear superheater reactor systems. (D.L.C.)

Bevilacqua, F.; Brown, G.M.

1963-10-18T23:59:59.000Z

15

Alternate Materials for Recovery Boiler Superheater Tubes  

SciTech Connect

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.

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

2009-01-01T23:59:59.000Z

16

Assembly fixture for cross-shaped control rods of boiling water nuclear reactors  

Science Conference Proceedings (OSTI)

An assembly fixture is disclosed for cross-shaped control rods of boiling-water nuclear reactors with an upper core grid mesh for holding a core cell formed of four fuel assemblies having a gap therebetween and means disposed beneath the reactor core for driving the control rods in the gap, including a frame having corners formed therein, the frame being substantially the size of a core cell and being disposable on the core grid, templates diagonally oppositely disposed on the frame and extending into the core cell for lateral guidance of the control rods, stops for the control rods disposed on the templates, and a carrying handle having a first portion thereof being pivotable at one of the corners of the frame and a second portion thereof being locked to an opposite corner of the frame in a disassembled condition and swung out of the locked condition in an assembled condition.

Lippert, H.J.

1983-10-18T23:59:59.000Z

17

Superheater Corrosion Produced By Biomass Fuels  

Science Conference Proceedings (OSTI)

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

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

2012-01-01T23:59:59.000Z

18

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

SciTech Connect

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.

NONE

1995-08-01T23:59:59.000Z

19

Recovery Boiler Superheater Ash Corrosion Field Study  

SciTech Connect

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.

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

2010-01-01T23:59:59.000Z

20

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

SciTech Connect

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.

Anh Bui; Nam Dinh; Brian Williams

2013-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

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

SciTech Connect

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).

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

1982-03-01T23:59:59.000Z

22

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

SciTech Connect

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.

Liu, C.; Theofanous, T. G.

2000-08-29T23:59:59.000Z

23

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

DOE Green Energy (OSTI)

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

Sharp, William (Sandy) [SharpConsultant

2011-12-01T23:59:59.000Z

24

Nuclear Desalination Complex with VK-300 Boiling-Type Reactor Facility  

E-Print Network (OSTI)

With regard to the global-scale development of desalination technologies and the stable growth demand for them, Russia also takes an active part in the development of these technologies. Two major aspects play a special role here: they are providing the desalination process with power and introducing new materials capable of making the production of fresh water cheaper and of raising the technical reliability of desalination units. In achieving these tasks, the focus is on the most knowledge-intensive issues, to which Russia is capable of making its contribution based both on the experience of developing national nuclear power and the experience of developing, manufacturing and operating desalination units, including the use of nuclear power (the experience of BN-350 in Aktau (formerly Shevchenko), Kazakhstan). In terms of design, the Nuclear Desalination Complex (NDC) with a VK-300 reactor facility is a modification of a nuclear power unit with a VK-300 reactor developed for application at Russian nuclear cogeneration plants. A power unit

B. A. Gabaraev; Yu. N. Kuznetzov; A. A. Romenkov; Yu. A. Mishanina

2004-01-01T23:59:59.000Z

25

Integrated boiler, superheater, and decomposer for sulfuric acid decomposition  

DOE Patents (OSTI)

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.

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

2010-01-12T23:59:59.000Z

26

Superheater Corrosion in Plants Burning High-Chlorine Coals  

Science Conference Proceedings (OSTI)

Corrosion caused by molten alkali sulfates can cause premature failure in superheaters and reheaters of coal-fired boilers. Coals with a high chlorine content are more likely to cause molten sulfate corrosion than those with a low chlorine content. Tests in a boiler burning coal with 0.37% chlorine and 1.3% sulfur show that stainless steels with at least 35% chromium are very corrosion resistant, while steels containing less than 20% chromium have high corrosion rates.

1992-12-01T23:59:59.000Z

27

Small Punch Creep of Service-Exposed SUS 316 HTB Superheater ...  

Science Conference Proceedings (OSTI)

Presentation Title, Small Punch Creep of Service-Exposed SUS 316 HTB Superheater Tubes of Fossil Boilers. Author(s), Maribel Leticia Saucedo-Muñoz,  ...

28

Technology, safety and costs of decommissioning a reference boiling water reactor power station: Comparison of two decommissioning cost estimates developed for the same commercial nuclear reactor power station  

SciTech Connect

This study presents the results of a comparison of a previous decommissioning cost study by Pacific Northwest Laboratory (PNL) and a recent decommissioning cost study of TLG Engineering, Inc., for the same commercial nuclear power reactor station. The purpose of this comparative analysis on the same plant is to determine the reasons why subsequent estimates for similar plants by others were significantly higher in cost and external occupational radiation exposure (ORE) than the PNL study. The primary purpose of the original study by PNL (NUREG/CR-0672) was to provide information on the available technology, the safety considerations, and the probable costs and ORE for the decommissioning of a large boiling water reactor (BWR) power station at the end of its operating life. This information was intended for use as background data and bases in the modification of existing regulations and in the development of new regulations pertaining to decommissioning activities. It was also intended for use by utilities in planning for the decommissioning of their nuclear power stations. The TLG study, initiated in 1987 and completed in 1989, was for the same plant, Washington Public Supply System's Unit 2 (WNP-2), that PNL used as its reference plant in its 1980 decommissioning study. Areas of agreement and disagreement are identified, and reasons for the areas of disagreement are discussed. 31 refs., 3 figs., 22 tabs.

Konzek, G.J.; Smith, R.I. (Pacific Northwest Lab., Richland, WA (USA))

1990-12-01T23:59:59.000Z

29

Genetic-Algorithm-Based Adaptive Control of Superheat Steam Temperature on a Power Plant Boiler  

Science Conference Proceedings (OSTI)

Superheat steam temperature control is critical to the normal and optimal operation of a power plant. Usually, cascade Proportional Integral Derivative (PID) control system is introduced to regulate the superheat temperature with the PID parameters fixed ... Keywords: Genetic Algorithm, Adaptive Control, Recursive Least Squares, Robustness

Yonghong Huang; Xuejun Yang

2008-12-01T23:59:59.000Z

30

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

E-Print Network (OSTI)

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

Li, Mo

31

bonus  

Office of Legacy Management (LM)

decommissioned Boiling Nuclear Superheater decommissioned Boiling Nuclear Superheater (BONUS) reactor, located northwest of Rincón, Puerto Rico, was developed as a prototype nuclear power plant to investigate the technical and economic feasibility of the integral boiling-superheating concept. This small- scale nuclear reactor produced saturated steam in the central portion of the reactor core, superheated it in four surrounding "superheater" sections of the same

32

Boiling Water in Microwave  

NLE Websites -- All DOE Office Websites (Extended Search)

Boiling Water in Microwave A 26-year old man decided to have a cup of coffee. He took a cup of water and put it in the microwave to heat it up (something that he had done numerous...

33

ADVANCED POWER PLANT MODELING WITH APPLICATIONS TO THE ADVANCED BOILING  

E-Print Network (OSTI)

The components of a modern Advanced Boiling Water Reactor (ABWR) nuclear power plant are modeled in this thesis) is a single-cycle, forced circulation, light-water nuclear reactor designed by the General Electric Company better control of the nuclear reaction in the fuel core. 2.1 Modifications to the BWR [1] · The reactor

Mitchell, John E.

34

Hard boiling eggs  

NLE Websites -- All DOE Office Websites (Extended Search)

Hard boiling eggs Hard boiling eggs Name: Sandburg J High Age: N/A Location: N/A Country: N/A Date: N/A Question: We have been studying chemical and physical changes in 6th grade science class and we were wondering whether hard boiling an egg would be a chemical or a physical change? Thanks for a reply! Replies: You decide. Here's what's going on: the proteins in the fresh egg are in the shape of tight little balls. When you boil the egg, these proteins unravel ("denature"), like balls of yarn unraveling into loose skeins. The strands of protein then get all tangled up with one another, so much so that they are locked in place and can no longer move. They also lock into place the other liquid components of the egg, forming all together what's called a "gel" instead of the liquid you started off with. The gel acts like a soft, rubbery solid because of the network of protein strands holding it all together. It's certainly true that when the protein denatures some chemical bonds are broken, but the most important effect is the tangling up process.

35

Boiling Water Reactor Zinc Addition Sourcebook  

Science Conference Proceedings (OSTI)

Boiling water reactors (BWRs) have been injecting zinc into the primary coolant via the feedwater system for over 25 years to control primary system radiation fields. The zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. This evolution included transition from natural zinc oxide to depleted zinc oxide and from active zinc injection skids (pumped systems) to passive injection systems (zinc pellet beds).  Also occurring were various ...

2013-11-15T23:59:59.000Z

36

OXIDE SCALE EXFOLIATION AND REGROWTH IN TP347H SUPERHEATER TUBES  

SciTech Connect

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

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

2012-01-01T23:59:59.000Z

37

A study of electrowetting-assisted boiling  

E-Print Network (OSTI)

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 ...

Bralower, Harrison L. (Harrison Louis)

2011-01-01T23:59:59.000Z

38

HORIZONTAL BOILING REACTOR SYSTEM  

DOE Patents (OSTI)

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

Treshow, M.

1958-11-18T23:59:59.000Z

39

SWR 1000: The Innovative Boiling Water Reactor  

SciTech Connect

Framatome ANP has developed the boiling water reactor SWR 1000 in close cooperation with German nuclear utilities and with support from various European partners. This advanced reactor design marks a new era in the successful tradition of boiling water reactor technology and, with a gross electric output of between 1290 and 1330 MW, is aimed at assuring competitive power generating costs compared to gas- and coal-fired stations. At the same time, the SWR 1000 meets the highest safety standards, including control of a core melt accident these objectives are met by supplementing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. The SWR 1000 fulfills international nuclear regulatory requirements and has been offered to TVO for the fifth nuclear unit in Finland. (authors)

Brettschuh, Werner [Framatome ANP GmbH, Berlinerstrasse 295, 63067 Offenbach (Germany); Hudson, Greg [Framatome ANP Inc., 400 South Tyron Street, Charlotte, NC 28285 (United States)

2004-07-01T23:59:59.000Z

40

Advanced Nuclear Technology: EPRI Materials Management Matrix Project—Toshiba Advanced Boiling Water Reactor Materials Managem ent Table Report, Revision 0  

Science Conference Proceedings (OSTI)

Experience gained through years of operating nuclear plants has shown that materials performance issues can be a significant concern related to economic and safe long-term plant operations. Although concerns remain, industry efforts to address materials performance issues at operating plants have led to several important advances in both the underlying scientific understanding of materials degradation and the implementation of practical mitigation and management technologies. The Electric Power Research...

2010-02-09T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Subcooled flow boiling of fluorocarbons  

E-Print Network (OSTI)

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 ...

Murphy, Richard Walter

1971-01-01T23:59:59.000Z

42

Fatigue Testing of Metallurgically-Bonded EBR-II Superheater Tubes  

Science Conference Proceedings (OSTI)

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.

Terry C. Totemeier

2006-12-01T23:59:59.000Z

43

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

E-Print Network (OSTI)

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 ...

Dawson, Phillip Eng

2007-01-01T23:59:59.000Z

44

Validation of IVA Computer Code for Flow Boiling Stability Analysis  

SciTech Connect

IVA is a computer code for modeling of transient multiphase, multi-component, non-equilibrium flows in arbitrary geometry including flow boiling in 3D nuclear reactors. This work presents part of the verification procedure of the code. We analyze the stability of flow boiling in natural circulation loop. Experimental results collected on the AREVA/FANP KATHY loop regarding frequencies, mass flows and decay ratio of the oscillations are used for comparison. The comparison demonstrates the capability of the code to successfully simulate such class of processes. (author)

Ivanov Kolev, Nikolay [Framatome-ANP, PO Box 3220, D-91058, Erlangen (Germany)

2006-07-01T23:59:59.000Z

45

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

SciTech Connect

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.

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

2013-01-15T23:59:59.000Z

46

Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2 field testing  

SciTech Connect

In Phase 1 of this project, a variety of developmental and commercial tubing alloys and claddings was exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347, RA85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 modified, NF 709, 690 clad, and 671 clad for over 10,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy are being exposed for 4,000, 12,000, and 16,000 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after approximately 4,400 hours of exposure.

Blough, J.L. [Foster Wheeler Development Corp., Livingston, NJ (United States)

1996-08-01T23:59:59.000Z

47

Modeling acid-gas generation from boiling chloride brines  

Science Conference Proceedings (OSTI)

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).

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

2009-11-16T23:59:59.000Z

48

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

E-Print Network (OSTI)

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 of 10 degrees Celsius and 20 degrees Celsius using PF-5060 as the test fluid and at atmospheric pressure. High-speed digital image acquisition techniques were used to perform hydrodynamic measurements. Boiling curves obtained from the experiments showed that the CNT coating enhanced critical heat flux (CHF) by 63% at 10 degrees Celsius subcooling. The CHF condition was not measured for the CNT sample at 20 degrees Celsius subcooling. Boiling incipience superheat for the CNT-coated surface is shown to be much lower than predicted by Hsu's hypothesis. It is proposed that bubble nucleation occurs within irregularities at the surface of the CNT coating. The irregularities could provide larger cavities than are available between individual nanotubes of the CNT coating. Measurements from high-speed imaging showed that the average bubble departing from the CNT coating in the nucleate boiling regime (excluding the much larger bubbles observed near CHF) was about 75% smaller (0.26 mm versus 1.01 mm)and had a departure frequency that was about 70% higher (50.46 Hz versus 30.10 Hz). The reduction in departure diameter is explained as a change in the configuration of the contact line, although further study is required. The increase in frequency is a consequence of the smaller bubbles, which require less time to grow. It is suggested that nucleation site density for the CNT coating must drastically increase to compensate for the smaller departure diameters if the rate of vapor creation is similar to or greater than that of a bare silicon surface.

Glenn, Stephen T.

2009-08-01T23:59:59.000Z

49

Fuel Reliability Project: Boiling Water Fuel Performance at Kernkraftwerk Leibstadt  

Science Conference Proceedings (OSTI)

The Kernkraftwerk Leibstadt (KKL) boiling water reactor (BWR), a General Electric BWR/6, performed a lead use assembly (LUA) program with fuel from three fuel suppliers. This program presented a unique opportunity to evaluate fuel performance on advanced 10x10 designs of AREVA, Global Nuclear Fuel (GNF), and Westinghouse Electric Company (Westinghouse). Fuel assemblies from each supplier (vendor) were loaded into the KKL core in 1997 and 1998. A number of fuel inspections have been performed during annua...

2007-05-16T23:59:59.000Z

50

Superheater/intermediate temperature airheater tube corrosion tests in the MHD Coal Fired Flow Facility (Eastern Coal Phase)  

DOE Green Energy (OSTI)

Corrosion data have been obtained for tub is exposed for 1500--2000 hours in a proof-of-concept magnetohydrodynamics (MHD) power generation test facility to conditions representative of superheater and intermediate temperature air heater (ITAH) components. The tubes, coated with K{sub 2}SO{sub 4}-rich deposits, were corroded more than in most pulverized coal fired superheater service, but much less than the highly aggressive liquid phase attack encountered in conventional plants with certain coals and temperatures. Results indicated that, with parabolic corrosion kinetics, type 310 and 253MA stainless steels should be usable to 1400F at hot end of ITAH. At final superheater temperatures, 2.25 and 5 Cr steels were indicated to have parabolic corrosion rates generally below a 0.5 mm/yr criterion, based on corrosion scale thickness. However, unknown amounts of scale loss from spallation made this determination uncertain. Stainless steels 304H, 316H, and 321H had parabolic rates variably above the criterion, but may be servicable under less cyclic conditions. Corrosion rates derived from scale thickness and intergranular corrosion depth measurements are reported, along with scale morphologies and compositions. Implications of results on commercial MHD utilization of the alloys are discussed, as well as the indicated need for more corrosion resistant alloys or coatings under the most severe exposure conditions.

White, M.K.

1993-11-01T23:59:59.000Z

51

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

DOE Green Energy (OSTI)

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.

Janes, J.

1984-06-01T23:59:59.000Z

52

CHIMNEY FOR BOILING WATER REACTOR  

DOE Patents (OSTI)

A boiling-water reactor is described which has vertical fuel-containing channels for forming steam from water. Risers above the channels increase the head of water radially outward, whereby water is moved upward through the channels with greater force. The risers are concentric and the radial width of the space between them is somewhat small. There is a relatively low rate of flow of water up through the radially outer fuel-containing channels, with which the space between the risers is in communication. (AE C)

Petrick, M.

1961-08-01T23:59:59.000Z

53

Microsoft Word - TR07-27.doc  

Office of Legacy Management (LM)

Boiling Nuclear Superheat (BONUS), Site, Rincón, Puerto Rico Boiling Nuclear Superheat (BONUS), Site, Rincón, Puerto Rico July 2010 Page 1 2010 Inspection and Status Report for the Former Boiling Nuclear Superheater (BONUS) Reactor Facility, Rincón, Puerto Rico Summary The Former Boiling Nuclear Superheater (BONUS) Reactor Facility, located on the west coast of Puerto Rico in the town of Rincón, was inspected on June 24, 2010. During the inspection radiation technicians from the Idaho National Laboratory (INL) safely packaged and shipped two legacy radioactive sources to INL for disposition. The BONUS facility consists of the containment building, which houses the entombed reactor system, and outside support facilities. The Puerto Rico Electric Power Authority (PREPA) uses the decommissioned BONUS facility as a history museum. It is opened to the public for

54

STEAM-SIDE OXIDE SCALE EXFOLIATION BEHAVIOR IN SUPERHEATERS AND REHEATERS  

SciTech Connect

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.

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

2011-01-01T23:59:59.000Z

55

Acoustically Enhanced Boiling Heat Transfer  

E-Print Network (OSTI)

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.

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

2008-01-07T23:59:59.000Z

56

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

DOE Green Energy (OSTI)

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.

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

2010-09-01T23:59:59.000Z

57

Numerical Simulations of Boiling Jet Impingement Cooling in Power Electronics  

DOE Green Energy (OSTI)

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

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

2006-12-01T23:59:59.000Z

58

Pool boiling heat transfer characteristics of nanofluids  

E-Print Network (OSTI)

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 ...

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

2007-01-01T23:59:59.000Z

59

Nucleate boiling bubble growth and departure  

E-Print Network (OSTI)

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

Staniszewski, Bogumil E.

1959-01-01T23:59:59.000Z

60

Performance Evaluation of Advanced LLW Liquid Processing Technology: Boiling Water Reactor Liquid Processing  

Science Conference Proceedings (OSTI)

This report provides condensed information on boiling water reactor (BWR) membrane based liquid radwaste processing systems. The report presents specific details of the technology, including design, configuration, and performance. This information provides nuclear plant personnel with data useful in evaluating the merits of applying advanced processes at their plant.

2001-11-26T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

SUPERHEATING IN A BOILING WATER REACTOR  

DOE Patents (OSTI)

A boiling-water reactor is described in which the steam developed in the reactor is superheated in the reactor. This is accomplished by providing means for separating the steam from the water and passing the steam over a surface of the fissionable material which is not in contact with the water. Specifically water is boiled on the outside of tubular fuel elements and the steam is superheated on the inside of the fuel elements.

Treshow, M.

1960-05-31T23:59:59.000Z

62

Life of Plant Activity Estimates for a Nominal 1000 MWe Pressurized Water Reactor and Boiling Water Reactor  

Science Conference Proceedings (OSTI)

Decommissioning nuclear power plant and disposal site managers must understand the radioactive source term of a nuclear power plant to effectively manage disposition of these materials. This study estimates the radioactive source term from nominal 1000 MWe pressurized water and boiling water reactors to support decisions related to radioactive waste storage, processing, and disposal through decommissioning.BackgroundThis study examines the radionuclide ...

2012-12-05T23:59:59.000Z

63

Evaluation of the behavior of shrouded plasma spray coatings in the platen superheater of coal-fired boilers  

SciTech Connect

Nickel- and cobalt-based coatings were formulated by a shrouded plasma spray process on boiler tube steels, namely, ASTM-SA210-grade A1 (GrA1), ASTM-SA213-T-11 (T11), and ASTM-SA213-T-22 (T22). The Ni-22Cr-10A1-1Y alloy powder was sprayed as a bond in each case before the final coating. The degradation behavior of the bared and coated steels was studied in the platen superheater of the coal-fired boiler. The samples were inserted through the soot blower dummy points with the help of stainless steel wires. The coatings were found to be effective in increasing resistance to degradation in the given boiler environment. The maximum protection was observed in the case of Stellite-6 (St-6) coating.

Sidhu, B.S.; Prakash, S. [GZS College of Engineering & Technology, Bathinda (India). Dept. of Mechanical Engineering

2006-06-15T23:59:59.000Z

64

Condensate Polishing Guidelines for Pressurized Water Reactor and Boiling Water Reactor Plants - 2004 Revision  

Science Conference Proceedings (OSTI)

Successful condensate polishing allows more reliable operation of nuclear units by maintaining control of ionic and particulate impurity transport to the pressurized water reactor (PWR) steam generators and the boiling water reactor (BWR) and recirculation system. This report presents revisions of EPRI's 1997 nuclear industry consensus guidelines for the design and operation of deep bed and filter demineralizer condensate polishers. These guidelines are consistent with the 2000 revisions of EPRI's "BWR W...

2004-03-16T23:59:59.000Z

65

NUCLEAR REACTOR  

DOE Patents (OSTI)

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

Treshow, M.

1961-09-01T23:59:59.000Z

66

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)

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 of the vessel's 2 inch bottom drain line while operating at full power. The accident was compounded by disabling the plant's isolation condenser system and as an initial condition, the loss of site power. The ability of the plant's passive safety systems to respond to this type of accident, and the code's ability to accurately predict the accidents phenomena was investigated. The overall conclusion was that the modeled plant maintained all relevant safety parameters within specifications supplied by General Electric (GE) in their Standard Safety Analysis Report (SAR) for the term of investigation (I 5,500 real time seconds). While no safety related parameters were exceeded, certain trends appearing near the end of the calculation suggest the need for further investigation. Both containment temperature and pressure were increasing when the transient was terminated. The RELAP5 code was able to simulate a representative model of the plant. Calculated steady state parameters for power, flow rates, recirculation ratio, and mass balance were within I% of those specified in the SAR. However the ability of the code to accurately model low flow, condensation heat transfer, in the presence of noncondensable gases should be verified. It is concluded that the simulation's results seem to pass an intuitive engineering inspection. That is to say, flow and heat transfer data calculated by the RELAP5 code reflect expected values and relational interactions are maintained, but that no quantitative significance could be justified. The uniqueness of the plant's design and the interactive nature of the transient, suggest Additional experimental data from test facilities is needed to validate the calculations.

Faust, Christophor Randall

1995-01-01T23:59:59.000Z

67

Boiling Springs Geothermal Area | Open Energy Information  

Open Energy Info (EERE)

Boiling Springs Geothermal Area Boiling Springs Geothermal Area Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Geothermal Resource Area: Boiling Springs Geothermal Area Contents 1 Area Overview 2 History and Infrastructure 3 Regulatory and Environmental Issues 4 Exploration History 5 Well Field Description 6 Geology of the Area 7 Geofluid Geochemistry 8 NEPA-Related Analyses (0) 9 Exploration Activities (0) 10 References Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"TERRAIN","zoom":6,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"500px","height":"300px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":44.3641,"lon":-115.856,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

68

Boiling characteristics of small multitube bundles  

SciTech Connect

Boiling characteristics of multitube bundles have been investigated experimentally. Small bundles of up to nine rows were used. Void fraction profiles in the test vessel, tube surface temperatures, power input to individual tubes, and critical heat fluxes were measured for different bundle arrangements and boiling conditions. The data were used to study the system hydrodynamics, bundle heat transfer coefficients, and bundle critical heat flux. The data showed that for lower heat fluxes, the heat transfer characteristics are affected by the system hydrodynamics resulting in higher heat transfer coefficients, whereas at higher heat fluxes nucleate boiling is the dominant mechanism. The data also showed that within a tube bundle, the vapor rising from lower tubes enhances the CHF characteristics of the upper tubes.

Chan, A.M.C. (Ontario Hydro Research Div., Toronto (Canada)); Shoukri, M. (McMaster Univ., Hamilton, Ontario (Canada))

1987-08-01T23:59:59.000Z

69

Pool boiling on nano-finned surfaces  

E-Print Network (OSTI)

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 experiments are conducted for saturation and subcooled conditions. Three different types of ordered nano-structured surfaces are fabricated using Step and flash imprint lithography on silicon substrates followed by Reactive Ion Etching (RIE) or Deep Reactive Ion Etching (DRIE). These nano-structures consist of a square array of cylindrical nanofins with a longitudinal pitch of 1 mm, transverse pitch of 0.9 mm and fixed (uniform) heights ranging from 15 nm - 650 nm for each substrate. The contact angle of de-ionized water on the substrates is measured before and after the boiling experiments. The contact-angle is observed to increase with the height of the nano-fins. Contact angle variation is also observed before and after the pool boiling experiments. The pool boiling curves for the nano-structured silicon surfaces are compared with that of atomically smooth single-crystal silicon (bare) surfaces. Data processing is performed to estimate the heat flux through the projected area (plan area) for the nano-patterned zone as well as the heat flux through the total nano-patterned area, which includes the surface area of the fins. Maximum heat flux (MHF) is enhanced by ~120 % for the nanofin surfaces compared to bare (smooth) surfaces, under saturation condition. The pool boiling heat flux data for the three nano-structured surfaces progressively overlap with each other in the vicinity of the MHF condition. Based on the experimental data several micro/nano-scale transport mechanisms responsible for heat flux enhancements are identified, which include: "microlayer" disruption or enhancement, enhancement of active nucleation site density, enlargement of cold spots and enhancement of contact angle which affects the vapor bubble departure frequency.

Sriraman, Sharan Ram

2007-12-01T23:59:59.000Z

70

Pool boiling on nano-finned surfaces  

E-Print Network (OSTI)

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 experiments are conducted for saturation and subcooled conditions. Three different types of ordered nano-structured surfaces are fabricated using Step and flash imprint lithography on silicon substrates followed by Reactive Ion Etching (RIE) or Deep Reactive Ion Etching (DRIE). These nano-structures consist of a square array of cylindrical nanofins with a longitudinal pitch of 1 mm, transverse pitch of 0.9 mm and fixed (uniform) heights ranging from 15 nm – 650 nm for each substrate. The contact angle of de-ionized water on the substrates is measured before and after the boiling experiments. The contact-angle is observed to increase with the height of the nano-fins. Contact angle variation is also observed before and after the pool boiling experiments. The pool boiling curves for the nano-structured silicon surfaces are compared with that of atomically smooth single-crystal silicon (bare) surfaces. Data processing is performed to estimate the heat flux through the projected area (plan area) for the nano-patterned zone as well as the heat flux through the total nano-patterned area, which includes the surface area of the fins. Maximum heat flux (MHF) is enhanced by ~120 % for the nanofin surfaces compared to bare (smooth) surfaces, under saturation condition. The pool boiling heat flux data for the three nano-structured surfaces progressively overlap with each other in the vicinity of the MHF condition. Based on the experimental data several micro/nano-scale transport mechanisms responsible for heat flux enhancements are identified, which include: “microlayer” disruption or enhancement, enhancement of active nucleation site density, enlargement of cold spots and enhancement of contact angle which affects the vapor bubble departure frequency.

Sriraman, Sharan Ram

2007-12-01T23:59:59.000Z

71

CONTINUOUS ANALYZER UTILIZING BOILING POINT DETERMINATION  

DOE Patents (OSTI)

A device is designed for continuously determining the boiling point of a mixture of liquids. The device comprises a distillation chamber for boiling a liquid; outlet conduit means for maintaining the liquid contents of said chamber at a constant level; a reflux condenser mounted above said distillation chamber; means for continuously introducing an incoming liquid sample into said reflux condenser and into intimate contact with vapors refluxing within said condenser; and means for measuring the temperature of the liquid flowing through said distillation chamber. (AEC)

Pappas, W.S.

1963-03-19T23:59:59.000Z

72

Nucleate boiling pressure drop in an annulus: Book 5  

Science Conference Proceedings (OSTI)

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 key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. Nineteen test series and a total of 178 tests were performed. Testing addressed the effects of: Heat flux; pressure; helium gas; power tilt; ribs; asymmetric heat flux. This document consists solely of the plato file index from 11/87 to 11/90.

Not Available

1992-11-01T23:59:59.000Z

73

Boiling Water Reactor Sampling Summary: 2012 Update  

Science Conference Proceedings (OSTI)

This report documents boiling water reactor (BWR) sampling practices for key reactor water and feedwater parameters. It includes information on analysis methods, sampling frequencies, and compliance with the recommended sampling frequencies in BWRVIP-190: BWR Vessels and Internals Project, BWR Water Chemistry Guidelines – 2008 Revision (EPRI report 1016579).

2013-03-28T23:59:59.000Z

74

Computations of Explosive Boiling in Microgravity  

Science Conference Proceedings (OSTI)

Dynamics of the explosive growth of a vapor bubble in microgravity is investigated by direct numerical simulation. A front tracking/finite difference technique is used to solve for the velocity and the temperature field in both phases and to account ... Keywords: front tracking, liquid/vapor phase change, microgravity, unstable boiling

Asghar Esmaeeli; Grétar Tryggvason

2003-12-01T23:59:59.000Z

75

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

E-Print Network (OSTI)

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. ...

Iloeje, Onwuamaeze C.

1972-01-01T23:59:59.000Z

76

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

E-Print Network (OSTI)

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 ...

Shai, Isaac

1967-01-01T23:59:59.000Z

77

Italy Nuclear Security Summit: Fact Sheet | National Nuclear...  

National Nuclear Security Administration (NNSA)

mid-1980s, Italy had an ambitious nuclear power research program which included heavy water, boiling water, light water, and fast reactors. In 1979, Italy signed the NPT which...

78

Corrosion Problems in Coal-Fired Boiler Superheater and Reheater Tubes: Steamside Oxidation and Exfoliation--Development of a Chroma te-Conversion Treatment  

Science Conference Proceedings (OSTI)

This report describes a chromate conversion treatment for preventing steam-side scale exfoliation in superheater and reheater tubes. The performance of scaled tubes that were first chemically cleaned by three techniques and then chromate-treated and tested in steam is evaluated. Test results on oxide growth rate reduction, improved scale stability, reduction of exfoliated scale, and compatibility of dissimilar metal welds are presented, and recommendations for further work are made.

1981-04-01T23:59:59.000Z

79

bonus.cdr  

NLE Websites -- All DOE Office Websites (Extended Search)

decommissioned decommissioned Boiling Nuclear Superheater (BONUS) reactor, located northwest of Rincón, Puerto Rico, was developed as a prototype nuclear power plant to investigate the technical and economic feasibil- ity of the integral boiling-superheating concept. This small-scale nuclear reactor produced saturated steam in the central portion of the reactor core, superheated it in four surrounding "superheater" sections of the same core, and then used the superheated steam in a direct loop to drive a turbine generator. It was one of only two boiling-water superheater reactors ever developed in the United States. The reactor was designed to be large enough to evaluate the major features of the integral boiling-superheating concept realistically without the high construction and operating costs associated with a large plant. Construction of the began in 1960 through a

80

BWRVIP-167NP, Rev. 3: Boiling Water Reactor Issue Management Tables  

Science Conference Proceedings (OSTI)

Nuclear utilities continue to face a number of ongoing issues related to degradation of boiling water reactor (BWR) pressure vessels, reactor internals, and American Society of Mechanical Engineers (ASME) Class 1 piping components. These issues have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) gaps and BWR Vessel and Internals Project (BWRVIP) requirements. The BWR Issue Management Tables in the report are living documents that ...

2013-08-23T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

BWRVIP-167NP, Revision 2: BWR Vessel and Internals Project, Boiling Water Reactor Issue Management Tables  

Science Conference Proceedings (OSTI)

Nuclear utilities face numerous ongoing issues related to degradation of boiling water reactor (BWR) pressure vessels, reactor internals, and American Society of Mechanical Engineers (ASME) Class 1 piping components. These issues have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) issues and BWR Vessel and Internals Project (BWRVIP) requirements. The BWR Issue Management Tables (IMTs) in the report are living documents that summarize the st...

2010-08-24T23:59:59.000Z

82

Advanced Light Water Reactor - Boiling Water Reactor Degradation Matrix (ALWR BWR DM), Revision 0  

Science Conference Proceedings (OSTI)

The advanced light water reactor–boiling water reactor degradation matrix (ALWR BWR DM) is an essential piece of the Electric Power Research Institute’s (EPRI’s) Advanced Nuclear Technology (ANT) materials management matrix initiative for advanced LWR designs. The materials management matrix provides a tool to assist the industry in proactive identification and consideration of materials issues as well as mitigation and management opportunities from the design phase, through component fabrication and pla...

2009-08-25T23:59:59.000Z

83

Boiling radial flow in fractures of varying wall porosity  

DOE Green Energy (OSTI)

The focus of this report is the coupling of conductive heat transfer and boiling convective heat transfer, with boiling flow in a rock fracture. A series of experiments observed differences in boiling regimes and behavior, and attempted to quantify a boiling convection coefficient. The experimental study involved boiling radial flow in a simulated fracture, bounded by a variety of materials. Nonporous and impermeable aluminum, highly porous and permeable Berea sandstone, and minimally porous and permeable graywacke from The Geysers geothermal field. On nonporous surfaces, the heat flux was not strongly coupled to injection rate into the fracture. However, for porous surfaces, heat flux, and associated values of excess temperature and a boiling convection coefficient exhibited variation with injection rate. Nucleation was shown to occur not upon the visible surface of porous materials, but a distance below the surface, within the matrix. The depth of boiling was a function of injection rate, thermal power supplied to the fracture, and the porosity and permeability of the rock. Although matrix boiling beyond fracture wall may apply only to a finite radius around the point of injection, higher values of heat flux and a boiling convection coefficient may be realized with boiling in a porous, rather than nonporous surface bounded fracture.

Barnitt, Robb Allan

2000-06-01T23:59:59.000Z

84

Superheater/intermediate temperature air heater tube corrosion tests in the MHD coal fired flow facility (Montana Rosebud POC tests)  

DOE Green Energy (OSTI)

Nineteen alloys have been exposed for approximately 1000 test hours as candidate superheater and intermediate temperature air heater tubes in a U.S. DOE facility dedicated to demonstrating Proof of Concept for the bottoming or heat and seed recovery portion of coal fired magnetohydrodynamic (MHD) electrical power generating plants. Corrosion data have been obtained from a test series utilizing a western United States sub-bituminous coal, Montana Rosebud. The test alloys included a broad range of compositions ranging from carbon steel to austenitic stainless steels to high chromium nickel-base alloys. The tubes, coated with K{sub 2}SO-containing deposits, developed principally, oxide scales by an oxidation/sulfidation mechanism. In addition to being generally porous, these scales were frequently spalled and/or non-compact due to a dispersed form of outward growth by oxide precipitation in the adjacent deposit. Austenitic alloys generally had internal penetration as trans Tranular and/or intergranular oxides and sulfides. While only two of the alloys had damage visible without magnification as a result of the relatively short exposure, there was some concern about Iona-term corrosion performance owing to the relatively poor quality scales formed. Comparison of data from these tests to those from a prior series of tests with Illinois No. 6, a high sulfur bituminous coal, showed less corrosion in the present test series with the lower sulfur coal. Although K{sub 2}SO{sub 4}was the principal corrosive agent as the supplier of sulfur, which acted to degrade alloy surface scales, tying up sulfur as K{sub 2}SO{sub 4} prevented the occurrence of complex alkali iron trisulfates responsible for severe or catastrophic corrosion in conventional power plants with certain coals and metal temperatures.

White, M.

1996-01-01T23:59:59.000Z

85

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

E-Print Network (OSTI)

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 ...

Rohsenow, Warren M.

1951-01-01T23:59:59.000Z

86

PRELIMINARY HAZARD SUMMARY REPORT ON THE BOILING EXPERIMENTAL REACTOR (BER)  

SciTech Connect

A preliminary evaluation of the hazards associated with a 20-Mw boiling reactor for the purpose of determining site requirements is presented. The Boiling Experimental Reactor design, safety features, and performance are given and the surroundings of the site at Argonne National Laboratory are described. (T.R.H.)

West, J.M.; Anderson, C.A.; Dietrich, J.R.; Harrer, J.M.; Jameson, A.S.; Untermyer, S.

1954-05-01T23:59:59.000Z

87

EA-1394: Final Environmental Assessment | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

394: Final Environmental Assessment 394: Final Environmental Assessment EA-1394: Final Environmental Assessment Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear Superheat (BONUS) Reactor Building, Ricon, Puerto Rico This Environmental Assessment (EA) addresses the proposed action by the U.S. Department of Energy (DOE) to authorize the Puerto Rico Electric Power Authority (PREPA) to allow public access to the Boiling Nuclear Superheat (BONUS) reactor building located near Rincón, Puerto Rico for use as a museum. PREPA, the owner of the facility, is proposing development of the facility as a museum. Environmental Assessment for Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear Superheat

88

Conversion of direct process high-boiling residue to monosilanes  

DOE Patents (OSTI)

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.

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

2000-01-01T23:59:59.000Z

89

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

E-Print Network (OSTI)

Comparison of various heat transfer coefficient models inpool boiling In summary, high heat transfer coefficientin boiling heat transfer can be generally explained by the

Lu, Ming-Chang

2010-01-01T23:59:59.000Z

90

Analysis of boiling experiment using inverse modeling  

DOE Green Energy (OSTI)

Numerical predictions of geothermal reservoir behavior strongly depend on the assumed steam-water relative permeabilities, which are difficult and time-consuming to measure in the laboratory. This paper describes the esti- mation of the parameters of the relative per- meability and capillary pressure functions by automatically matching simulation results to data from a transient boiling experiment performed on a Berea sandstone. A sensitivity analysis reveals the strong dependence of the observed system behavior on effects such as heat transfer from the heater to the core, as well as heat losses through the insulation. Parameters of three conceptual models were estimated by inverse modeling. Each calibra- tion yields consistent effective steam perme- abilities, but the shape of the liquid relative permeability remains ambiguous.

Finsterle, S.; Guerrero, M.; Satik, C.

1998-05-01T23:59:59.000Z

91

Enhanced Natural Convection in a Metal Layer Cooled by Boiling Water  

Science Conference Proceedings (OSTI)

An experimental study is performed to investigate the natural convection heat transfer characteristics and the solidification of the molten metal pool concurrently with forced convective boiling of the overlying coolant to simulate a severe accident in a nuclear power plant. The relationship between the Nusselt number (Nu) and the Rayleigh number (Ra) in the molten metal pool region is determined and compared with the correlations in the literature and experimental data with subcooled water. Given the same Ra condition, the present experimental results for Nu of the liquid metal pool with coolant boiling are found to be higher than those predicted by the existing correlations or measured from the experiment with subcooled boiling. To quantify the observed effect of the external cooling on the natural convection heat transfer rate from the molten pool, it is proposed to include an additional dimensionless group characterizing the temperature gradients in the molten pool and in the external coolant region. Starting from the Globe and Dropkin correlation, engineering correlations are developed for the enhancement of heat transfer in the molten metal pool when cooled by an overlying coolant. The new correlations for predicting natural convection heat transfer are applicable to low-Prandtl-number (Pr) materials that are heated from below and solidified by the external coolant above. Results from this study may be used to modify the current model in severe accident analysis codes.

Cho, Jae-Seon [Seoul National University (Korea, Republic of); Suh, Kune Y. [Seoul National University (Korea, Republic of); Chung, Chang-Hyun [Seoul National University (Korea, Republic of); Park, Rae-Joon [Korea Atomic Energy Research Institute (Korea, Republic of); Kim, Sang-Baik [Korea Atomic Energy Research Institute (Korea, Republic of)

2004-12-15T23:59:59.000Z

92

NUCLEAR POWER PLANT  

DOE Patents (OSTI)

A nuclear power plant for use in an airless environment or other environment in which cooling is difficult is described. The power plant includes a boiling mercury reactor, a mercury--vapor turbine in direct cycle therewith, and a radiator for condensing mercury vapor. (AEC)

Carter, J.C.; Armstrong, R.H.; Janicke, M.J.

1963-05-14T23:59:59.000Z

93

Forced-convection, dispersed-flow film boiling  

E-Print Network (OSTI)

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. ...

Hynek, Scott Josef

1969-01-01T23:59:59.000Z

94

Effects of surface parameters on boiling heat transfer phenomena  

E-Print Network (OSTI)

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 ...

Truong, Bao H. (Bao Hoai)

2011-01-01T23:59:59.000Z

95

Dryout droplet distribution and dispersed flow film boiling  

E-Print Network (OSTI)

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

Hill, Wayne S.

1982-01-01T23:59:59.000Z

96

BWRVIP-241: BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Ve ssel Shell Welds and Nozzle Blend Radii  

Science Conference Proceedings (OSTI)

This report documents supplemental analyses for boiling water reactor (BWR) reactor pressure vessel (RPV) recirculation inlet and outlet nozzle-to-shell welds and nozzle inner radii to address limitations imposed by the U.S. Nuclear Regulatory Commission (NRC) regarding the reduction of inspections specified in Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.

2010-10-26T23:59:59.000Z

97

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

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.

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

2006-10-20T23:59:59.000Z

98

Confined boiling rates of liquefied petroleum gas on water  

SciTech Connect

Results of a program to measure the rate of boiling of liquefied petroleum gas (LPG) on water surface and to develop an analytical model to describe the phenomena involved are reported. Primary emphasis was placed on liquid propane or LPG mixtures containing small quantities of ethane or butane or both. A few exploratory tests were, however, made with pure liquid ethane, ethylene, and n-butane. The investigation was conducted to provide quantitative data and analytical models to delineate the rate of vaporization, the spread rate and the degree of fractionation, should an LPG tanker suffer an accident leading to a major spill on water. For propane or LPG spills on water, immediately following the contact, violent boiling commenced. Ice quickly formed; in most cases, ice was even thrown onto the sidewalls of the vessel. In some instances sprays of water/ice and propane were ejected from the calorimeter. Within a few seconds, however, the interaction quieted and the surface was covered by a rough ice sheet. The LPG boiled on the surface of this ice, but large gas bubbles occasionally appeared under the ice shield and were trapped. The boiling rate decreased with time with a concomitant increase in the thickness of the ice shield. In the first second or two, very high boiling heat fluxes were experienced. The mass of LPG lost was approximately half that spilled originally. It is estimated that only 5 to 15% could have been ejected as liquid if the water loss is used as a reference. However, since the water surface is very agitated during this period, it is not possible to obtain reliable quantitative values of the boiling flux. Also, as noted, the mass lost in the very early time period was approximately proportional to the original mass of LPG used. It may be inferred that larger spills lead to more mixing and boiling before the ice shield prevents a direct contact between the LPG and the water.

Reid, R.C.; Smith, K.A.

1978-05-01T23:59:59.000Z

99

Large scale nuclear sensor monitoring and diagnostics by means of an ensemble of regression models based on Evolving Clustering Methods  

E-Print Network (OSTI)

signals measured at a nuclear Boiling Water Reactor (BWR) located in Oskarshamn, Sweden. A total number NLarge scale nuclear sensor monitoring and diagnostics by means of an ensemble of regression models the validation and reconstruction of 792 signals measured at the Swedish boiling water reactor located

100

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

Science Conference Proceedings (OSTI)

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.

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

2008-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Enhanced boiling heat transfer in horizontal test bundles  

Science Conference Proceedings (OSTI)

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

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

1994-08-01T23:59:59.000Z

102

Soap Manufacturing TechnologyChapter 9 Semi-Boiled Soap Production Systems  

Science Conference Proceedings (OSTI)

Soap Manufacturing Technology Chapter 9 Semi-Boiled Soap Production Systems Surfactants and Detergents eChapters Surfactants - Detergents Press Downloadable pdf of\tChapter 9 Semi-Boiled Soap Production Systems fr

103

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

E-Print Network (OSTI)

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 ...

Deane, Charles William

1969-01-01T23:59:59.000Z

104

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

E-Print Network (OSTI)

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. ...

Bergles A. E.

1962-01-01T23:59:59.000Z

105

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

E-Print Network (OSTI)

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 ...

Truong, Bao H.

106

A new approach in signal processing for sodium boiling noise detection by probability density function estimates  

Science Conference Proceedings (OSTI)

The probability density function (pdf) method of noise signal processing has been investigated for its capability and quality in detecting sodium boiling noise. In an attempt to identify proper features of the pdf for sodium boiling noise detection, the segmented areas under the pdf curves have been found sensitive to sodium boiling noise. New approaches have been followed in selecting the feature threshold and achieving the targeted probabilities for false and missed sodium boiling noise detection.

Reddy, C.P.; Singh, O.P.; Vyjayanthi, R.K.; Prabhakar, R.

1988-03-01T23:59:59.000Z

107

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

DOE Patents (OSTI)

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

White, Terry L. (Oak Ridge, TN)

1991-01-01T23:59:59.000Z

108

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

DOE Patents (OSTI)

An acoustic emission based feedback system for controlling the boiling level of a liquid medium in a microwave oven is provided. The acoustic emissions from the medium correlated with surface boiling is used to generate a feedback control signal proportional to the level of boiling of the medium. This signal is applied to a power controller to automatically and continuously vary the power applied to the oven to control the boiling at a selected level. 2 figs.

White, T.L.

1990-05-02T23:59:59.000Z

109

DESIGN STUDY OF SMALL BOILING REACTORS FOR POWER AND HEAT PRODUCTION  

SciTech Connect

A design study has been made of a small "Package" nuclear power plant for the production of electric power and heat in remotely located, inaccessible areas devoid of natural fuels. The design utilizes a horizontal boiling reactor as a steam generator consistent with safe and simple equipment and a minimum building height. A reactor design of 51/2 Mw capacity, with a combined net electric power output of 750 kw and a heat plant output of 4500 kw, was studied in detail. Tertative cost estimates are presented on the basis of this combination. General comparisons have been made between different systems designed for either independent or combined production of 425 kw net electric power and 2500 kw available heat. (auth)

Treshow, M.

1954-11-01T23:59:59.000Z

110

An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory  

SciTech Connect

Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

1989-12-01T23:59:59.000Z

111

Efficiency of a solar collector with internal boiling  

DOE Green Energy (OSTI)

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.

Neeper, D.A.

1986-01-01T23:59:59.000Z

112

Efficiency of a solar collector with internal boiling  

DOE Green Energy (OSTI)

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 week 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.

Neeper, D.A.

1986-06-01T23:59:59.000Z

113

Boiling Water Reactor Shutdown Chemistry and Dose Summary: September 2010  

Science Conference Proceedings (OSTI)

This 2010 update provides an annual report of shutdown radiation dose rates at 46 boiling water reactors (BWRs) that participate in the Electric Power Research Institute's (EPRI's) BWR Chemistry Monitoring and Assessment program and supersedes the BWR Radiation Assessment and Control (BRAC) Summary that was issued twice a year. In addition to BRAC dose rates, the report also includes information on operating and shutdown water chemistry and worker outage dose and contamination.

2010-09-23T23:59:59.000Z

114

Great Boiling Springs Geothermal Area | Open Energy Information  

Open Energy Info (EERE)

Boiling Springs Geothermal Area Boiling Springs Geothermal Area Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Geothermal Resource Area: Great Boiling Springs Geothermal Area Contents 1 Area Overview 2 History and Infrastructure 3 Regulatory and Environmental Issues 4 Exploration History 5 Well Field Description 6 Geology of the Area 7 Geofluid Geochemistry 8 NEPA-Related Analyses (0) 9 Exploration Activities (0) 10 References Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"TERRAIN","zoom":6,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"500px","height":"300px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":40.66166667,"lon":-119.3616667,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

115

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

Science Conference Proceedings (OSTI)

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.

Not Available

1994-07-01T23:59:59.000Z

116

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

SciTech Connect

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.

Not Available

1994-07-01T23:59:59.000Z

117

Study of plutonium disposition using existing GE advanced Boiling Water Reactors  

SciTech Connect

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.

Not Available

1994-06-01T23:59:59.000Z

118

Liquid-vapour phase change : nucleate boiling of pure fluid and nanofluid under different gravity levels.  

E-Print Network (OSTI)

??This research was a step towards the comprehension of the nano-particles interaction with bubbles created during boiling. It was aimed at solving the controversies of… (more)

Diana, Antoine

2014-01-01T23:59:59.000Z

119

The Development of a Non-Equilibrium Dispersed Flow Film Boiling Heat Transfer Modeling Package.  

E-Print Network (OSTI)

??The dispersed flow film boiling (DFFB) heat transfer regime is important to several applications including cryogenics, rocket engines, steam generators, and in the safety analysis… (more)

Meholic, Michael

2011-01-01T23:59:59.000Z

120

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

E-Print Network (OSTI)

??This work examines the effects of gravitational acceleration on the flow boiling process. A test facility focusing on an annular heat exchanger was designed, built… (more)

Westheimer, David Thomas

2012-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

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

E-Print Network (OSTI)

??This dissertation presents a study exploring the limits of phase-change heat transfer with the aim of enhancing critical heat flux (CHF) in pool boiling and… (more)

Lu, Ming-Chang

2010-01-01T23:59:59.000Z

122

Boiling Water Reactor (BWR) Zinc Injection Strategy Evaluation  

Science Conference Proceedings (OSTI)

All U.S. boiling water reactors (BWRs) inject depleted zinc oxide (DZO) into the reactor feedwater for the purpose of suppressing drywell shutdown radiation dose rates. Current guidance in BWRVIP-190: BWR Vessel and Internals Project, BWR Water Chemistry Guidelines2008 Revision (EPRI report 1016579) is to inject sufficient zinc to achieve a Co-60(s)/Zn(s) ratio of Utility-specific goals may encourage even lower Co-60(s)/Zn(s) levels. This may be in part because BWR e...

2010-11-24T23:59:59.000Z

123

Water inventory management in condenser pool of boiling water reactor  

DOE Patents (OSTI)

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.

Gluntz, Douglas M. (San Jose, CA)

1996-01-01T23:59:59.000Z

124

Water inventory management in condenser pool of boiling water reactor  

DOE Patents (OSTI)

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.

Gluntz, D.M.

1996-03-12T23:59:59.000Z

125

Boiling water neutronic reactor incorporating a process inherent safety design  

DOE Patents (OSTI)

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.

Forsberg, Charles W. (Kingston, TN)

1987-01-01T23:59:59.000Z

126

Boiling water neutronic reactor incorporating a process inherent safety design  

DOE Patents (OSTI)

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.

Forsberg, C.W.

1985-02-19T23:59:59.000Z

127

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

SciTech Connect

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.

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

2004-10-06T23:59:59.000Z

128

A METHOD FOR ESTIMATING THE CONFIDENCE IN THE IDENTIFICATION OF NUCLEAR TRANSIENTS BY A BAGGED  

E-Print Network (OSTI)

in the feedwater system of a nuclear Boiling Water Reactor (BWR). The obtained results indicate that the bagging-fuzzy system for fault detection and isolation in nuclear reactors," Advanced Engineering Informatics, 19 (1A METHOD FOR ESTIMATING THE CONFIDENCE IN THE IDENTIFICATION OF NUCLEAR TRANSIENTS BY A BAGGED

129

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

Science Conference Proceedings (OSTI)

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

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

1990-01-01T23:59:59.000Z

130

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

E-Print Network (OSTI)

-flow boiling over circular tube bundles has been meticulously studied; collected data and correlations for circular tube bundles. For exam- ple, Jensen and Hsu [81] conducted a parametric study of boiling heat transfer in a horizontal tube bundle and reported an increase in local heat transfer coefficient

Peles, Yoav

131

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

E-Print Network (OSTI)

in saturated pool boiling. In this model the analysis of heat conduction within the heater is added on the heater surface itself [10]. Bhat et al. [11] put forward a theoretical model of macrolayer formation to their model and ob- tained the simulated boiling curve of water. In addition, they compared Haramura and Katto

Maruyama, Shigeo

132

Technical Basis for Water Chemistry Control of IGSCC in Boiling ...  

Science Conference Proceedings (OSTI)

... Degradation of Materials in Nuclear Power Systems – Water Reactors ... However, even the utilization of near theoretical conductivity water cannot prevent ...

133

BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES  

DOE Patents (OSTI)

This patent covers the use of injection nozzles for pumping water into the lower ends of reactor fuel tubes in which water is converted directly to steam. Pumping water through fuel tubes of this type of boiling water reactor increases its power. The injection nozzles decrease the size of pump needed, because the pump handles only the water going through the nozzles, additional water being sucked into the tubes by the nozzles independently of the pump from the exterior body of water in which the fuel tubes are immersed. The resulting movement of exterior water along the tubes holds down steam formation, and thus maintains the moderator effectiveness, of the exterior body of water. (AEC)

Treshow, M.

1963-04-30T23:59:59.000Z

134

Interfacing systems LOCAs (Loss of Coolant Accidents) at boiling water reactors  

Science Conference Proceedings (OSTI)

The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency (CDF).

Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

1987-01-01T23:59:59.000Z

135

A Numerical Model for Evaluating the Impact of Noble Metal Chemical Addition in Boiling Water Reactors  

SciTech Connect

The technique of noble metal chemical addition (NMCA), accompanied by a low-level hydrogen water chemistry (HWC), is being employed by several U.S. nuclear power plants for mitigating intergranular stress corrosion cracking in the vessel internals of their boiling water reactors (BWRs). An improved computer model by the name of DEMACE was employed to evaluate the performance of NMCA throughout the primary coolant circuit (PCC) of a commercial BWR. The molar ratios of hydrogen to oxidizing species in the PCC under normal water chemistry and HWC are analyzed. The effectiveness of NMCA is justified by calculated electrochemical corrosion potential (ECP) around the PCC and in a local power range monitoring (LPRM) housing tube, in which practical in-vessel ECP measurements are normally taken.Prior to the modeling work for the BWR, the Mixed Potential Model, which is embedded in DEMACE and responsible for ECP calculation, was calibrated against both laboratory and plant ECP data. After modeling for various HWC conditions, it is found that the effectiveness of NMCA in the PCC of the selected BWR varies from region to region. In particular, the predicted ECP in the LPRM housing tube is notably different from that in the nearby bulk environment under NMCA, indicating that cautions must be given to a possible, undesirable outcome due to a distinct ECP difference between a locally confined area and the actual bulk environment.

Yeh, T.-K. [National Tsing-Hua University, Taiwan (China)

2002-10-15T23:59:59.000Z

136

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

SciTech Connect

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.

NONE

1994-04-30T23:59:59.000Z

137

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

E-Print Network (OSTI)

The objective of this investigation is to measure and analyze surface temperature fluctuations in pool boiling. The surface temperature fluctuations were recorded on silicon surfaces with and without multi-walled carbon nanotubes (MWCNT). Novel Thin Film Thermocouples (TFT) are micro-fabricated on test substrates to measure surface temperatures. A dielectric liquid refrigerant (PF-5060) is used as test fluid. Both nucleate and lm boiling regimes are investigated for the silicon test substrates. Dynamics of nucleate boiling is investigated on the CNT coated substrates. High frequency temperature fluctuation data is analyzed for the presence of determinism using non-linear time series analysis techniques in TISEAN(copyright) software. The impact of subcooling and micro/nano-scale surface texturing using MWCNT coatings on the dynamics of pool boiling is assessed. Dynamic invariants such as correlation dimensions and Lyapunov spectrum are evaluated for the reconstructed attractor. A non-linear noise reduction scheme is employed to reduce the level of noise in the data. Previous investigations in pool boiling chaos, reported in literature were based on temperature measurements underneath the test surface consisting of single or few active nucleation sites. Previous studies have indicated the presence of low-dimensional behavior in nucleate boiling and high-dimensional behavior in CHF and film boiling. Currently, there is no study detailing the effects of multiple nucleation sites, subcooling and surface texturing on pool boiling dynamics. The investigation comprises of four parts: i) in situ micro-machining of Chromelalumel (K-type) TFT, ii) calibration of these sensors, iii) utilizing these sensors in pool boiling experiments iv) analysis of these fluctuations using techniques of nonlinear time series analysis. Ten TFT are fabricated on a rectangular silicon surface within an area of ~ 3.00 cm x 3.00 cm. The sensing junctions of the TFT measure 50 mm in width and 250 nm in depth. Surface temperature fluctuations of the order of i) 0.65-0.93 degrees C are observed near ONB ii) 2.3-6.5 degrees C in FDNB iii) 2.60-5.00 degrees C at CHF and iv) 2.3-3.5 degrees C in film boiling. Investigations show the possible presence of chaotic dynamics near CHF and in film-boiling in saturated and subcooled pool boiling. Fully-developed nucleate boiling (FDNB) is chaotic. No clear assessment of the dynamics could be made in the onset of nucleate boiling (ONB) and partial nucleate boiling (PNB) regimes due to the effects of noise. However, the frequency spectra in these regimes appear to have two independent frequencies and their integral combinations indicating a possible quasiperiodic bifurcation route to chaos. The dimensionality in FDNB, at CHF and in film-boiling is lower in saturated pool boiling as compared to values in corresponding regimes in subcooled pool boiling. Surface temperature fluctuations can damage electronic components and need to be carefully controlled. Understanding the nature of these fluctuations will aid in deciding the modeling approach for surface temperature transients on an electronic chip. Subsequently, the TFT signals can be employed in a suitable feedback control loop to prevent the occurrence of hotspots.

Sathyamurthi, Vijaykumar

2009-12-01T23:59:59.000Z

138

DOE - Office of Legacy Management -- Bonus  

Office of Legacy Management (LM)

Puerto Rico Puerto Rico Boiling Nuclear Superheater (BONUS), Puerto Rico, Decommissioned Reactor Site La Antiqua Central Nuclear de Aqua Hirviente Sobrecalentada Reactor Fuera de Servico Activo A D&D Program Site bonus_map3 As part of the DOE Defense Decontamination and Decommissioning (D&D) Program, the Office of Legacy Management manages the Boiling Nuclear Superheater (BONUS) Decommissioned Reactor Site and ensures compliance with applicable federal, state, and local environmental protection laws and regulations, executive orders, and internal DOE policies. The site transferred to the Office of Legacy Management in 2004 and requires routine inspection and maintenance, records-related activities, and stakeholder support. For more information about the BONUS site, view the fact sheet.

139

FUNDAMENTAL INVESTIGATION OF BOILING HEAT TRANSFER AND TWO-PHASE FLOW  

SciTech Connect

Significantly improved theories of two-phase heat transfer and prediction of departure from nucleate boiling have recently been developed which for the first time are not based on empirical relationships. These theories should be critically analyzed in relation to naval reactor work and tested with all existing data from both classified and unclassified sources. Conflicting analyses of two-phase fluid fiow regimes confuse this area, and essentially no data or theories are avsilable for twophase fiow with superimposed boiling. Theories and understanding of two-phase flow with boiling should be developed, starting from proven theories without boiling, and tested against all existing data or new data as necessary. A substantial start hss been made in analysis of the case of upward annular two-phase flow in vertical channels, based upon modern knowledge of boundary layer and vapor condensation principles. (auth)

Grohse, E.W.; Mueller, G.O.; Findlay, J.A.

1958-10-17T23:59:59.000Z

140

Prediction of departure from nucleate boiling in PWR fast power transients  

E-Print Network (OSTI)

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 ...

Lenci, Giancarlo

2013-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

1514 JOURNAL OF MICROELECTROMECHANICAL SYSTEMS, VOL. 15, NO. 6, DECEMBER 2006 Bubble Dynamics During Boiling in  

E-Print Network (OSTI)

. At lower heat fluxes the void fraction increase is insufficient to change the flow pattern to annular, and P. Mercier, "Experimental investigations on boiling of n-pentane across a horizontal tube bundle

Peles, Yoav

142

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

E-Print Network (OSTI)

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 ...

Levin, Alan Edward

1980-01-01T23:59:59.000Z

143

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

E-Print Network (OSTI)

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 ...

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

2010-01-01T23:59:59.000Z

144

Bubble behavior in subcooled flow boiling on surfaces of variable wettability  

E-Print Network (OSTI)

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 ...

Tow, Emily W

2012-01-01T23:59:59.000Z

145

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

E-Print Network (OSTI)

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 ...

Laverty, W. F.

1964-01-01T23:59:59.000Z

146

Boiling-Water Reactor internals aging degradation study. Phase 1  

SciTech Connect

This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.

Luk, K.H. [Oak Ridge National Lab., TN (United States)

1993-09-01T23:59:59.000Z

147

Robust nuclear signal reconstruction by a novel ensemble model aggregation procedure P. Baraldi1  

E-Print Network (OSTI)

of a nuclear boiling water reactor and 215 signals measured at a pressurized water reactor. The advantagesRobust nuclear signal reconstruction by a novel ensemble model aggregation procedure P. Baraldi1 Reactor Project, 1751, Halden, Norway Abstract Monitoring of sensor operation is important for detecting

148

DOE LM/GJ596  

NLE Websites -- All DOE Office Websites (Extended Search)

LM/GJ596 LM/GJ596 2004 - - U.S. Department of Energy Long-Term Surveillance and Maintenance Plan for the Boiling Nuclear Superheater (BONUS) Reactor Facility, Rincón, Puerto Rico May 2005 Work Performed Under DOE Contract No. for the U.S. Department of Energy Office of Legacy Management. DE-AC01-02GJ79491 Approved for public release; distribution is unlimited. Office of Legacy Management Office of Legacy Management Office of Legacy Management Office of Legacy Management S0109100 DOE-LM/GJ596-2004 Long-Term Surveillance and Maintenance Plan for the Boiling Nuclear Superheater (BONUS) Reactor Facility Rincón, Puerto Rico May 2005 Work Performed by S.M. Stoller Corporation under DOE Contract No. DE-AC01-02GJ79491 for the U.S. Department of Energy Office of Legacy Management, Grand Junction, Colorado

149

DOE - Office of Legacy Management -- Bonus  

Office of Legacy Management (LM)

Puerto Rico Puerto Rico Boiling Nuclear Superheater (BONUS), Puerto Rico, Decommissioned Reactor Site This Site All Sites All LM Quick Search Key Documents and Links All documents are Adobe Acrobat files. pdf_icon Key Documents Fact Sheet BONUS, Puerto Rico, Reactor Clausurado, Hoja de Datos Long-Term Surveillance and Maintenance Plan Plan para la vigilance y mantenimiento a largo plaza Please be green. Do not print these documents unless absolutely necessary. Request a paper copy of any document by submitting a Document Request. All Site Documents All documents are Adobe Acrobat files. pdf_icon Fact Sheet Reports Other Documents Español Documents Fact Sheet Boiling Nuclear Superheater, Puerto Rico, Site Fact Sheet August 2012 Reports PREPA-DOE Letter of Agreement July 2010

150

Method of controlling crystallite size in nuclear-reactor fuels  

DOE Patents (OSTI)

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.

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

1985-01-01T23:59:59.000Z

151

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

E-Print Network (OSTI)

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 in three parts. As an introduction, a brief description of the code RAMONA-4B. Later, the implemented SBWR model, based on the General Electric Standard Safety Analysis Report (SSAR), is described and discussed. Finally, the reactor behavior during a turbine trip transient is numerically simulated through the description of nuclear and thermal hydraulic parameters and under the scenario conditions suggested by General Electric. The SBWR model consists of the representation of the vessel internal components through parameters such as areas, diameters and volumes, and the one-quarter-core neutron parameters which were obtained using the transport theory lattice physics code CASMO-3. The thermohydraulic equations are solved by RAMONA-4B in a closed-contour inside the vessel and in a hundred eighty four parallel channels (including bypass) in the core. The tridimensional representation of the reactor core is accomplished through a proposed fuel load which was obtained from a selection of out of three fuel loads and using some standard fuel design parameters. The cross sections are represented using a polynomial as a function of the bumup, void fraction, fuel and moderator temperatures. The six-group delayed neutron equation and the one-and-a-half neutron diffusion equation are solved and the power distribution in the reactor core is obtained.Also, RAMONA-4B has implemented a (adiabatic) steam line model to represent the acoustic effects of the turbine stop valve closure during the transient. Finally, the two-phase coolant and neutronic parameters are calculated in steady state and during the turbine trip transient. The results are discussed and compared against the ones shown in the chapter XV of the SSAR.

Cuevas Vivas, Gabriel Francisco

1995-01-01T23:59:59.000Z

152

CIVILIAN POWER REACTOR PROGRAM. PART II. ECONOMIC POTENTIAL AND DEVELOPMENT PROGRAM AS OF 1959  

SciTech Connect

The status of technology of nuclear power reactors in 1959 is reviewed. General research and engineering development activities are discussed. The reactors considered include the pressurized water, boiling water, light water moderated superheat, organic cooled, sodium graphite, gas cooled enriched fuel, gas cooled natural uranium, fast breeder, aqueous homogeneous, and heavy water. Power costs are compared with the cost of power from conventional plants. (C.H.)

1960-01-01T23:59:59.000Z

153

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

DOE Patents (OSTI)

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.

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

1998-02-24T23:59:59.000Z

154

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

DOE Patents (OSTI)

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.

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

1998-01-01T23:59:59.000Z

155

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

SciTech Connect

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

NONE

1997-05-01T23:59:59.000Z

156

Bench-scale screening tests for a boiling sodium-potassium alloy solar receiver  

DOE Green Energy (OSTI)

Bench-scale tests were carried out in support of the design of a second-generation 75-kW{sub t} reflux pool-boiler solar receiver. The receiver will be made from Haynes Alloy 230 and will contain the sodium-potassium alloy NaK-78. The bench-scale tests used quartz-lamp-heated boilers to screen candidate boiling-stabilization materials and methods at temperatures up to 750{degree}C. Candidates that provided stable boiling were tested for hot-restart behavior. Poor stability was obtained with single 1/4-inch diameter patches of powdered metal hot-press-sintered onto the wetted side of the heat-input area. Laser-drilled and electric-discharge-machined cavities in the heated surface also performed poorly. Small additions of xenon, and heated-surface tilt out of the vertical dramatically improved poor boiling stability; additions of helium or oxygen did not. The most stable boiling was obtained when the entire heat-input area was covered by a powdered-metal coating. The effect of heated-area size was assessed for one coating: at low incident fluxes, when even this coating performed poorly, increasing the heated-area size markedly improved boiling stability. Good hot-restart behavior was not observed with any candidate, although results were significantly better with added xenon in a boiler shortened from 3 to 2 feet. In addition to the screening tests, flash-radiography imaging of metal-vapor bubbles during boiling was attempted. Contrary to the Cole-Rohsenow correlation, these bubble-size estimates did not vary with pressure; instead they were constant, consistent with the only other alkali metal measurements, but about 1/2 their size.

Moreno, J.B.; Moss, T.A.

1993-06-01T23:59:59.000Z

157

Boiling Water Reactor Chemistry Performance Monitoring Update--2007 Edition  

Science Conference Proceedings (OSTI)

Successful operation of a nuclear plant demands careful monitoring of water chemistry, particularly in BWRs, where control of iron and copper in the reactor coolant is essential. Since the advent of hydrogen water chemistry (HWC), plant operators have successfully applied other chemistry regimes such as noble metal chemical addition (NMCA) and zinc injection to control radiation fields and provide additional mitigation for intergranular stress corrosion cracking (IGSCC). This report compiles recent BWR p...

2007-12-12T23:59:59.000Z

158

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

Science Conference Proceedings (OSTI)

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)

Ishii, M.; Xu, Y.; Revankar, S.T. [Purdue University, West Lafayette, IN 47907 (United States)

2002-07-01T23:59:59.000Z

159

Apparatus to measure liquid helium boil-off from low-loss superconducting current leads  

DOE Green Energy (OSTI)

A low-loss liquid helium dewar was constructed to measure the liquid helium boil-off rate from high-temperature superconducting current leads. The dewar has a measured background heat leakage rate of 12 mW. Equations calculating the heat leakage rate from the measured vapor mass flow rate in liquid helium boil-off experiments are derived. Parameters that affect the experiments, such as density ratio, absolute pressure, and rate of pressure variation, are discussed. This study is important as superconducting current leads may be used in superconducting magnetic energy storage systems.

Cha, Y.S.; Niemann, R.C.; Hull, J.R. [Argonne National Lab., IL (United States). Energy Technology Div.

1995-06-01T23:59:59.000Z

160

Analysis and Measurement of Bubble Dynamics and Associated Flow Field in Subcooled Nucleate Boiling Flows  

SciTech Connect

In recent years, subooled nucleate boiling (SNB) has attrcted expanding research interest owing to the emergence of axial offset anomaly (AOA) or crud-induced power shigt (CIPS) in many operating US PWRs, which is an unexpected deviation in the core axial power distribution from the predicted power curves. Research indicates that the formation of the crud, which directly leads to AOA phenomena, results from the presence of the subcooled nucleate boiling, and is especially realted to bubble motion occurring in the core region.

Barclay G. Jones

2008-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Boiling and condensation processes in the Cerro Prieto beta reservoir under exploitation  

DOE Green Energy (OSTI)

The deep Cerro Prieto (Baja California, Mexico) beta reservoir is offset vertically by the southwest-northeast trending, normal H fault. Under exploitation pressures in the upthrown block have decreased strongly resulting in boiling and high-enthalpy production fluids. Significant differences in fluid chemical and isotopic compositions are observed in the two parts of the reservoir and particularly in an anomalous zone associated with the H fault. These differences result from intense boiling and adiabatic steam condensation, as well as from leakage of overlying cooler water along the fault.

Truesdell, A. (Truesdell (Alfred), Menlo Park, CA (United States)); Manon, A.; Quijano, L. (Comision Federal de Electricidad, Morelia (Mexico)); Coplen, T. (Geological Survey, Reston, VA (United States)); Lippmann, M. (Lawrence Berkeley Lab., CA (United States))

1992-01-01T23:59:59.000Z

162

Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review  

SciTech Connect

In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path.

Lund, A.L.

1997-11-01T23:59:59.000Z

163

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

SciTech Connect

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

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

2003-06-16T23:59:59.000Z

164

Identification of Boiling Two-phase Flow Patterns in Water Wall Tube Based on BP Neural Network  

Science Conference Proceedings (OSTI)

In this paper, the boiling phenomena of steam boiler under atmospheric pressure are simulated by using the UDF program of CFD software. Characteristics including pressure, temperature and vapor fraction respectively for bubble, slug and annular flow ... Keywords: Boiling heat transfer, BP neural network, flow pattern, coefficient of heat transfer

Lei Guo; Shusheng Zhang; Yaqun Chen; Lin Cheng

2010-06-01T23:59:59.000Z

165

DEUTERIUM-HYDROGEN EXCHANGE IN BOEHMITE CORROSION PRODUCT FORMED ON PURE ALUMINUM IN BOILING WATER  

SciTech Connect

Proton-deuteron exchange is rapid in boehmite corrosion product formed on pure aluminum in boiling water. In addition, deuterated boehmite films undergo rapid exchange with the humidity of the atmosphere. This explains the previously reported anomaly in the H-D exchange rate for the growing corrosion product on 1100 aluminum. (auh)

Mori, S.; Draley, J.E.; Bernstein, R.B.

1961-10-31T23:59:59.000Z

166

NUMERICAL SIMULATION OF POOL BOILING FOR STEADY STATE AND TRANSIENT HEATING  

E-Print Network (OSTI)

boiling. The developed model includes the analysis of heat conduction within the heater coupled-dimensional transient heat conduction within the heater coupled with the macrolayer model was considered. Being employed-averaged model from experimental measurements of void fraction close to the heater surface. In the model

Maruyama, Shigeo

167

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

E-Print Network (OSTI)

Simulation of subcooled boiling was carried out using RELAP5 thermal hydraulic computer programs. Both one-dimensional and three-dimensional analyses were carried out with one-dimensional RELAP5/MOD3.2 and three-dimensional RELAP5-3D code. Experimental data from the subcooled boiling experiment at low pressure conditions of Bartel, and Zeitoun and Shoukri were simulated. The RELAP5/MOD3.2 was executed to determine the axial void faction distribution. The predictions of void fraction distributions 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 of RELAP5/MOD3.2 were in a good agreement with the experimental data. Two sets of both Bartel's and Zeitoun and Shoukri's experiments were chosen for three-dimensional simulation. Three-dimensional input model resembling the annular test section was constructed. The simulation results using RELAP5-3D program achieved a good agreement with low and high-pressure experimental data. Sensitivity study, with various nodalization schemes, was performed to obtain the optimum simulation parameters.

Reza, S.M. Mohsin

2002-01-01T23:59:59.000Z

168

Advanced Power Plant Modeling with Applications to an Advanced Boiling Water  

E-Print Network (OSTI)

wave fronts. However, in most power plant transient performance models, there are few heat exchangersAdvanced Power Plant Modeling with Applications to an Advanced Boiling Water Reactor and a Heat Introduction This paper presents two advanced modeling methods, and two applications, for power plant

Mitchell, John E.

169

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

E-Print Network (OSTI)

2007-No54-BoilingPoint Theme Health and Greenhouse Gas Impacts of Biomass and Fossil Fuel Energy nations. In sub-Saharan Africa (SSA), biomass provides more than 90% of household energy needs in many nations. The combustion of biomass emits pollutants that currently cause over 1.6 million annual deaths

Kammen, Daniel M.

170

Technical and economic analysis of the thermal performance of a solar boiling concentrator for power generation  

SciTech Connect

A system for power generation using solar energy collected by compound parabolic concentrators (CPC) incorporated into a Rankine cycle system is studied by developing a model to simulate the CPC performance. The power cycle is also modeled under quasi-steady and transient conditions. An economic analysis is performed through a model developed to study the economic viability of the power system. The CPC performance is sensitive to the ratio of diffuse to beam components of the solar incident irradiation. This ratio, along with the concentration ratio, govern the CPC optical efficiency which in turn determine the thermal efficiency. The performance of the CPC working under boiling and superheating conditions is governed by the axial fractional lengths of the non-boiling and the superheating regions. The overall thermal loss coefficient is formulated as a function of the local thermal loss coefficient in the different regions and the length of each region. The thermal efficiency of CPC's and flat plates, whether under non-boiling, boiling or superheating conditions, is evaluated. The CPC working under superheating conditions has a good potential for solar powered Rankine cycles. System efficiencies as high as 11.3% could be obtained at R-11 evaporation temperature of 120/sup 0/C and a condensation temperature of 20/sup 0/ C.

El-Assy, A.Y.

1985-01-01T23:59:59.000Z

171

Nondestructive Evaluation: Boiling Water Reactor Bottom Head Drain Line Examination - Field Trial  

Science Conference Proceedings (OSTI)

This report describes newly developed technology for the examination of the boiling water reactor (BWR) vessel drain line. The technology targets the examination of the elbow and piping section deemed most susceptible to flow-accelerated corrosion (FAC) attack. The technology developed includes a remotely operated sensor manipulator and an ultrasound data acquisition system to perform thickness measurements throughout the affected components.

2007-12-12T23:59:59.000Z

172

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

SciTech Connect

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.

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

1981-04-01T23:59:59.000Z

173

Letter Report: Progress in developing EQ3/6 for modeling boiling processes  

DOE Green Energy (OSTI)

EQ3/6 is a software package for geochemical modeling of aqueous systems, such as water/rock or waste/water rock. It is being developed for a variety of applications in geochemical studies for the Yucca Mountain Site Characterization Project. The present focus is on development of capabilities to be used in studies of geochemical processes which will take place in the near-field environment and the altered zone of the potential repository. We have completed the first year of a planned two-year effort to develop capabilities for modeling boiling processes. These capabilities will interface with other existing and future modeling capabilities to provide a means of integrating the effects of various kinds of geochemical processes in complex systems. This year, the software has been modified to allow the formation of a generalized gas phase in a closed system for which the temperature and pressure are known (but not necessarily constant). The gas phase forms when its formation is thermodynamically favored; that is, when the system pressure is equal to the sum of the partial pressures of the gas species as computed from their equilibrium fugacities. It disappears when this sum falls below that pressure. `Boiling` is the special case in which the gas phase which forms consists mostly of water vapor. The reverse process is then `condensation.` To support calculations of boiling and condensation, we have added a capability to calculate the fugacity coefficients of gas species in the system H{sub 2}O-CO{sub 2}-CH{sub 4}-H{sub 2},-Awe{sub 2}-N{sub 2},-H{sub 2}S-NH3. This capability at present is accurate only at relatively low pressures, but is adequate for all likely repository boiling conditions. We have also modified the software to calculate changes in enthalpy (heat) and volume functions. Next year we will be extending the boiling capability to calculate the pressure or the temperature at known enthalpy. We will also add an option for open system boiling.

Wolery, T. J., LLNL

1995-08-28T23:59:59.000Z

174

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

DOE Patents (OSTI)

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.

Tokarz, R.D.

1981-10-27T23:59:59.000Z

175

Nuclear data for nuclear transmutation  

Science Conference Proceedings (OSTI)

Current status on nuclear data for the study of nuclear transmutation of radioactive wastes is reviewed

Hideo Harada

2009-01-01T23:59:59.000Z

176

Integrated Boiler, Superheater & Decomposer Bayonet for ...  

With the growing pressure placed on energy efficiency and reliance on fossil fuels, alternative sources of energy are increasingly important. The primary function can ...

177

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

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.

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

2011-01-19T23:59:59.000Z

178

Project plan for the decontamination and decommissioning of the Argonne National Laboratory Experimental Boiling Water Reactor  

SciTech Connect

In 1956, the Experimental Boiling Water Reactor (EBWR) Facility was first operated at Argonne National Laboratory (ANL) as a test reactor to demonstrate the feasibility of operating an integrated power plant using a direct cycle boiling water reactor as a heat source. In 1967, ANL permanently shut down the EBWR and placed it in dry lay-up. This project plan presents the schedule and organization for the decontamination and decommissioning of the EBWR Facility which will allow it to be reused by other ANL scientific research programs. The project total estimated cost is $14.3M and is projected to generate 22,000 cubic feet of low-level radioactive waste which will be disposed of at an approved DOE burial ground. 18 figs., 3 tabs.

Boing, L.E.

1989-12-01T23:59:59.000Z

179

Preliminary results of the US pool-boiling coils from the IFSMTF full-array tests  

SciTech Connect

The Large Coil Task to develop superconducting magnets for fusion reactors, is now in the midst of full-array tests in the International Fusion Superconducting Magnet Test Facility at Oak Ridge National Laboratory. Included in the test array are two pool-boiling coils designed and fabricated by US manufacturers, General Dynamics/Convair Division and General Electric/Union Carbide Corporation. So far, both coils have been energized to full design currents in the single-coil tests, and the General Dynamics coil has reached the design point in the first Standard-I full-array test. Both coils performed well in the charging experiments. Extensive heating tests and the heavy instrumentation of these coils have, however, revealed some generic limitations of large pool-boiling superconducting coils. Details of these results and their analyses are reported.

Lue, J.W.; Dresner, L.; Lubell, M.S.; Luton, J.N.; McManamy, T.J.; Shen, S.S.

1986-01-01T23:59:59.000Z

180

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

Science Conference Proceedings (OSTI)

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

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

1980-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Effect of surfactant additive on pool boiling of concentrated lithium bromide solution  

SciTech Connect

The measurements of nucleate pool boiling heat transfer rate and surface tension were made for pure water and 50 wt.% lithium bromide solution with various amounts of n-octanol. Regardless of low concentration, n-octanol additive depresses considerably the surface tension of the liquids. The pool boiling data, however, reveal that the addition of surfactant results in insignificant enhancement of heat transfer for both pure water and the concentrated LiBr solution. With the results of this work, the performance improvement received from using n-octanol additive in working liquid of an absorption heat pump (AHP) is consequently due to the enhancement of heat and mass transfer in the absorber (but not generator) by the induced interfacial turbulence.

Wu, W.T.; Yang, Y.M.; Maa, J.R. [National Cheng Kung Univ., Tainan (Taiwan, Province of China). Dept. of Chemical Engineering] [National Cheng Kung Univ., Tainan (Taiwan, Province of China). Dept. of Chemical Engineering

1998-11-01T23:59:59.000Z

182

Subcooled and saturated water flow boiling pressure drop in small diameter helical coils at low pressure  

SciTech Connect

Experimental pressure drop results on boiling water flow through three helical coils of tube inner diameter of 4.03 mm and 4.98 mm and coil diameter to tube diameter ratio of 26.1, 64.1 and 93.3 are presented. Both subcooled and saturated flow boiling are investigated, covering operating pressures from 120 to 660 kPa, mass fluxes from 290 to 690 kg m{sup -2} s{sup -1} and heat fluxes from 50 to 440 kW m{sup -2}. Existing correlations for subcooled flow pressure drop are found not capable to fit the present subcooled database, while the measurements in saturated flow conditions are successfully reproduced by existing correlations for both straight and coiled pipe two-phase flow. The experimental database is included in tabular form. (author)

Cioncolini, Andrea; Santini, Lorenzo; Ricotti, Marco E. [Department of Nuclear Engineering, Politecnico di Milano, via Ponzio 34/3, 20133 Milano (Italy)

2008-05-15T23:59:59.000Z

183

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

E-Print Network (OSTI)

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 ...

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

2009-01-01T23:59:59.000Z

184

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

E-Print Network (OSTI)

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 ...

Schor, Andrei L.

1982-01-01T23:59:59.000Z

185

Survey of Optimization of Reactor Coolant Cleanup Systems: For Boiling Water Reactors and Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Optimization of the reactor coolant cleanup systems in the boiling water reactor (BWR) and pressurized water reactor (PWR) environment is important for controlling the transport of corrosion products (metals and activated metals), fission products, and coolant impurities (soluble and insoluble) throughout the reactor coolant loop, and this optimization contributes to reducing primary system radiation fields. The removal of radionuclides and corrosion products is just one of many functions (both ...

2013-09-27T23:59:59.000Z

186

Corrosion Product Transport during Boiling Water Reactor and Pressurized Water Reactor Startups  

Science Conference Proceedings (OSTI)

Corrosion product transport to Pressurized Water Reactor (PWR) steam generators and to the Boiling Water Reactor (BWR) reactor vessel during startups is of increased interest due to reductions in feedwater transport rates during normal operation and the recent emphasis on minimizing total transport during the cycle. Reductions in transport will reduce deposition on the fuel and the tendency for hot spot formation in BWRs and reduce surface fouling and the tendency for formation of aggressive chemical sol...

2010-12-17T23:59:59.000Z

187

BWRVIP-167: BWR Vessel and Internals Project, Boiling Water Reactor Issue Management Tables  

Science Conference Proceedings (OSTI)

Ongoing issues related to degradation of boiling water reactor (BWR) pressure vessels, reactor internals, and American Society of Mechanical Engineers (ASME) Class 1 piping components have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) issues. This BWR Vessel and Internals Project (BWRVIP) report provides BWR Issue Management Tables that identify, rank, and describe R&D gaps.

2007-03-20T23:59:59.000Z

188

Impact of Chemical Injections on Boiling Water Reactor Dose Rates: Interim Report  

Science Conference Proceedings (OSTI)

This report investigates the effects of Boiling Water Reactor (BWR) chemistry parameters on radiation field generation, with a focus on the higher reactor water Co-60 activity levels observed at plants using On-line NobleChem™ (OLNC) injections. Correlation and response curves were developed to relate reactor water and feedwater chemistry to dose rates, with the goal of improving reactor recirculation system (RRS) piping shutdown dose rate ...

2012-12-21T23:59:59.000Z

189

COST STUDY OF A 100-Mw(e) DIRECT-CYCLE BOILING WATER REACTOR PLANT  

SciTech Connect

A technical and economic evaluation is presented of a direct-cycle light- water boiling reactor designed for natural circulation and internal steam-water separation. The reference lOO-Mw(e) reactor power plant design evolved from the study should have the best chance (compared to similar plants) of approaching the 8 to 9 mill/kwh total power-cost level. (W.D.M.)

Bullinger, C.F.; Harrer, J.M.

1960-07-01T23:59:59.000Z

190

Early Hydrogen Water Chemistry Injection in Boiling Water Reactors: Impact on Fuel Performance and Reliability  

Science Conference Proceedings (OSTI)

Early injection of hydrogen during plant startup has been proposed to further mitigate intergranular stress corrosion cracking (IGSCC) in boiling water reactors (BWRs). To assess the effectiveness of early hydrogen water chemistry (EHWC), laboratory tests were performed under simulated BWR startup conditions at 200-400°F in the absence of radiation with pre-oxidized stainless steel specimens treated with noble metals to simulate plant surfaces. The ...

2012-12-13T23:59:59.000Z

191

Microscale flow visualization of nucleate boiling in small channels: Mechanisms influencing heat transfer  

SciTech Connect

This paper describes the use of a new test apparatus employing flow visualization via ultra-high-speed video and microscope optics to study microscale nucleate boiling in a small, rectangular, heated channel. The results presented are for water. Because of confinement effects produced by the channel cross section being of the same nominal size as the individual vapor bubbles nucleating at discrete wall sites, flow regimes and heat transfer mechanisms that occur in small channels are shown to be considerably different than those in large channels. Flow visualization data are presented depicting discrete bubble/bubble and bubble/wall interactions for moderate and high heat flux. Quantitative data are also presented on nucleate bubble growth behavior for a single nucleation site in the form of growth rates, bubble sizes, and frequency of generation in the presence and absence of a thin wall liquid layer. Mechanistic boiling behavior and trends are observed which support the use of this type of research as a powerful means to gain fundamental insights into why, under some conditions, nucleate boiling heat transfer coefficients are considerably larger in small channels than in large channels.

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

1997-07-01T23:59:59.000Z

192

On the hot-spot-controlled critical heat flux mechanism in pool boiling of saturated fluids  

SciTech Connect

In this paper, we further investigate the hypothesis that the critical heat flux (CHF) occurs when some point on the heated surface reaches a high enough temperature that liquid can no longer contact that point, resulting in a gradual but continuous increase in the overall surface temperature. This hypothesis unifies the occurrence of the CHF and the quenching of hot surfaces by relating both to the same concept, i.e., the ability of a liquid to contact a hot surface. We use a two-dimensional transient conduction model to study the boiling phenomenon in the second transition region of saturated pool nucleate boiling on a horizontal surface. The heater surface is assumed to consist of two regions: a dry patch region formed as a result of complete evaporation of the thinner liquid macrolayers and a two-phase macrolayer region formed by numerous vapor stems penetrating relatively thick liquid macrolayers. The constitutive relations used to determine the stem-macrolayer configuration in the two-phase macrolayer region of the boiling surface were reevaluated for Gaertner's clean water and water-nickel/salt solution. 29 refs.

Unal, C.; Sadasivan, P.; Nelson, R.A.

1992-01-01T23:59:59.000Z

193

On the hot-spot-controlled critical heat flux mechanism in pool boiling of saturated fluids  

SciTech Connect

In this paper, we further investigate the hypothesis that the critical heat flux (CHF) occurs when some point on the heated surface reaches a high enough temperature that liquid can no longer contact that point, resulting in a gradual but continuous increase in the overall surface temperature. This hypothesis unifies the occurrence of the CHF and the quenching of hot surfaces by relating both to the same concept, i.e., the ability of a liquid to contact a hot surface. We use a two-dimensional transient conduction model to study the boiling phenomenon in the second transition region of saturated pool nucleate boiling on a horizontal surface. The heater surface is assumed to consist of two regions: a dry patch region formed as a result of complete evaporation of the thinner liquid macrolayers and a two-phase macrolayer region formed by numerous vapor stems penetrating relatively thick liquid macrolayers. The constitutive relations used to determine the stem-macrolayer configuration in the two-phase macrolayer region of the boiling surface were reevaluated for Gaertner`s clean water and water-nickel/salt solution. 29 refs.

Unal, C.; Sadasivan, P.; Nelson, R.A.

1992-05-01T23:59:59.000Z

194

Bubble confinement in flow boiling of FC-72 in a ''rectangular'' microchannel of high aspect ratio  

SciTech Connect

Boiling in microchannels remains elusive due to the lack of full understanding of the mechanisms involved. A powerful tool in achieving better comprehension of the mechanisms is detailed imaging and analysis of the two-phase flow at a fundamental level. Boiling is induced in a single microchannel geometry (hydraulic diameter 727 {mu}m), using a refrigerant FC-72, to investigate the effect of channel confinement on bubble growth. A transparent, metallic, conductive deposit has been developed on the exterior of the rectangular microchannel, allowing simultaneous uniform heating and visualisation to be achieved. The data presented in this paper is for a particular case with a uniform heat flux applied to the microchannel and inlet liquid mass flowrate held constant. In conjunction with obtaining high-speed images and videos, sensitive pressure sensors are used to record the pressure drop across the microchannel over time. Bubble nucleation and growth, as well as periodic slug flow, are observed in the microchannel test section. The periodic pressure fluctuations evidenced across the microchannel are caused by the bubble dynamics and instances of vapour blockage during confined bubble growth in the channel. The variation of the aspect ratio and the interface velocities of the growing vapour slug over time, are all observed and analysed. We follow visually the nucleation and subsequent both 'free' and 'confined' growth of a vapour bubble during flow boiling of FC-72 in a microchannel, from analysis of our results, images and video sequences with the corresponding pressure data obtained. (author)

Barber, Jacqueline [School of Engineering, University of Edinburgh, The King's Buildings, Mayfield Road, Edinburgh, EH9 3JL (United Kingdom); Aix-Marseille Universite (UI, UII) - CNRS Laboratoire IUSTI, UMR 6595, 5 Rue Enrico Fermi, Marseille 13453 (France); Brutin, David; Tadrist, Lounes [Aix-Marseille Universite (UI, UII) - CNRS Laboratoire IUSTI, UMR 6595, 5 Rue Enrico Fermi, Marseille 13453 (France); Sefiane, Khellil [School of Engineering, University of Edinburgh, The King's Buildings, Mayfield Road, Edinburgh, EH9 3JL (United Kingdom)

2010-11-15T23:59:59.000Z

195

IMPROVEMENTS RELATING TO NUCLEAR REACTORS  

SciTech Connect

In order to reduce the pumping power for the coolant in a steam-cooled reactor, in which the steam being passed through successive sections of the reactor core and being superheated there, the sections are connected in series with one another, while a plurality of de-superheaters is provided such that steam flowing from one section to the next passes through a de-superheater. The condensed steam returning to the reactor from the means utilizing the steam heat content is divided into a number of separate streams. The first stream going to the first section in the reactor core is raised at least to saturated steam outside the reactor, while the remaining streams of condensed steam are conveyed to the de-superheaters to be mixed with steam passing therethrough between successive sections of the reactor, cooling in this manner said steam and being themselves converted into steam. Increasing amounts of condensate are added in successive de-superheaters until the steam returning to the reactor from the final desuperheater is equivalent to the full mass flow of steam circulating to the heat utilizing means. (NPO)

1960-08-01T23:59:59.000Z

196

Nuclear & Uranium  

U.S. Energy Information Administration (EIA)

Nuclear & Uranium. Uranium fuel ... nuclear reactors, generation, spent fuel. Total Energy. Comprehensive data summaries, comparisons, analysis, and projections ...

197

Nuclear power and nuclear weapons  

SciTech Connect

The proliferation of nuclear weapons and the expanded use of nuclear energy for the production of electricity and other peaceful uses are compared. The difference in technologies associated with nuclear weapons and nuclear power plants are described.

Vaughen, V.C.A.

1983-01-01T23:59:59.000Z

198

Verification of physics parameters for BWR (boiling water reaction) one-dimensional transient analysis  

SciTech Connect

A data-processing method was developed to generate physics parameters for use with the one-dimensional kinetics model of the RETRAN-02/MOD3 code. The physics parameters were verified to assure the consistency in collapsing procedures and to identify the need for further improvements. In the present study, calculations were performed during the boiling water reactor-4 Chinshan-1 cycle-7 (CS1CY7) end-of-cycle (EOC) Hailing condition, CS2CY6 middle-of-cycle (MOC), and CS1CY1 beginning-of-cycle (BOC) rated conditions. This paper describes the results of verification and their implications for plant transient analyses.

Chou, H.P. (National Tsing-Hua Univ., Hsinchu (Taiwan)); Chen, Y.J. (Institute of Nuclear Energy Research, Lung-Tan (Taiwan))

1989-11-01T23:59:59.000Z

199

Numerical modeling of boiling due to production in a fractured reservoir and its field application  

Science Conference Proceedings (OSTI)

Numerical simulations were carried out to characterize the behaviors of fractured reservoirs under production which causes in-situ boiling. A radial flow model with a single production well, and a two-dimensional geothermal reservoir model with several production and injection wells were used to study the two-phase reservoir behavior. The behavior can be characterized mainly by the parameters such as the fracture spacing and matrix permeability. However, heterogeneous distribution of the steam saturation in the fracture and matrix regions brings about another complicated feature to problems of fractured two-phase reservoirs.

Yusaku Yano; Tsuneo Ishido

1995-01-26T23:59:59.000Z

200

Forced-convection boiling tests performed in parallel simulated LMR fuel assemblies  

SciTech Connect

Forced-convection tests have been carried out using parallel simulated Liquid Metal Reactor fuel assemblies in an engineering-scale sodium loop, the Thermal-Hydraulic Out-of-Reactor Safety facility. The tests, performed under single- and two-phase conditions, have shown that for low forced-convection flow there is significant flow augmentation by thermal convection, an important phenomenon under degraded shutdown heat removal conditions in an LMR. The power and flows required for boiling and dryout to occur are much higher than decay heat levels. The experimental evidence supports analytical results that heat removal from an LMR is possible with a degraded shutdown heat removal system.

Rose, S.D.; Carbajo, J.J.; Levin, A.E.; Lloyd, D.B.; Montgomery, B.H.; Wantland, J.L.

1985-04-21T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Some investigations on the enhancement of boiling heat transfer from planer surface embedded with continuous open tunnels  

Science Conference Proceedings (OSTI)

Boiling heat transfer from a flat surface can be enhanced if continuous open tunnel type structures are embedded in it. Further, improvement of boiling heat transfer from such surfaces has been tried by two separate avenues. At first, inclined tunnels are embedded over the solid surface and an effort is made to optimize the tunnel inclination for boiling heat transfer. Surfaces are manufactured in house with four different inclinations of the tunnels with or without a reentrant circular pocket at the end of the tunnel. Experiments conducted in the nucleate boiling regime showed that 45 deg inclination of the tunnels for both with and without base geometry provides the highest heat transfer coefficient. Next, active fluid rotation was imposed to enhance the heat transfer from tunnel type surfaces with and without the base geometry. Rotational speed imparted by mechanical stirrer was varied over a wide range. It was observed that fluid rotation enhances the heat transfer coefficient only up to a certain value of stirrer speed. Rotational speed values, beyond this limit, reduce the boiling heat transfer severely. A comparison shows that embedding continuous tunnel turns out to be a better option for the increase of heat transfer coefficient compared to the imposition of fluid rotation. But the behavior of inclined tunnels under the action of fluid rotation is yet to be established and can be treated as a future scope of the work. (author)

Das, A.K.; Das, P.K.; Saha, P. [Department of Mechanical Engineering, Indian Institute of Technology, Kharagpur 721 302 (India)

2010-11-15T23:59:59.000Z

202

Resistivity During Boiling in the SB-15-D Core from the Geysers Geothermal Field: The Effects of Capillarity  

DOE Green Energy (OSTI)

In a laboratory study of cores from borehole SB-15-D in The Geysers geothermal area, we measured the electrical resistivity of metashale with and without pore-pressure control, with confining pressures up to 100 bars and temperatures between 20 and 150 C, to determine how the pore-size distribution and capillarity affected boiling. We observed a gradual increase in resistivity when the downstream pore pressure or confining pressure decreased below the phase boundary of free water. For the conditions of this experiment, boiling, as indicated by an increase in resistivity, is initiated at pore pressures of approximately 0.5 to 1 bar (0.05 to 0.1 MPa) below the free-water boiling curve, and it continues to increase gradually as pressure is lowered to atmospheric. A simple model of the effects of capillarity suggests that at 145 C, less than 15% of the pore water can boil in these rocks. If subsequent experiments bear out these preliminary observations, then boiling within a geothermal reservoir is controlled not just by pressure and temperature but also by pore-size distribution. Thus, it may be possible to determine reservoir characteristics by monitoring changes in electrical resistivity as reservoir conditions change.

Roberts, J.; Duba, A.; Bonner, B.; Kasameyer, P.

1997-01-01T23:59:59.000Z

203

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207

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212

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215

Development of a Scatter Search Optimization Algorithm for Boiling Water Reactor Fuel Lattice Design  

Science Conference Proceedings (OSTI)

Technical Paper / Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications

Juan-Luis François; Cecilia Martín-del-Campo; Luis B. Morales; Miguel-Angel Palomera

216

Nuclear Resonance Fluorescence for Nuclear Materials Assay  

E-Print Network (OSTI)

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

Quiter, Brian Joseph

2010-01-01T23:59:59.000Z

217

Gas processing/The boiling behavior of LPG and liquid ethane, ethylene, propane, and n-butane spilled on water  

SciTech Connect

Boiling-rate calorimeter studies showed that unlike liquid nitrogen, methane, and LNG, LPG (84.7% propane, 6.0% ethane, and 9.3% n-butane; 442/sup 0/C bp), or pure propane, when rapidly spilled on water, reacted violently, ejecting water and ice into the vapor space; but in 1-2 sec, a coherent ice layer was formed and further boiloff was quiet and well predicted by a simple one-dimensional, moving-boundary-value, heat transfer model with a growing ice shield. Increasing the content of ethane and butane in LPG to 20% and 10%, respectively, had almost no effect on the LPG boiling, indicating that boiling may be modeled by using pure propane. Ethane, ethylene, and n-butane behaved quite differently from LPG. In spills of pure liquid propane on solid ice, the boiloff rate was almost identical to that predicted by the moving-boundary model.

Reid, R.C.; Smith, K.A.

1978-04-01T23:59:59.000Z

218

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

Science Conference Proceedings (OSTI)

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

Murphy, T.J.

1987-09-01T23:59:59.000Z

219

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

Science Conference Proceedings (OSTI)

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

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

2011-06-01T23:59:59.000Z

220

Heating surface material’s effect on subcooled flow boiling heat transfer of R134a  

Science Conference Proceedings (OSTI)

In this study, subcooled flow boiling of R134a on copper (Cu) and stainless steel (SS) heating surfaces was experimentally investigated from both macroscopic and microscopic points of view. By utilizing a high-speed digital camera, bubble growth rate, bubble departure size, and nucleation site density, were able to be observed and analyzed from the microscopic point of view. Macroscopic characteristics of the subcooled flow boiling, such as heat transfer coefficient, were able to be measured as well. Experimental results showed that there are no obvious difference between the copper and the stainless surface with respect to bubble dynamics, such as contact angle, growth rate and departure size. On the contrary, the results clearly showed a trend that the copper surface had a better performance than the stainless steel surface in terms of heat transfer coefficient. It was also observed that wall heat fluxes on both surfaces were found highly correlated with nucleation site density, as bubble hydrodynamics are similar on these two surfaces. The difference between these two surfaces was concluded as results of different surface thermal conductivities.

Ling Zou; Barclay G. Jones

2012-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Operating experience of natural circulation core cooling in boiling water reactors  

SciTech Connect

General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed.

Kullberg, C.; Jones, K.; Heath, C.

1993-08-01T23:59:59.000Z

222

AtomicNuclear Properties  

NLE Websites -- All DOE Office Websites (Extended Search)

HTML_PAGES HTML_PAGES This AtomicNuclearProperties page is under intermittent development. Suggestions and comments are welcome. Please report errors. Chemical elements: For entries in red, a pull-down menu permits selection of the physical state. Cryogenic liquid densties are at the boiling point at 1 atm. 0n 1Ps 1H 2He 3Li 4Be 5B 6C 7N 8O 9F 10Ne 11Na 12Mg 13Al 14Si 15P 16S 17Cl 18Ar 19K 20Ca 21Sc 22Ti 23V 24Cr 25Mn 26Fe 27Co 28Ni 29Cu 30Zn 31Ga 32Ge 33As 34Se 35Br 36Kr 37Rb 38Sr 39Y 40Zr 41Nb 42Mo 43Tc 44Ru 45Rh 46Pd 47Ag 48Cd 49In 50Sn 51Sb 52Te 53I 54Xe 55Cs 56Ba 57La 72Hf 73Ta 74W 75Re 76Os 77Ir 78Pt 79Au 80Hg 81Tl 82Pb 83Bi 84Po 85At 86Rn 87Fr 88Ra 89Ac 104Rf 105Db 106Sg 107Bh 108Hs 109Mt 110Ds 111Rg 112 113 114 115 116 mt 118

223

Physical modeling and numerical simulation of subcooled boiling in one- and three-dimensional representation of bundle geometry  

Science Conference Proceedings (OSTI)

Numerical simulation of subcooled boiling in one-dimensional geometry with the Homogeneous Equilibrium Model (HEM) may yield difficulties related to the very low sonic velocity associated with the HEM. These difficulties do not arise with subcritical flow. Possible solutions of the problem include introducing a relaxation of the vapor production rate. Three-dimensional simulations of subcooled boiling in bundle geometry typical of fast reactors can be performed by using two systems of conservation equations, one for the HEM and the other for a Separated Phases Model (SPM), with a smooth transition between the two models.

Bottoni, M.; Lyczkowski, R.; Ahuja, S.

1995-07-01T23:59:59.000Z

224

Critical heat flux and boiling heat transfer to water in a 3-mm-diameter horizontal tube.  

DOE Green Energy (OSTI)

Boiling of the coolant in an engine, by design or by circumstance, is limited by the critical heat flux phenomenon. As a first step in providing relevant engine design information, this study experimentally addressed both rate of boiling heat transfer and conditions at the critical point of water in a horizontal tube of 2.98 mm inside diameter and 0.9144 m heated length. Experiments were performed at system pressure of 203 kPa, mass fluxes in range of 50 to 200 kg/m{sup z}s, and inlet temperatures in range of ambient to 80 C. Experimental results and comparisons with predictive correlations are presented.

Yu, W.; Wambsganss, M. W.; Hull, J. R.; France, D. M.

2000-12-04T23:59:59.000Z

225

Analytical and experimental simulation of boiling oscillations in sodium with a low-pressure water system. [LMFBR  

SciTech Connect

An experimental and analytical program designed to simulate sodium boiling under low-power, low-flow conditions has been completed. Experiments were performed using atmospheric- pressure water as a simulant fluid and a simple one-dimensional model was developed for the system. Results indicate that water is a suitable simulant for liquid sodium under certain conditions and that the model does a fair job of modeling the system. In addition, oscillations that occur during the boiling process appear to augment substantially the heat transfer between liquid and vapor in condensation.

Levin, A.E.; Griffith, P.

1981-01-01T23:59:59.000Z

226

Modeling the Thermal Mechanical Behavior of a 300 K Vacuum Vesselthat is Cooled by Liquid Hydrogen in Film Boiling  

DOE Green Energy (OSTI)

This report discusses the results from the rupture of a thin window that is part of a 20-liter liquid hydrogen vessel. This rupture will spill liquid hydrogen onto the walls and bottom of a 300 K cylindrical vacuum vessel. The spilled hydrogen goes into film boiling, which removes the thermal energy from the vacuum vessel wall. This report analyzes the transient heat transfer in the vessel and calculates the thermal deflection and stress that will result from the boiling liquid in contact with the vessel walls. This analysis was applied to aluminum and stainless steel vessels.

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

2004-05-07T23:59:59.000Z

227

Nuclear Reactions  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactions Nuclear reactions and nuclear scattering are used to measure the properties of nuclei. Reactions that exchange energy or nucleons can be used to measure the energies of...

228

Nuclear Safety  

Energy.gov (U.S. Department of Energy (DOE))

Nuclear Safety information site that provides assistance and resources to field elements in implementation of requirements and resolving nuclear safety, facility safety, and quality assurance issues.

229

Nuclear Materials  

Science Conference Proceedings (OSTI)

Materials and Fuels for the Current and Advanced Nuclear Reactors III ... response of oxide ceramics for nuclear applications through experiment, theory, and ...

230

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)

with macrolayer model of Maruyama, we simulated the transient boiling curve for water and fluorinert FC-72(C6F14 transient CHF in saturated pool boiling. The developed model includes the analysis of thermal energy conduction within the heater coupled with a macrolayer- thinning model. The prediction indicated favorable

Maruyama, Shigeo

231

Fluid flow and reactive transport around potential nuclear waste emplacement tunnels at Yucca Mountain, Nevada  

E-Print Network (OSTI)

and refluxing of steam condensate towards the boiling front.and refluxing of steam condensate towards the boiling front.

Spycher, N.F.; Sonnenthal, E.L.; Apps, J.A.

2002-01-01T23:59:59.000Z

232

Nuclear Matter and Nuclear Dynamics  

E-Print Network (OSTI)

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

M Colonna

2009-02-26T23:59:59.000Z

233

Prediction of Boiling-Induced Natural-Circulation Flow in Engineered Cooling Channels  

Science Conference Proceedings (OSTI)

Technical Paper / Special Issue on the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) / Thermal Hydraulics

Kwang Soon Ha; Fan-Bill Cheung; Jinho Song; Rae Joon Park; Sang Baik Kim

234

NREL: Energy Analysis - Nuclear Power Results - Life Cycle Assessment  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Power Results - Life Cycle Assessment Harmonization Nuclear Power Results - Life Cycle Assessment Harmonization Over the last 30 years, analysts have conducted life cycle assessments on the environmental impacts associated with a variety of nuclear power technologies and systems. These life cycle assessments have had wide-ranging results. To better understand greenhouse gas (GHG) emissions from nuclear power systems, NREL completed a comprehensive review and analysis of life cycle assessments focused on light water reactors (LWRs)-including both boiling water reactors (BWRs) and pressurized water reactors (PWRs)-published between 1980 and 2010. NREL developed and applied a systematic approach to review life cycle assessment literature, identify primary sources of variability and, where possible, reduce variability in GHG emissions

235

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

DOE Patents (OSTI)

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.

Hill, P.R.

1994-12-27T23:59:59.000Z

236

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

SciTech Connect

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.

PACE, M.E.

2004-01-13T23:59:59.000Z

237

Oxygen suppression in boiling water reactors. Quarterly report 2, January 1--March 31, 1978  

DOE Green Energy (OSTI)

Boiling water reactors (BWR's) generally use high purity, no-additive feedwater. Primary recirculating coolant is neutral pH, and contains 100 to 300 ppB oxygen and stoichiometrically related dissolved hydrogen. However, oxygenated water increases austenitic stainless steel susceptibility to intergranular stress-corrosion cracking (IGSCC) when other requisite factors such as stress and sensitization are present. Thus, reduction or elimination of the oxygen in BWR water may preclude cracking incidents. One approach to reduction of the BWR coolant oxygen concentration is to adopt alternate water chemistry (AWC) conditions using an additive(s) to suppress or reverse radiolytic oxygen formation. Several additives are available to do this but they have seen only limited and specialized application in BWR's. The objective of this program is to perform an in-depth engineering evaluation of the potential suppression additives supported by critical experiments where required to resolve substantive uncertainties.

Burley, E.L.

1978-10-01T23:59:59.000Z

238

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

DOE Patents (OSTI)

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.

Hill, Paul R. (Tucson, AZ)

1994-01-01T23:59:59.000Z

239

Measurement of Key Pool BOiling Parameters in nanofluids for Nuclerar Applications  

Science Conference Proceedings (OSTI)

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

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

2007-01-01T23:59:59.000Z

240

A DESIGN STUDY OF A LOW POWER AQUEOUS HOMOGENEOUS BOILING REACTOR POWER PLANT  

SciTech Connect

This design study describes a reactor and associated power plant that has been designed to produce 100 kv of net electric power and 400 kv of hot water space heating at a total thermal output of 1300 kw. The fuel consists of a solution of UO/sub 2/SO/sub 4/ in light water. Power is removed from the core by boiling the fuel solution and transferring the heat to the secondary steam system by condensing primary water on the external surface of a bayonet type boiler and boiling secondary water within the tubes. Saturated steam, produced in the boiler at 225 psia (Full Power) is used to drive a turbo generator, Extraction steam from the turbine is used, at a reduced pressure, for space heating. The initial loading of the reactor is approximately 4.8 kg of U/sub 235/ and operation based on an average load factor of 80% will require fuel addition at the rate of about 580 grams per year. It may be desirable to replace the fuel in the core after a period of 5 years operation due to the accumulation of corrosion products. The reactor control is affected automatically by power demand. The major objective has been to design a reactor that is reliable and simple, requiring little if any operating personnel and routine maintenance only which can be performed by one man. The design should stress simplicity of the system, ease of erection at the site, initial transportability, reliability and ease of operation; these characteristics are then expected to result in greatly reduced effort and manpower support over a conventional system. (auth)

Mong, B.A.; Colgan, J.E.; D' Elia, R.A.; Mooradian, J.S.; Rhode, G.K.; Wood, P.M.

1955-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

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

SciTech Connect

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

Nelson, R.A.; Unal, C.

1991-01-01T23:59:59.000Z

242

Fusion Nuclear Science | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Systems Modeling, Simulation & Validation Nuclear Systems Technology Reactor Technology Nuclear Science Home | Science & Discovery | Nuclear Science | Research...

243

Nuclear Analytical Chemistry Portal  

Science Conference Proceedings (OSTI)

NIST Home > Nuclear Analytical Chemistry Portal. Nuclear Analytical Chemistry Portal. ... see all Nuclear Analytical Chemistry news ... ...

2010-08-02T23:59:59.000Z

244

State Nuclear Profiles 2009  

U.S. Energy Information Administration (EIA)

Vermont Yankee 1 620 5,361 98.7 BWR 11/30/1972 3/21/2012 620 5,361 98.7 Data for 2009 BWR = Boiling Water Reactor. License Expiration Date

245

Future of Nuclear Data for Nuclear Astrophysics  

Science Conference Proceedings (OSTI)

Nuclear astrophysics is an exciting growth area in nuclear science. Because of the enormous nuclear data needs of this field

Michael S. Smith

2005-01-01T23:59:59.000Z

246

Countering Nuclear Terrorism | National Nuclear Security Administratio...  

NLE Websites -- All DOE Office Websites (Extended Search)

Countering Nuclear Terrorism | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response...

247

Nuclear Detonation Detection | National Nuclear Security Administratio...  

National Nuclear Security Administration (NNSA)

Nuclear Nonproliferation Program Offices > Office of Nonproliferation Research & Development > Nuclear Detonation Detection Nuclear Detonation Detection Develop, Demonstrate, and...

248

Chernobyl Nuclear Accident | National Nuclear Security Administration  

NLE Websites -- All DOE Office Websites (Extended Search)

Chernobyl Nuclear Accident | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response...

249

GE Hitachi Nuclear Energy | Open Energy Information  

Open Energy Info (EERE)

GE Hitachi Nuclear Energy GE Hitachi Nuclear Energy Jump to: navigation, search Name GE Hitachi Nuclear Energy Place Wilmington, North Carolina Zip 28402 Sector Efficiency, Services Product GE Hitachi Nuclear Energy develops advanced light water reactors and offers products and services used by operators of boiling water reactor (BWR) nuclear power plants to improve efficiency and boost output. Coordinates 42.866922°, -72.868494° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":42.866922,"lon":-72.868494,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

250

Top U.S. Nuclear Official Commends Industry for Submitting 3rd Combined  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Top U.S. Nuclear Official Commends Industry for Submitting 3rd Top U.S. Nuclear Official Commends Industry for Submitting 3rd Combined Construction & Operating License Application to the NRC Top U.S. Nuclear Official Commends Industry for Submitting 3rd Combined Construction & Operating License Application to the NRC November 28, 2007 - 4:45pm Addthis RICHMOND, VA - The U.S. Department of Energy (DOE) Assistant Secretary for Nuclear Energy Dennis Spurgeon today commended Dominion North Anna, LLC (Dominion) for submission of a combined Construction and Operating License (COL) application to the Nuclear Regulatory Commission (NRC) for construction of a new nuclear power plant in the United States. Dominion's application seeks approval to build and operate one General Electric-Hitachi Economic Simplified Boiling Water Reactor (ESBWR) at its

251

Nuclear Science  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Science Science and Engineering Education Sourcebook 2013 American Nuclear Society US Department of Energy Nuclear Science & Engineering Education Sourcebook 2013 North American Edition American Nuclear Society Education, Training, and Workforce Division US Department of Energy Office of Nuclear Energy Editor and Founder John Gilligan Professor of Nuclear Engineering North Carolina State University Version 5.13 Welcome to the 2013 Edition of the Nuclear Science and Engineering Education (NS&EE) Sourcebook. We have evolved and improved! The core mission of the Sourcebook has not changed, however. Our purpose is to facilitate interaction among faculty, students, industry, and government agencies to accomplish nuclear research, teaching and service activities. Since 1986 we have compiled critical information on nuclear

252

Nuclear forces  

Science Conference Proceedings (OSTI)

These lectures present an introduction into the theory of nuclear forces. We focus mainly on the modern approach

2013-01-01T23:59:59.000Z

253

Nuclear weapons, nuclear effects, nuclear war  

SciTech Connect

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

Bing, G.F.

1991-08-20T23:59:59.000Z

254

Nuclear-Coupled Flow Instabilities and Their Effects on Dryout  

SciTech Connect

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.

M. Ishii; X. Sunn; S. Kuran

2004-09-27T23:59:59.000Z

255

FUEL CYCLE PROGRAM, A BOILING WATER REACTOR RESEARCH DEVELOPMENT PROGRAM. First Summary Report for March 1959-July 1960  

SciTech Connect

The Fuel Cycle Development Program is a basic development program for boiling and other water technology. It covers the areas of oxide fuel fabrication. irradiation. and examination; the physics of water-moderated reactore; and boiling-water heat transfer and stability. Schedules for the fuel- cycle program were examined. and it was concluded that portions of the Task A program should be conducted during the period May to Dec. 1959 in order to keep costs of the work as low as possible and to allow initiation of the fuel-cycle program at the earliest possible date after the Vallecitos BWR was returned to service. The basis for the scheduling of the work is discussed. and a chronological summary describing the content of the work is given. Technical progress is outlined and details are summarized. Subsequent reports issued monthly and quarterly will summarize the progress of the prognam. (W.D.M.)

Cook, W.H.

1961-10-31T23:59:59.000Z

256

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

E-Print Network (OSTI)

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 ...

Gerardi, Craig Douglas

2009-01-01T23:59:59.000Z

257

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

E-Print Network (OSTI)

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, ...

No, Hee Cheon

1983-01-01T23:59:59.000Z

258

BWRVIP-270, Revision 1: BWR Vessel and Internals Project, Compilation of Fluence Estimates for Boiling Water Reactor Materials  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP) is an association of utilities focused on BWR vessel and internals issues. Many of the BWR internal components receive high exposure to neutron flux due to their proximity to the fuel in the Reactor Pressure Vessel (RPV). Identifying how predicted fluence values will impact the materials at these locations is a focus of the BWRVIP proactive materials strategy. As part of this approach, this report provides visual and tabular summaries ...

2013-12-09T23:59:59.000Z

259

Nuclear Deterrence  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Deterrence Nuclear Deterrence Nuclear Deterrence LANL's mission is to develop and apply science and technology to ensure the safety, security, and effectiveness of the U.S. nuclear deterrent; reduce global threats; and solve other emerging national security and energy challenges. April 12, 2012 A B-2 Spirit bomber refuels from a KC-135 Stratotanker A B-2 Spirit bomber refuels from a KC-135 Stratotanker. Contact Operator Los Alamos National Laboratory (505) 667-5061 Charlie McMillan, Director: "For the last 70 years there has not been a world war, and I have to think that our strong deterrent has something to do with that fact." Mission nuclear weapons Charlie McMillan, Director of Los Alamos National Laboratory 1:06 Director McMillan on nuclear deterrence While the role and prominence of nuclear weapons in U.S. security policy

260

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

DOE Green Energy (OSTI)

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

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

1990-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Multivent effects in a large scale boiling water reactor pressure suppression system  

Science Conference Proceedings (OSTI)

The steam-driven GKSS pressure suppression test facility, which contains 3 full scale vent pipes, has been used for 5 years to investigate the postulated loss-of-coolant accident in a Mark II and Type 69 boiling water reactor. Using the results from several of these tests, wetwell boundary load data (peak pressures and spectral power) during the chugging stage, have been evaluated for sparse pool response (one and two vents in the three vent pool) and for full pool response (one, two, or three vent operation in pools of constant wetwell pool area per vent). The sparse pool results indicate the pool-system, chug event boundary loads are strongly dependent on wetwell pool area per vent, with the load increasing with decreasing area. The full pool results show a substantial increase in the pool-system, chug event boundary loads upon a change from single cell to double cell operation; only minor change occurs in going from double to triple cell operation.

McCauley, E.W.; Aust, E.; Schwan, H.

1984-07-06T23:59:59.000Z

262

An Improved Model for Assessing the Effectiveness of Hydrogen Water Chemistry in Boiling Water Reactors  

Science Conference Proceedings (OSTI)

For nearly two decades, hydrogen water chemistry (HWC) has been used as a remedial measure to protect boiling water reactor (BWR) structural components against intergranular stress corrosion cracking (IGSCC). In this paper, computer modeling is used to evaluate the effectiveness of HWC for BWRs. The DEMACE computer code, equipped with an updated chemical reaction set, G values, and a Sherwood number, is adopted to predict the chemical species concentration and electrochemical corrosion potential (ECP) responses to HWC in the primary heat transport circuit of a typical BWR. In addition, plant-specific neutron and gamma dose rate profiles are reported. DEMACE is calibrated against the data of oxygen concentration variation as a function of feedwater hydrogen concentration in the recirculation system of the Chinshan Unit 2 BWR.The determinant result for assessing the effectiveness of HWC is the ECP. For a typical BWR/4-type reactor such as Chinshan Unit 2, it is found that protecting the core channel and the lower plenum outlet is quite difficult even though the feedwater hydrogen concentration is as high as 2 ppm, based on the predicted species concentration and ECP data. However, for regions other than those mentioned earlier, a moderate amount of hydrogen added to the feedwater (0.9 ppm) is enough to achieve the desired protection against IGSCC.

Yeh, T.-K. [National Tsing-Hua University, Taiwan (China); Chu Fang [Taiwan Power Company (China)

2001-10-15T23:59:59.000Z

263

Decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR): Project final report, Argonne National Laboratory  

SciTech Connect

The Final Report for the Decontamination and Decommissioning (D&D) of the Argonne National Laboratory - East (ANL-E) Experimental Boiling Water Reactor (EBWR) facility contains the descriptions and evaluations of the activities and the results of the EBWR D&D project. It provides the following information: (1) An overall description of the ANL-E site and EBWR facility. (2) The history of the EBWR facility. (3) A description of the D&D activities conducted during the EBWR project. (4) A summary of the final status of the facility, including the final and confirmation surveys. (5) A summary of the final cost, schedule, and personnel exposure associated with the project, including a summary of the total waste generated. This project report covers the entire EBWR D&D project, from the initiation of Phase I activities to final project closeout. After the confirmation survey, the EBWR facility was released as a {open_quotes}Radiologically Controlled Area,{close_quotes} noting residual elevated activity remains in inaccessible areas. However, exposure levels in accessible areas are at background levels. Personnel working in accessible areas do not need Radiation Work Permits, radiation monitors, or other radiological controls. Planned use for the containment structure is as an interim transuranic waste storage facility (after conversion).

Fellhauer, C.R.; Boing, L.E. [Argonne National Lab., IL (United States); Aldana, J. [NES, Inc., Danbury, CT (United States)

1997-03-01T23:59:59.000Z

264

Experimental studies of adiabatic flow boiling in fractal-like branching microchannels  

SciTech Connect

Experimental results of adiabatic boiling of water flowing through a fractal-like branching microchannel network are presented and compared to numerical model simulations. The goal is to assess the ability of current pressure loss models applied to a bifurcating flow geometry. The fractal-like branching channel network is based on channel length and width ratios between adjacent branching levels of 2{sup -1/2}. There are four branching sections for a total flow length of 18 mm, a channel height of 150 {mu}m and a terminal channel width of 100 {mu}m. The channels were Deep Reactive Ion Etched (DRIE) into a silicon disk. A Pyrex disk was anodically bonded to the silicon to form the channel top to allow visualization of the flow within the channels. The flow rates ranged from 100 to 225 g/min and the inlet subcooling levels varied from 0.5 to 6 C. Pressure drop along the flow network and time averaged void fraction in each branching level were measured for each of the test conditions. The measured pressure drop ranged from 20 to 90 kPa, and the measured void fraction ranged from 0.3 to 0.9. The measured pressure drop results agree well with separated flow model predictions accounting for the varying flow geometry. The measured void fraction results followed the same trends as the model; however, the scatter in the experimental results is rather large. (author)

Daniels, Brian J.; Liburdy, James A.; Pence, Deborah V. [Mechanical Engineering, Oregon State University, Corvallis, OR 97330 (United States)

2011-01-15T23:59:59.000Z

265

Impact of aspect ratio on flow boiling of water in rectangular microchannels  

SciTech Connect

In this paper we focus on the impact of varying the aspect ratio of rectangular microchannels, on the overall pressure drop involving water boiling. An integrated system comprising micro-heaters, sensors and microchannels has been realized on (110) silicon wafers, following CMOS compatible process steps. Rectangular microchannels were fabricated with varying aspect ratios (width [W] to depth [H]) but constant hydraulic diameter of 142{+-}2{mu}m and length of 20 mm. The invariant nature of the hydraulic diameter is confirmed through two independent means: physical measurements using profilometer and by measuring the pressure drop in single-phase fluid flow. The experimental results show that the pressure drop for two-phase flow in rectangular microchannels experiences minima at an aspect ratio of about 1.6. The minimum is possibly due to opposing trends of frictional and acceleration pressure drops, with respect to aspect ratio. In a certain heat flux and mass flux range, it is observed that the two-phase pressure drop is lower than the corresponding single-phase value. This is the first study to investigate the effect of aspect ratio in two-phase flow in microchannels, to the best of our knowledge. The results are in qualitative agreement with annular flow model predictions. These results improve the possibility of designing effective heat-sinks based on two-phase fluid flow in microchannels. (author)

Singh, S.G.; Kulkarni, A.; Duttagupta, S.P. [Nanoelectronics Center, Department of Electrical Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400 076 (India); Puranik, B.P.; Agrawal, A. [Suman Mashruwala Lab, Department of Mechanical Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400 076 (India)

2008-10-15T23:59:59.000Z

266

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

SciTech Connect

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.

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

1983-03-01T23:59:59.000Z

267

Nuclear Energy  

Nuclear Energy Environmental Mgmt. Study Objectives: Respond to the pressing need to refine existing corrosion models: Predict performance in wide range of environments

268

Nuclear Reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactors Nuclear reactors created not only large amounts of plutonium needed for the weapons programs, but a variety of other interesting and useful radioisotopes. They produced...

269

Nuclear Astrophysics  

Science Conference Proceedings (OSTI)

I review progress that has been made in nuclear astrophysics over the past few years and summarize some of the questions that remain. Topics selected include solar neutrinos

W. C. Haxton

2006-01-01T23:59:59.000Z

270

Nuclear & Uranium  

U.S. Energy Information Administration (EIA)

Table 17. Purchases of enrichment services by owners and operators of U.S. civilian nuclear power reactors by contract type in delivery year, 2012

271

Nuclear Weapons  

NLE Websites -- All DOE Office Websites (Extended Search)

nuclear science that has had a significant global influence. Following the observation of fission products of uranium by Hahn and Strassmann in 1938, a uranium fission weapon...

272

NUCLEAR ENERGY  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

could improve the economic and safety performance of these advanced reactors. Nuclear power can reduce GHG emissions from electricity production and possibly in co-generation...

273

Nuclear Forces and Nuclear Systems  

NLE Websites -- All DOE Office Websites (Extended Search)

Forces and Nuclear Systems Forces and Nuclear Systems Our goal is to achieve a description of nuclear systems ranging in size from the deuteron to nuclear matter and neutron stars using a single parameterization of the nuclear forces. Our work includes both the construction of two- and three-nucleon potentials and the development of many-body techniques for computing nuclear properties with these interactions. Detailed quantitative, computationally intense studies are essential parts of this work. In the last decade we have constructed several realistic two- and three-nucleon potential models. The NN potential, Argonne v18, has a dominant charge-independent piece plus additional charge-dependent and charge-symmetry-breaking terms, including a complete electromagnetic interaction. It fits 4301 pp and np elastic scattering data with a chi**2

274

Nuclear Weapons Journal Archive  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Weapons Journal Archive Nuclear Weapons Journal The Nuclear Weapons Journal ceased publication after Issue 2, 2009. Below are Nuclear Weapons Journal archived issues. Issue...

275

Nonreactor Nuclear Facilities Division  

NLE Websites -- All DOE Office Websites (Extended Search)

role in developing science and technology for nuclear power programs, nuclear propulsion, nuclear medicine, and the nation's nuclear weapon program among others. Many...

276

Nuclear hadrodynamics  

Science Conference Proceedings (OSTI)

The role of hadron dynamics in the nucleus is illustrated to show the importance of nuclear medium effects in hadron interactions. The low lying hadron spectrum is considered to provide the natural collective variable for nuclear systems. Recent studies of nucleon?nucleon and delta?nucleon interactions are reviewed

D. F. Geesaman

1984-01-01T23:59:59.000Z

277

PROBING DENSE NUCLEAR MATTER VIA NUCLEAR COLLISIONS  

E-Print Network (OSTI)

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

Stocker, H.

2012-01-01T23:59:59.000Z

278

Nuclear Materials Management & Safeguards System | National Nuclear...  

National Nuclear Security Administration (NNSA)

Management & Safeguards System Nuclear Materials Management & Safeguards System NMMSS U.S. Department of Energy U.S. Nuclear Regulatory Commission Nuclear Materials...

279

Nuclear Materials Management & Safeguards System | National Nuclear...  

NLE Websites -- All DOE Office Websites (Extended Search)

Our Jobs Our Jobs Working at NNSA Blog Nuclear Materials Management & Safeguards System Home > About Us > Our Programs > Nuclear Security > Nuclear Materials Management &...

280

Nuclear Resonance Fluorescence for Nuclear Materials Assay  

E-Print Network (OSTI)

that are of interest for nuclear security applications. Theof interest to nuclear security. To either make theseother targets of nuclear security interest, such kilogram-

Quiter, Brian Joseph

2010-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Nuclear Resonance Fluorescence for Nuclear Materials Assay  

E-Print Network (OSTI)

and Diablo Canyon 2 nuclear reactors. Data were taken fromCapacity Operation of nuclear reactors for power generationby the operation of nuclear reactors. Therefore, ap-

Quiter, Brian Joseph

2010-01-01T23:59:59.000Z

282

Nuclear Materials Management & Safeguards System | National Nuclear...  

National Nuclear Security Administration (NNSA)

System Nuclear Materials Management & Safeguards System NMMSS U.S. Department of Energy U.S. Nuclear Regulatory Commission Nuclear Materials Management & Safeguards System...

283

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

NLE Websites -- All DOE Office Websites (Extended Search)

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

284

Nuclear Resonance Fluorescence for Nuclear Materials Assay  

E-Print Network (OSTI)

Energy Transmission say for Nuclear Fuel Assemblies 4.1Facilities Spent nuclear fuel is another example wherein intact spent nuclear fuel would be a technological

Quiter, Brian Joseph

2010-01-01T23:59:59.000Z

285

Nuclear Halos  

Science Conference Proceedings (OSTI)

We show that extreme nuclear halos are caused only by pairs of s?wave neutrons (or single s?wave neutrons) and that such states occur much more frequently in the periodic table than previously believed. Besides lingering long near zero neutron separation energy such extreme halos have very remarkable properties: they can contribute significantly to the nuclear density at more than twice the normal nuclear radius and their spreading width can be very narrow. The properties of these states are primarily determined by the “thickness” of the nuclear surface in the mean?free nuclear potential and thus their importance increases greatly as we approach the neutron drip line. We discuss what such extreme halos are

Erich Vogt

2010-01-01T23:59:59.000Z

286

Nuclear fuel elements made from nanophase materials  

SciTech Connect

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.

Heubeck, Norman B. (Schenectady, NY)

1998-01-01T23:59:59.000Z

287

Nuclear fuel elements made from nanophase materials  

DOE Patents (OSTI)

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 3,000 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.

Heubeck, Norman B.

1997-12-01T23:59:59.000Z

288

Nuclear fuel elements made from nanophase materials  

DOE Patents (OSTI)

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.

Heubeck, N.B.

1998-09-08T23:59:59.000Z

289

Nuclear Astrophysics  

E-Print Network (OSTI)

Nuclear physics has a long and productive history of application to astrophysics which continues today. Advances in the accuracy and breadth of astrophysical data and theory drive the need for better experimental and theoretical understanding of the underlying nuclear physics. This paper will review some of the scenarios where nuclear physics plays an important role, including Big Bang Nucleosynthesis, neutrino production by our sun, nucleosynthesis in novae, the creation of elements heavier than iron, and neutron stars. Big-bang nucleosynthesis is concerned with the formation of elements with A nuclear physics inputs required are few-nucleon reaction cross sections. The nucleosynthesis of heavier elements involves a variety of proton-, alpha-, neutron-, and photon-induced reactions, coupled with radioactive decay. The advent of radioactive ion beam facilities has opened an important new avenue for studying these processes, as many involve radioactive species. Nuclear physics also plays an important role in neutron stars: both the nuclear equation of state and cooling processes involving neutrino emission play a very important role. Recent developments and also the interplay between nuclear physics and astrophysics will be highlighted.

Carl R. Brune

2005-02-28T23:59:59.000Z

290

Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios  

Science Conference Proceedings (OSTI)

For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

1998-04-01T23:59:59.000Z

291

(Nuclear theory). [Research in nuclear physics  

SciTech Connect

This report discusses research in nuclear physics. Topics covered in this paper are: symmetry principles; nuclear astrophysics; nuclear structure; quark-gluon plasma; quantum chromodynamics; symmetry breaking; nuclear deformation; and cold fusion. (LSP)

Haxton, W.

1990-01-01T23:59:59.000Z

292

Development of a fully-consistent reduced order model to study instabilities in boiling water reactors  

SciTech Connect

A simple nonlinear Reduced Order Model to study global, regional and local instabilities in Boiling Water Reactors is described. The ROM consists of three submodels: neutron-kinetic, thermal-hydraulic and heat-transfer models. The neutron-kinetic model allows representing the time evolution of the three first neutron kinetic modes: the fundamental, the first and the second azimuthal modes. The thermal-hydraulic model describes four heated channels in order to correctly simulate out-of-phase behavior. The coupling between the different submodels is performed via both void and Doppler feedback mechanisms. After proper spatial homogenization, the governing equations are discretized in the time-domain. Several modifications, compared to other existing ROMs, have been implemented, and are reported in this paper. One novelty of the ROM is the inclusion of both azimuthal modes, which allows to study combined instabilities (in-phase and out-of-phase), as well as to investigate the corresponding interference effects between them. The second modification concerns the precise estimation of so-called reactivity coefficients or C{sub mn}{sup *V,D} - coefficients by using direct cross-section data from SIMULATE-3 combined with the CORE SIM core simulator in order to calculate Eigenmodes. Furthermore, a non-uniform two-step axial power profile is introduced to simulate the separate heat production in the single and two-phase regions, respectively. An iterative procedure was developed to calculate the solution to the coupled neutron-kinetic/thermal-hydraulic static problem prior to solving the time-dependent problem. Besides, the possibility of taking into account the effect of local instabilities is demonstrated in a simplified manner. The present ROM is applied to the investigation of an actual instability that occurred at the Swedish Forsmark-1 BWR in 1996/1997. The results generated by the ROM are compared with real power plant measurements performed during stability tests and show a good qualitative agreement. The present study provides some insight in a deeper understanding of the physical principles which drive both core-wide and local instabilities. (authors)

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

2012-07-01T23:59:59.000Z

293

Flow instabilities in the core and the coolant circuit of advances low-boiling light water reacto: classification of causes and development of simulator for the future analysis  

E-Print Network (OSTI)

The coolant flow instability, apparent in the coolant mass flow fluctuations in the separate parallel heating channels and also in a closed loop of the primary circuit under some operating conditions, is observed in the core fuel assemblies of light water reactors. In some ways this phenomenon is identical with the fluctuations in the once-through steam generators pipes, and changes of the coolant mass flows and length of flow patterns are initiating this phenomenon. The parameters at the core output and the secondary circuit parameters have influence on each other. These parameter changes have significant influences on the operating processes, operating and control algorithms, operating and control system design, and reliability of the operating power plant's machines and equipment. Changes of heating surface temperatures, displacement borders of the flow patterns, and critical heat flux entail changes of the coolant flow parameters, finally causing changes of the initial primary system parameters due to closed loop system feedback. In turn, these cause over-circuit instability in the reactor. Core power generation changes are carried out by means of influencing the nuclear fission process through changing the multiplication factor. Additionally, these local side-to-side power irregularities in sub-zones may appear due to the influence of various hydrodynamic instabilities. The local side-to-side power in these sub-zones may differ significantly from each other. The aforesaid arguments are correct for the both light water reactor types. But, as is shown by our investigations and operational practice of low-boiling reactors, behavior of the core-circuit hydrodynamic system is significantly different from its behavior in the boiling or pressurized reactors with pumping circulation. The coolant flow regimes in typical reactors are defined through pump operating regimes and are not adjustable inside a certain power range. The objective of this thesis is to understand more precisely the influence and the nature of these phenomena. After analyzing the problem from different points of view and showing the necessity of its comprehensive understanding, we present recommendations for engineering solutions and plans for further investigations. We will try to determine limits of their reliable practical application with modern low-to-medium power reactor design and investigate this dynamic system behavior. Finally, it is necessary to take into consideration not separate phenomena, but their complex influence on the whole primary system (i.e. a kind of macro-system is examined without being separated into its individual elements). But, the analysis of every phenomenon is fulfilled separately and a process of formation of a block-scheme, consisting of several sub-systems, is given in this thesis. The final block-scheme is presented as a simulator model, taking into consideration design components chosen for the analysis of system dynamics as the first step of model development.

Rezvyi, Aleksey

2002-01-01T23:59:59.000Z

294

Nuclear Forensics | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

295

Nuclear Incident Team | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

296

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

National Nuclear Security Administration (NNSA)

Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

297

Countering Nuclear Terrorism and Trafficking | National Nuclear...  

National Nuclear Security Administration (NNSA)

Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

298

Nuclear Safeguards | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

299

Nuclear Controls | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

300

Nuclear Nonproliferation Treaty | National Nuclear Security Administra...  

National Nuclear Security Administration (NNSA)

Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Nuclear Verification | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Verification | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our...

302

Net energy from nuclear power  

SciTech Connect

An analysis of net energy from nuclear power plants is dependent on a large number of variables and assumptions. The energy requirements as they relate to reactor type, concentration of uranium in the ore, enrichment tails assays, and possible recycle of uranium and plutonium were examined. Specifically, four reactor types were considered: pressurized water reactor, boiling water reactor, high temperature gas-cooled reactor, and heavy water reactor (CANDU). The energy requirements of systems employing both conventional (current) ores with uranium concentration of 0.176 percent and Chattanooga Shales with uranium concentration of 0.006 percent were determined. Data were given for no recycle, uranium recycle only, and uranium plus plutonium recycle. Starting with the energy requirements in the mining process and continuing through fuel reprocessing and waste storage, an evaluation of both electrical energy requirements and thermal energy requirements of each process was made. All of the energy, direct and indirect, required by the processing of uranium in order to produce electrical power was obtained by adding the quantities for the individual processes. The energy inputs required for the operation of a nuclear power system for an assumed life of approximately 30 years are tabulated for nine example cases. The input requirements were based on the production of 197,100,000 MWH(e), i.e., the operation of a 1000 MW(e) plant for 30 years with an average plant factor of 0.75. Both electrical requirements and thermal energy requirements are tabulated, and it should be emphasized that both quantities are needed. It was found that the electricity generated far exceeded the energy input requirements for all the cases considered. (auth)

Rotty, R.M.; Perry, A.M.; Reister, D.B.

1975-11-01T23:59:59.000Z

303

Alcohol-free alkoxide process for containing nuclear waste  

DOE Patents (OSTI)

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.

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

1984-01-01T23:59:59.000Z

304

Examination of turbine discs from nuclear power plants  

SciTech Connect

Investigations were performed on a cracked turbine disc from the Cooper Nuclear Power Station, and on two failed turbine discs (governor and generator ends) from the Yankee-Rowe Nuclear Power Station. Cooper is a boiling water reactor (BWR) which went into commercial operation in July 1974, and Yankee-Rowe is a pressurized water reactor (PWR) which went into commercial operation in June 1961. Cracks were identified in the bore of the Cooper disc after 41,913 hours of operation, and the disc removed for repair. At Yankee-Rowe two discs failed after 100,000 hours of operation. Samples of the Cooper disc and both Yankee-Rowe disc (one from the governor and one from the generator end of the LP turbine) were sent to Brookhaven National Laboratory (BNL) for failure analysis.

Czajkowski, C.J.; Weeks, J.R.

1982-01-01T23:59:59.000Z

305

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

306

Nuclear Magnetic Resonance Laboratory  

Science Conference Proceedings (OSTI)

Nuclear Magnetic Resonance Laboratory. ... A 600 MHz Nuclear Magnetic Resonance Spectrometer. Analytical Data Compilation Reference Materials. ...

2012-10-01T23:59:59.000Z

307

Nuclear Chirality  

Science Conference Proceedings (OSTI)

Nuclear chirality is a novel manifestation of spontaneous symmetry breaking resulting from an orthogonal coupling of angular momentum vectors in triaxial nuclei. Three perpendicular angular momenta can form two systems of opposite handedness; the time reversal operator

Krzysztof Starosta

2005-01-01T23:59:59.000Z

308

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

1962-10-23T23:59:59.000Z

309

Nuclear Materials  

Science Conference Proceedings (OSTI)

Assessing the Thermal Stability of Bulk Metallic Glasses for Nuclear Waste Applications by K. Hildal, J.H. Perepezko, and L. Kaufman, $10.00 ($10.00), $25.00.

310

Nuclear & Uranium  

U.S. Energy Information Administration (EIA)

Table 21. Foreign sales of uranium from U.S. suppliers and owners and operators of U.S. civilian nuclear power reactors by origin and delivery year, 2008-2012

311

Nuclear Nonproliferation  

Science Conference Proceedings (OSTI)

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.

Atkins-Duffin, C E

2008-12-10T23:59:59.000Z

312

Roughness and surface material effects on nucleate boiling heat transfer from cylindrical surfaces to refrigerants R-134a and R-123  

SciTech Connect

This paper presents results of an experimental investigation carried out to determine the effects of the surface roughness of different materials on nucleate boiling heat transfer of refrigerants R-134a and R-123. Experiments have been performed over cylindrical surfaces of copper, brass and stainless steel. Surfaces have been treated by different methods in order to obtain an average roughness, Ra, varying from 0.03 {mu}m to 10.5 {mu}m. Boiling curves at different reduced pressures have been raised as part of the investigation. The obtained results have shown significant effects of the surface material, with brass being the best performing and stainless steel the worst. Polished surfaces seem to present slightly better performance than the sand paper roughened. Boiling on very rough surfaces presents a peculiar behavior characterized by good thermal performance at low heat fluxes, the performance deteriorating at high heat fluxes with respect to smoother surfaces. (author)

Jabardo, Jose M. Saiz [Escuela Politecnica Superior, Universidad de la Coruna, Mendizabal s/n Esteiro, 15403 Ferrol, Coruna (Spain); Ribatski, Gherhardt; Stelute, Elvio [Department of Mechanical Engineering, Escola de Engenharia de Sao Carlos (EESC), University of Sao Paulo (USP), Av. Trabalhador Saocarlense 400 Centro, 13566-590 Sao Carlos, SP (Brazil)

2009-04-15T23:59:59.000Z

313

International Cooperation on Safety of Nuclear Plants - Nuclear...  

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Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

314

Current R&D Activities in Nuclear Criticality Safety - Nuclear...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

315

NUCLEAR DATA AND MEASUREMENTS REPORTS 161-180 - Nuclear Data...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

316

Analysis Tools for Nuclear Criticality Safety - Nuclear Engineering...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

317

Countering Nuclear Terrorism | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Countering Nuclear Terrorism | National Nuclear Security Administration Countering Nuclear Terrorism | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog The National Nuclear Security Administration Countering Nuclear Terrorism Home > Our Mission > Countering Nuclear Terrorism Countering Nuclear Terrorism NNSA provides expertise, practical tools, and technically informed policy

318

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

SciTech Connect

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.

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

1996-01-01T23:59:59.000Z

319

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

SciTech Connect

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.

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

1995-01-01T23:59:59.000Z

320

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

DOE Patents (OSTI)

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.

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

1995-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

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

DOE Patents (OSTI)

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.

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

1996-01-01T23:59:59.000Z

322

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

SciTech Connect

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.

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

1982-03-01T23:59:59.000Z

323

Organization - Nuclear Engineering Division (Argonne)  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

324

Achievements: Nuclear Engineering Division (Argonne)  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

325

Nuclear scales  

Science Conference Proceedings (OSTI)

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

Friar, J.L.

1998-12-01T23:59:59.000Z

326

Nuclear Power  

E-Print Network (OSTI)

The world of the twenty first century is an energy consuming society. Due to increasing population and living standards, each year the world requires more energy and new efficient systems for delivering it. Furthermore, the new systems must 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 that offer electricity and a multitude of co-generation energy products ranging from potable water to heat for industrial applications. The goal of the book is to show the current state-of-the-art in the covered technical areas as well as to demonstrate how general engineering principles and methods can be applied to nuclear power systems.

Tsvetkov, Pavel

2010-08-01T23:59:59.000Z

327

Nuclear Reactions  

E-Print Network (OSTI)

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.

C. A. Bertulani

2009-08-22T23:59:59.000Z

328

Nuclear Models  

SciTech Connect

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.

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

2010-09-10T23:59:59.000Z

329

Collaborating Organizations - Nuclear Data Program, Nuclear Engineering  

NLE Websites -- All DOE Office Websites (Extended Search)

Collaborating Organizations Collaborating Organizations Nuclear Data Program Overview Current Projects & Recent Activities Collaborating Organizations Publications Nuclear Data Measurements (NDM) Reports Experimental Nuclear Data Resources Contact ND Program Related Resources Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Data Program Collaborating Organizations Bookmark and Share National Nuclear Data Center, Brookhaven National Laboratory, Upton, New York. International Nuclear Structure and Decay Data Network, coordinated by IAEA, Vienna, Austria Heavy-Ion Nuclear Physics Group, Physics Division, Argonne National Laboratory, Argonne, Illinois. Nuclear Spectroscopy Group, Department of Nuclear Physics,

330

Nuclear Data Program - Nuclear Engineering Division (Argonne)  

NLE Websites -- All DOE Office Websites (Extended Search)

Data Program Data Program Nuclear Data Program Overview Current Projects & Recent Activities Collaborating Organizations Publications Nuclear Data Measurements (NDM) Reports Experimental Nuclear Data Resources Contact ND Program Related Resources Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Data Program We contribute to the development of comprehensive nuclear reactions and nuclear structure databases, including nuclear data measurement, analysis, modeling and evaluation methodologies, that are implemented in basic science research and advanced nuclear technologies. Bookmark and Share Recent Events Nuclear Structure 2012 Conference Argonne National Laboratory hosted the

331

Powering the Nuclear Navy | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

The National Nuclear Security Administration Powering the Nuclear Navy Home > Our Mission > Powering the Nuclear Navy Powering the Nuclear Navy The Naval Nuclear Propulsion Program...

332

Radionuclides in United States commercial nuclear power reactors  

SciTech Connect

In the next ten to twenty years, many of the commercial nuclear power reactors in the United States will be reaching their projected lifetime of forty years. As these power plants are decommissioned, it seems prudent to consider the recycling of structural materials such as stainless steel. Some of these materials and components have become radioactive through either nuclear activation of the elements within the components or surface contamination with radioactivity form the operational activities. In order to understand the problems associated with recycling stainless steel from decommissioned nuclear power reactors, it is necessary to have information on the radionuclides expected on or in the contaminated materials. A study has been conducted of radionuclide contamination information that is available for commercial nuclear power reactors in the United States. There are two types of nuclear power reactors in commercial use in the United States, pressurized water reactors (PWRs) and boiling water reactors (BWRs). Before presenting radionuclide activities information, a brief discussion is given on the major components and operational differences for the PWRs and BWRs. Radionuclide contamination information is presented from 11 PWRs and over 8 BWRs. These data include both the radionuclides within the circulating reactor coolant water as well as radionuclide contamination on and within component parts.

Bechtold, T.E. [ed.] [Westinghouse Idaho Nuclear Co., Inc., Idaho Falls, ID (United States); Dyer, N.C. [Oregon Graduate Inst. of Science and Technology, Beaverton, OR (United States)

1994-01-01T23:59:59.000Z

333

National Nuclear Data Center Nuclear Energy  

E-Print Network (OSTI)

National Nuclear Data Center and Nuclear Energy Pavel Oblozinsky National Nuclear Data Center;National Nuclear Data Center Probably the oldest active organization at BNL History · Founded in 1952 as Sigma Center, neutron cross sections · Changed to National Nuclear Data Center in 1977 · 40 staff

334

Midwest Nuclear Compact (Iowa)  

Energy.gov (U.S. Department of Energy (DOE))

The Midwest Nuclear Compact establishes a Midwest Nuclear Board to cooperatively evaluate and make recommendations regarding the development of nuclear technology, distribute information about...

335

Nuclear Science & Technology  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Science & Technology Nuclear Science & Technology Nuclear Science & Technology1354608000000Nuclear Science & TechnologySome of these resources are LANL-only and will require Remote Access. /No/ Nuclear Science & Technology Some of these resources are LANL-only and will require Remote Access. Key Resources Databases Organizations Journals Key Resources International Atomic Energy Agency IAEA scientific and technical publications cover areas of nuclear power, radiation therapy, nuclear security, nuclear law, and emergency repose. Search under Publications/Books and Reports for scientific books, standards, technical guides and reports National Nuclear Data Center Nuclear physics data for basic nuclear research and for applied nuclear technologies, operated by Brookhaven.

336

Nuclear | Department of Energy  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Nuclear Radioisotope Power Systems, a strong partnership between the Energy Department's Office of Nuclear Energy and NASA, has been providing the energy for deep space...

337

Brookhaven Nuclear Physics  

NLE Websites -- All DOE Office Websites (Extended Search)

Brookhaven Nuclear Physics Historically, nuclear physicists have studied the structure, characteristics, and behavior of the atomic nucleus and the nature of the nuclear force....

338

NUCLEAR PROXIMITY FORCES  

E-Print Network (OSTI)

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

Randrup, J.

2011-01-01T23:59:59.000Z

339

Nuclear | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Science & Innovation Energy Sources Nuclear Nuclear Radioisotope Power Systems, a strong partnership between the Energy Department's Office of Nuclear Energy and NASA, has...

340

Nuclear explosions  

Science Conference Proceedings (OSTI)

A summary of the physics of a nuclear bomb explosion and its effects on human beings is presented at the level of a sophomore general physics course without calculus. It is designed to supplement a standard text for such a course and problems are included.

A. A. Broyles

1982-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Nuclear ferromagnetism  

Science Conference Proceedings (OSTI)

The possibility of producing ordered states of nuclear spins by DNP followed by ADRF was first demonstrated in 1969. The spins of 19F in a crystal of CaF2 were cooled below one microdegree (with the applied field along the [100] axis) and their antiferromagnetic ordering was exhibited through the characteristic behaviour of their transverse and (later) longitudinal susceptibilities.

A. Abragam

1975-01-01T23:59:59.000Z

342

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described that includes spaced vertical fuel elements centrally disposed in a pressure vessel, a mass of graphite particles in the pressure vessel, means for fluidizing the graphite particles, and coolant tubes in the pressure vessel laterally spaced from the fuel elements. (AEC)

Post, R.G.

1963-05-01T23:59:59.000Z

343

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

Starr, C.

1963-01-01T23:59:59.000Z

344

Nuclear Terrorism.  

SciTech Connect

As pointed out by several speakers, the level of violence and destruction in terrorist attacks has increased significantly during the past decade. Fortunately, few have involved weapons of mass destruction, and none have achieved mass casualties. The Aum Shinrikyo release of lethal nerve agent, sarin, in the Tokyo subway on March 20, 1995 clearly broke new ground by crossing the threshold in attempting mass casualties with chemical weapons. However, of all weapons of mass destruction, nuclear weapons still represent the most frightening threat to humankind. Nuclear weapons possess an enormous destructive force. The immediacy and scale of destruction are unmatched. In addition to destruction, terrorism also aims to create fear among the public and governments. Here also, nuclear weapons are unmatched. The public's fear of nuclear weapons or, for that matter, of all radioactivity is intense. To some extent, this fear arises from a sense of unlimited vulnerability. That is, radioactivity is seen as unbounded in three dimensions - distance, it is viewed as having unlimited reach; quantity, it is viewed as having deadly consequences in the smallest doses (the public is often told - incorrectly, of course - that one atom of plutonium will kill); and time, if it does not kill you immediately, then it will cause cancer decades hence.

Hecker, Siegfried S.

2001-01-01T23:59:59.000Z

345

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

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.

NONE

1993-09-15T23:59:59.000Z

346

Executive Director for Operations CONSIDERATION OF ADDITIONAL REQUIREMENTS FOR CONTAINMENT VENTING SYSTEMS FOR BOILING WATER REACTORS WITH MARK I AND MARK II CONTAINMENTS  

E-Print Network (OSTI)

information, options, and a recommendation from the NRC staff to impose new requirements for containment venting systems for boiling-water reactors (BWRs) with Mark I and Mark II containments. This paper is provided in response to the Commission’s staff requirements memorandum (SRM) for SECY-11-0137, “Prioritization of Recommended Actions To Be

R. W. Borchardt

2012-01-01T23:59:59.000Z

347

Neutrino nuclear response and photo nuclear reaction  

E-Print Network (OSTI)

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 calculation...

Ejiri, H; Boswell, M; Young, A

2013-01-01T23:59:59.000Z

348

Neutrino nuclear response and photo nuclear reaction  

E-Print Network (OSTI)

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.

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

2013-11-10T23:59:59.000Z

349

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

Science Conference Proceedings (OSTI)

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.

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

2008-07-08T23:59:59.000Z

350

Review and evaluation of the RELAP5YA computer code and the Vermont Yankee LOCA (Loss-of-Coolant Accident) licensing analysis model for use in small and large break BWR (Boiling Water Reactor) LOCAS  

SciTech Connect

A review has been completed of the RELAP5YA computer code to determine its acceptability for performing licensing analyses. The review was limited to Boiling Water Reactor (BWR) reactor applications. In addition, a Loss-Of-Coolant Accident (LOCA) licensing analysis method, using the RELAP5YA computer code, has been reviewed. This method is applicable to the Vermont Yankee Nuclear Power Station to perform full break spectra LOCA and fuel cycle independent analyses. The review of the RELAP5YA code consisted of an evaluation of all Yankee Atomic Electric Company (YAEC) incorporated modifications to the RELAP5/MOD1 Cycle 18 computer code from which the licensing version of the code originated. Qualifying separate and integral effects assessment calculations were reviewed to evaluate the validity and proper implementation of the various added models. The LOCA licensing method was assessed by reviewing two RELAP5YA system input models and evaluating several small and large break qualifying transient calculations. A review of the RELAP5YA code modifications and their assessments, as well as the submitted LOCA licensing method, is given and the results of the review are provided.

Jones, J.L.

1987-01-01T23:59:59.000Z

351

Impact of Casting Superheat on the Mechanical Properties of ...  

Science Conference Proceedings (OSTI)

Corrosion and Materials Degradation in Microturbines · Development of Cast Alumina-Forming Austenitic Stainless Steel Alloys · Effect of Al-Substitution and ...

352

Superheater Tube Corrosion in Wood Gasifier Ash Deposits  

DOE Green Energy (OSTI)

The upper operating temperature of tubes in heat exchangers/steam generators is strongly influenced by the degradation that can occur because of the reaction of the exchanger/generator tubing with the deposits that accumulate on the surface of the tubes. In fact, severe corrosion has been observed in some biomass fired systems, particularly with elevated potassium and chlorine concentrations in the deposits. Wood gasifiers have recently been and are currently being constructed at several sites in North America. In these systems, the syngas is burned to produce steam and the performance of the heat exchanger tubes under ash deposits is of great concern. As temperatures of the heat exchangers are increased in an effort to increase their operating efficiency, the performance of the tubes is of greater interest. The corrosion behavior of alloy steel tubes as a function of temperature has been investigated by exposing samples of selected alloys to ash collected from the steam generator fired by syngas produced in wood gasifiers. This study compares corrosion rates from laboratory exposures of synthesis gas and ash at 500 C and 600 C. This study investigated the material performance of four ferritic steels and one austenitic steel exposed to conditions expected on the fireside of a wood gasifier. The purpose of this study was to identify an effective method for determining material performance for samples exposed to both the process gas and the fly ash that is typically observed within the steam generator for times up to 1000 hours. Mass changes were measured for all of the samples, but this information can be misleading concerning material performance due to the difficulty in sufficiently cleaning the samples after exposure in the ash. Therefore, small cross sections of the samples were collected and imaged using optical microscopy. Oxide thicknesses were measured along with metal losses. The metal loss information provides a clear indication of material performance. The metal loss rates for the ferritic steels at 500 C were almost half of those observed at 600 C and the rates decreased with increasing exposure time. It was also reported that the metal loss rates generally decrease with increasing chromium concentration.

Bestor, Michael A [ORNL; Keiser, James R [ORNL; Meisner, Roberta A [University of Tennessee, Knoxville (UTK) & Oak Ridge National Laboratory (ORNL)

2011-01-01T23:59:59.000Z

353

INTEGRATED OILER, SUPERHEATER & DE OMPOSER AYONET FOR SO? ...  

POTENTIAL APPLI ATIONS For more information or licensing Hydrogen production Renewable energy Energy storage Agricultural Automotive/Transportation

354

Investigation of the physical and numerical foundations of two-fluid representation of sodium boiling with applications to LMFBR experiments  

Science Conference Proceedings (OSTI)

This work involves the development of physical models for the constitutive relations of a two-fluid, three-dimensional sodium boiling code, THERMIT-6S. The code is equipped with a fluid conduction model, a fuel pin model, and a subassembly wall model suitable for stimulating LMFBR transient events. Mathematically rigorous derivations of time-volume averaged conservation equations are used to establish the differential equations of THERMIT-6S. These equations are then discretized in a manner identical to the original THERMIT code. A virtual mass term is incorporated in THERMIT-6S to solve the ill-posed problem. Based on a simplified flow regime, namely cocurrent annular flow, constitutive relations for two-phase flow of sodium are derived. The wall heat transfer coefficient is based on momentum-heat transfer analogy and a logarithmic law for liquid film velocity distribution. A broad literature review is given for two-phase friction factors. It is concluded that entrainment can account for some of the discrepancies in the literature. Mass and energy exchanges are modelled by generalization of the turbulent flux concept. Interfacial drag coefficients are derived for annular flows with entrainment. Code assessment is performed by simulating three experiments for low flow-high power accidents and one experiment for low flow/low power accidents in the LMFBR. While the numerical results for pre-dryout are in good agreement with the data, those for post-dryout reveal the need for improvement of the physical models. The benefits of two-dimensional non-equilibrium representation of sodium boiling are studied.

No, H.C.; Kazimi, M.S.

1983-03-01T23:59:59.000Z

355

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

SciTech Connect

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

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

2008-01-15T23:59:59.000Z

356

TEMperature Pressure ESTimation of a homogeneous boiling fuel-steel mixture in an LMFBR core. [TEMPEST code  

SciTech Connect

The paper describes TEMPEST, a simple computer program for the temperature and pressure estimation of a boiling fuel-steel pool in an LMFBR core. The time scale of interest of this program is large, of the order of ten seconds. Further, the vigorous boiling in the pool will generate a large contact, and hence a large heat transfer between fuel and steel. The pool is assumed to be a uniform mixture of fuel and steel, and consequently vapor production is also assumed to be uniform throughout the pool. The pool is allowed to expand in volume if there is steel melting at the walls. In this program, the total mass of liquid and vapor fuel is always kept constant, but the total steel mass in the pool may change by steel wall melting. Because of a lack of clear understanding of the physical phenomena associated with the progression of a fuel-steel mixture at high temperature, various input options have been built-in to enable one to perform parametric studies. For example, the heat transfer from the pool to the surrounding steel structure may be controlled by input values for the heat transfer coefficients, or, the heat transfer may be calculated by a correlation obtained from the literature. Similarly, condensation of vapor on the top wall can be specified by input values of the condensation coefficient; the program can otherwise calculate condensation according to the non-equilibrium model predictions. Meltthrough rates of the surrounding steel walls can be specified by a fixed melt-rate or can be determined by a fraction of the heat loss that goes to steel-melting. The melted steel is raised to the pool temperature before it is joined with the pool material. Several applications of this program to various fuel-steel pools in the FFTF and the CRBR cores are discussed.

Pyun, J.J.; Majumdar, D.

1976-11-01T23:59:59.000Z

357

Nuclear Forensics | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Forensics | National Nuclear Security Administration Forensics | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Nuclear Forensics Home > About Us > Our Programs > Emergency Response > Responding to Emergencies > Nuclear Forensics Nuclear Forensics Forensics Operations The National Technical Nuclear Forensics (NTNF) program is a Homeland Security Council and National Security

358

Nuclear Detonation Detection | National Nuclear Security Administratio...  

National Nuclear Security Administration (NNSA)

Research and Development > Nuclear Detonation Detection Nuclear Detonation Detection NNSA builds the nation's operational sensors that monitor the entire planet from space to...

359

Why Nuclear Energy?  

NLE Websites -- All DOE Office Websites (Extended Search)

nuclear Why nuclear energy? energy? Nuclear energy already meets a significant share of the Nuclear energy already meets a significant share of the world world' 's energy needs s...

360

Civilian Nuclear Programs  

NLE Websites -- All DOE Office Websites (Extended Search)

Civilian Nuclear Programs Civilian Nuclear Programs Civilian Nuclear Programs Los Alamos is committed to using its advanced nuclear expertise and unique facilities to meet the civilian nuclear national security demands of the future. CONTACT US Program Director Bruce Robinson (505) 667-1910 Email Los Alamos partners extensively with other laboratories, universities, industry, and the international nuclear community to address real-world technical challenges The Civilian Nuclear Programs Office is the focal point for nuclear energy research and development and next-generation repository science at Los Alamos National Laboratory. The Civilian Nuclear Programs Office manages projects funded by the Department of Energy's offices of Nuclear Energy Environmental Management Nuclear Regulatory Commission

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

ANS Nuclear Historic Landmark  

Science Conference Proceedings (OSTI)

... NCNR declared a Nuclear Historic Landmark by the American Nuclear Society. The NIST Center for Neutron Research ...

362

WORKSHOP ON NUCLEAR DYNAMICS  

E-Print Network (OSTI)

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

Myers, W.D.

2010-01-01T23:59:59.000Z

363

National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

364

Nuclear Analytical Methods  

Science Conference Proceedings (OSTI)

... Nuclear Analytical Methods. Research activities in the Nuclear Analytical Methods Group are focused on the science that ...

365

Nuclear rockets  

SciTech Connect

A systems analysis is made of a class of nuclear-propelled rockets in combination with chemical boosters. Various missions are considered including the delivery of 5000-lb payload 5500 nautical miles, the placement of a satellite in an orbit about the earth and the delivery of a payload to escape velocity. The reactors considered are of the heterogeneous type utilizing graphite fuel elements in a matrix of Be or hydrogenous moderator. Liquid hydrogen and ammonia are considered as propellants. Graphical results are presented which show the characteristics and performance of the nuclear rockets as the design parameters are varied. It should be emphasized that this report is not in any sense intended as a handbook of rocket parameters; it is intended only as a guide for determining areas of interest.

York, H.F.; Biehl, A.T.

1955-04-26T23:59:59.000Z

366

NUCLEAR REACTOR  

DOE Patents (OSTI)

High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

Grebe, J.J.

1959-07-14T23:59:59.000Z

367

Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - main report. Final report  

SciTech Connect

The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s Washington Nuclear Plant Two (WNP-2), which is a boiling water reactor (BWR), located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a {open_quotes}green field{close_quotes} condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low- level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined.

Smith, R.I.; Bierschbach, M.C.; Konzek, G.J.; McDuffie, P.N.

1996-07-01T23:59:59.000Z

368

Nuclear Hydrogen Initiative  

NLE Websites -- All DOE Office Websites (Extended Search)

Advanced Nuclear Research Advanced Nuclear Research Office of Nuclear Energy, Science and Technology FY 2003 Programmatic Overview Nuclear Hydrogen Initiative Nuclear Hydrogen Initiative Office of Nuclear Energy, Science and Technology Henderson/2003 Hydrogen Initiative.ppt 2 Nuclear Hydrogen Initiative Nuclear Hydrogen Initiative Program Goal * Demonstrate the economic commercial-scale production of hydrogen using nuclear energy by 2015 Need for Nuclear Hydrogen * Hydrogen offers significant promise for reduced environmental impact of energy use, specifically in the transportation sector * The use of domestic energy sources to produce hydrogen reduces U.S. dependence on foreign oil and enhances national security * Existing hydrogen production methods are either inefficient or produce

369

Nuclear Nonproliferation Program Offices | National Nuclear Security  

NLE Websites -- All DOE Office Websites (Extended Search)

Nonproliferation Program Offices | National Nuclear Security Nonproliferation Program Offices | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Nuclear Nonproliferation Program Offices Home > About Us > Our Programs > Nonproliferation > Nuclear Nonproliferation Program Offices Nuclear Nonproliferation Program Offices One of the gravest threats the United States and the international

370

Nuclear Nonproliferation Program Offices | National Nuclear Security  

National Nuclear Security Administration (NNSA)

Nonproliferation Program Offices | National Nuclear Security Nonproliferation Program Offices | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Nuclear Nonproliferation Program Offices Home > About Us > Our Programs > Nonproliferation > Nuclear Nonproliferation Program Offices Nuclear Nonproliferation Program Offices One of the gravest threats the United States and the international

371

Nuclear Systems Technology | Nuclear Science | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Advanced Fuel Cycle Systems Criticality Safety Irradiation Experiment Development and Execution Robotics & Remote Systems Engineering and Applications Thermal & Hydraulic Experiments & Analysis Used Nuclear Fuel Storage, Transportation, and Disposal Reactor Technology Nuclear Science Home | Science & Discovery | Nuclear Science | Research Areas | Nuclear Systems Technology SHARE Nuclear Systems Technology Nuclear Systems Technology Image 2 ORNL has had historic involvement in a broad set of nuclear research areas: irradiated materials and isotopes R&D, fission and fusion reactors development, neutron scattering, fuel enrichment, used fuel recycling and disposal, etc. The skills and knowledge required to succeed in these research areas often cultivated core areas of expertise in which ORNL is

372

Nuclear / Radiological Advisory Team | National Nuclear Security  

NLE Websites -- All DOE Office Websites (Extended Search)

/ Radiological Advisory Team | National Nuclear Security / Radiological Advisory Team | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Nuclear / Radiological Advisory Team Home > About Us > Our Programs > Emergency Response > Responding to Emergencies > Operations > Nuclear / Radiological Advisory Team Nuclear / Radiological Advisory Team

373

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

Young, G.

1963-01-01T23:59:59.000Z

374

Nuclear Photonics  

E-Print Network (OSTI)

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 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.

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

2012-01-21T23:59:59.000Z

375

Collecting and recirculating condensate in a nuclear reactor containment  

DOE Patents (OSTI)

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.

Schultz, Terry L. (Murrysville Boro, PA)

1993-01-01T23:59:59.000Z

376

Collecting and recirculating condensate in a nuclear reactor containment  

DOE Patents (OSTI)

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.

Schultz, T.L.

1993-10-19T23:59:59.000Z

377

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

Science Conference Proceedings (OSTI)

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.

Ritterbusch, S.E.

2000-08-01T23:59:59.000Z

378

GTRI: Reducing Nuclear Threats | National Nuclear Security Administrat...  

NLE Websites -- All DOE Office Websites (Extended Search)

Reducing Nuclear Threats | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response...

379

Global Nuclear Energy Partnership Fact Sheet - Minimize Nuclear...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Global Nuclear Energy Partnership Fact Sheet - Minimize Nuclear Waste Global Nuclear Energy Partnership Fact Sheet - Minimize Nuclear Waste GNEP will increase the efficiency in the...

380

National Nuclear SecurityAdministration's Nuclear ExplosiveSafety...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

National Nuclear SecurityAdministration's Nuclear ExplosiveSafety Study Program, IG-0581 National Nuclear SecurityAdministration's Nuclear ExplosiveSafety Study Program, IG-0581 To...

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Nuclear Weapons Testing Resumes | National Nuclear Security Administra...  

National Nuclear Security Administration (NNSA)

> Nuclear Weapons Testing Resumes Nuclear Weapons Testing Resumes September 01, 1961 Washington, DC Nuclear Weapons Testing Resumes The Soviet Union breaks the nuclear test...

382

Nuclear Waste Policy Act Signed | National Nuclear Security Administra...  

National Nuclear Security Administration (NNSA)

> Nuclear Waste Policy Act Signed Nuclear Waste Policy Act Signed January 07, 1983 Washington, DC Nuclear Waste Policy Act Signed President Reagan signs the Nuclear Waste...

383

WEB RESOURCE: Nuclear Materials and Nuclear Fuel/Waste  

Science Conference Proceedings (OSTI)

Feb 12, 2007 ... Select, Sandbox, Open Discussion Regarding Materials for Nuclear ... Trends in Nuclear Power, The Nuclear Fuel Cycle, Nuclear Science ...

384

Regulation of nuclear envelope breakdown by the nuclear pore complex;.  

E-Print Network (OSTI)

??In higher eukaryotes, each time a cell divides dramatic changes occur at the nuclear periphery. The nuclear envelope, nuclear pore complexes, and nuclear lamina must… (more)

Prunuske, Amy Jeanette

2006-01-01T23:59:59.000Z

385

Nuclear Security Enterprise | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Enterprise | National Nuclear Security Administration Enterprise | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Nuclear Security Enterprise Home > About Us > Our Programs > Defense Programs > Nuclear Security Enterprise Nuclear Security Enterprise The Nuclear Security Enterprise (NSE) mission is to ensure the Nation sustains a safe, secure, and effective nuclear deterrent through the

386

Innovations in Nuclear Infrastructure  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Innovations in Nuclear Infrastructure Innovations in Nuclear Infrastructure and Education (INIE) Innovations in Nuclear Infrastructure and Education (INIE) Presented to the Nuclear Energy Research Advisory Committee Crystal City, Virginia John Gutteridge Director, University Programs Office of Nuclear Energy, Science and Technology September 30 - October 1, 2002 Office of Nuclear Energy, Science and Technology Gutteridge/Sep-Oct_02 INIE-NERAC.ppt (2) INIE The Stimuli .... INIE The Stimuli .... 6 Declining number of operating university research/training reactors 6 Dwindling student population in nuclear engineering 6 Closing or loss of identity of university nuclear engineering programs 6 Looming shortage of nuclear engineering graduates 6 Threat of additional reactor closures -- Cornell, Michigan, MIT

387

Reconversion of nuclear weapons  

E-Print Network (OSTI)

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?

Kapitza, Sergei P

1993-01-01T23:59:59.000Z

388

Capabilities - Nuclear Engineering Division (Argonne)  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Waste Form and Repository Performance Modeling Nuclear Systems Technologies Nuclear Criticality Safety Research Reactor Analysis System Process Monitoring,...

389

National Nuclear Data Center  

NLE Websites -- All DOE Office Websites (Extended Search)

Internal Radiation Dose Evaluated Nuclear (reaction) Data File Experimental nuclear reaction data Sigma Retrieval & Plotting Nuclear structure & decay Data Nuclear Science References Experimental Unevaluated Nuclear Data List Evaluated Nuclear Structure Data File NNDC databases Ground and isomeric states properties Nuclear structure & decay data journal Nuclear reaction model code Tools and Publications US Nuclear Data Program Cross Section Evaluation Working Group Nuclear data networks Basic properties of atomic nuclei Parameters & thermal values Basic properties of atomic nuclei Internal Radiation Dose Evaluated Nuclear (reaction) Data File Experimental nuclear reaction data Sigma Retrieval & Plotting Nuclear structure & decay Data Nuclear Science References Experimental Unevaluated Nuclear Data List Evaluated Nuclear Structure Data File NNDC databases Ground and isomeric states properties Nuclear structure & decay data journal Nuclear reaction model code Tools and Publications US Nuclear Data Program Cross Section Evaluation Working Group Nuclear data networks Basic properties of atomic nuclei Parameters & thermal values Basic properties of atomic nuclei Homepage BNL Home Site Index - Go USDNP and CSEWG November 18-22! USNDP CSEWG Agenda Thanks for attending! EXFOR 20,000 Milestone EXFOR Milestone 20,000 experimental works are now in the EXFOR database!

390

Pilot Application of Risk Informed Safety Margins to Support Nuclear Plant Long-Term Operation Decisions: Impacts on Safety Margins of Extended Power Uprates for BWR Station Blackout Events  

Science Conference Proceedings (OSTI)

The risk-informed safety margin characterization (RISMC) framework is a technically robust approach that could be used to analyze nuclear power plant (NPP) safety margins for issues of significance to NPP safety. This report describes application of the RISMC framework to analysis of the impacts of an extended power uprate (EPU) to a boiling water reactor (BWR) station blackout (SBO) event, with emphasis on changes in safety margins due to elevated power levels. The analysis focused on probabilistic ...

2013-08-27T23:59:59.000Z

391

Nuclear Data | More Science | ORNL  

NLE Websites -- All DOE Office Websites (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...

392

Nuclear Power and the Environment  

Reports and Publications (EIA)

This Nuclear Issue Paper discusses Nuclear Plant Wastes, Interactions of Fossil Fuel and Nuclear Power Waste Decisions, and the Environmental Position of Nuclear Power.

2013-05-30T23:59:59.000Z

393

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

SciTech Connect

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.

Not Available

1993-05-13T23:59:59.000Z

394

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

Christy, R.F.

1958-07-15T23:59:59.000Z

395

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

Science Conference Proceedings (OSTI)

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

Not Available

1984-08-01T23:59:59.000Z

396

Electrochemistry of Water-Cooled Nuclear Reactors  

DOE Green Energy (OSTI)

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

Dgiby Macdonald; Mirna Urquidi-Macdonald; John Mahaffy, Amit Jain, Han Sang Kim, Vishisht Gupta; Jonathan Pitt

2006-08-08T23:59:59.000Z

397

NUCLEAR DEFORMATION ENERGIES  

E-Print Network (OSTI)

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

Blocki, J.

2009-01-01T23:59:59.000Z

398

NUCLEAR STRUCTURE DATABASE  

E-Print Network (OSTI)

d UNIVERSITY OF CALIFORNIA NUCLEAR STRUCTURE DATABASE R. B.IS UNLfflfTEO LBL-11089 NUCLEAR STRUCTURE DATABASE by R.B.and E. Browne June 1980 Nuclear Science Division University

Firestone, R.B.

2010-01-01T23:59:59.000Z

399

WORKSHOP ON NUCLEAR DYNAMICS  

E-Print Network (OSTI)

Physics of the Office of High Energy and Nuclear Physics ofPhysics of the Office of High Energy and Nuclear Physics ofPhysics of the Office of High Energy and Nuclear Physics of

Myers, W.D.

2010-01-01T23:59:59.000Z

400

WORKSHOP ON NUCLEAR DYNAMICS  

E-Print Network (OSTI)

Complete Events in Medium-Energy Nuclear Collisions" C-Y.+ corrections. (A) The nuclear potential-energy problem isquantum dynamics in high-energy nuclear collisions. We have

Myers, W.D.

2010-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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401

RELATIVISTIC NUCLEAR COLLISIONS: THEORY  

E-Print Network (OSTI)

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.

Gyulassy, M.

2010-01-01T23:59:59.000Z

402

Asians Resist Nuclear Threat  

E-Print Network (OSTI)

Midway carries soma 100 nuclear weapons and the missiles onthe removal of U. S. nuclear weapons from Asia. It is ti-aeof U. S. tactical nuclear weapons This set the figure for

Schirmer, Daniel Boone

1981-01-01T23:59:59.000Z

403

Nuclear Security | National Nuclear Security Administration  

NLE Websites -- All DOE Office Websites (Extended Search)

| National Nuclear Security Administration | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Nuclear Security Home > About Us > Our Programs > Nuclear Security Nuclear Security The Office of Defense Nuclear Security (DNS) is responsible for the development and implementation of security programs for NNSA. In this capacity, DNS is the NNSA line management organization responsible for

404

Nuclear Security 101 | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

101 | National Nuclear Security Administration 101 | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > Nuclear Security 101 Fact Sheet Nuclear Security 101 Mar 23, 2012 The goal of United States Government's nuclear security programs is to prevent the illegal possession, use or transfer of nuclear material,

405

Nuclear Security | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

| National Nuclear Security Administration | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Nuclear Security Home > About Us > Our Programs > Nuclear Security Nuclear Security The Office of Defense Nuclear Security (DNS) is responsible for the development and implementation of security programs for NNSA. In this capacity, DNS is the NNSA line management organization responsible for

406

Nuclear Security 101 | National Nuclear Security Administration  

NLE Websites -- All DOE Office Websites (Extended Search)

101 | National Nuclear Security Administration 101 | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > Nuclear Security 101 Fact Sheet Nuclear Security 101 Mar 23, 2012 The goal of United States Government's nuclear security programs is to prevent the illegal possession, use or transfer of nuclear material,

407

Nuclear Structure and Nuclear Reactions | Argonne Leadership...  

NLE Websites -- All DOE Office Websites (Extended Search)

value of 92.16 MeV and the point rms radius is 2.35 fm vs 2.33 from experiment. Nuclear Structure and Nuclear Reactions PI Name: James Vary PI Email: jvary@iastate.edu...

408

Nuclear Structure and Nuclear Reactions | Argonne Leadership...  

NLE Websites -- All DOE Office Websites (Extended Search)

the ab initio no-core full configuration approach," Phys. Rev. C 86, 034325 (2012) Nuclear Structure and Nuclear Reactions PI Name: James Vary PI Email: jvary@iastate.edu...

409

Nuclear power and nuclear-weapons proliferation  

SciTech Connect

The danger that fissile isotopes may be diverted from nuclear power production to the construction of nuclear weapons would be aggravated by a switch to the plutonium breeder: but future uranium supplies are uncertain.

Moniz, E.J.; Neff, T.L.

1978-04-01T23:59:59.000Z

410

Nuclear Quadrupole Moments and Nuclear Shell Structure  

DOE R&D Accomplishments (OSTI)

Describes a simple model, based on nuclear shell considerations, which leads to the proper behavior of known nuclear quadrupole moments, although predictions of the magnitudes of some quadrupole moments are seriously in error.

Townes, C. H.; Foley, H. M.; Low, W.

1950-06-23T23:59:59.000Z

411

Nuclear Fuel Cycle & Vulnerabilities  

Science Conference Proceedings (OSTI)

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.

Boyer, Brian D. [Los Alamos National Laboratory

2012-06-18T23:59:59.000Z

412

Nuclear Systems Technologies - Nuclear Engineering Division ...  

NLE Websites -- All DOE Office Websites (Extended Search)

Departments involved: Research & Test Reactor | Engineering Development and Applications "Decommissioning of Nuclear Facilities" training courses Argonne Decommissioning Training...

413

Publications 2000 - Nuclear Data Program - Nuclear Engineering...  

NLE Websites -- All DOE Office Websites (Extended Search)

0 Nuclear Data Program Overview Current Projects & Recent Activities Collaborating Organizations Publications Publications 2011 Publications 2010 Publications 2009...

414

Nuclear Operations | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering...

415

NUCLEAR PROXIMITY FORCES  

E-Print Network (OSTI)

theory describing the structure of the nuclear surface region, so that one may take two flat nuclear surfaces and calculate their interaction energy

Randrup, J.

2011-01-01T23:59:59.000Z

416

Sustainable Nuclear Energy  

NLE Websites -- All DOE Office Websites (Extended Search)

Energy Enabling a Sustainable Nuclear Energy Future Since its inception, Argonne R&D has supported U.S. Department of Energy nuclear programs and initiatives, including today's...

417

6 Nuclear Fuel Designs  

NLE Websites -- All DOE Office Websites (Extended Search)

Message from the Director Message from the Director 2 Nuclear Power & Researrh Reactors 3 Discovery of Promethium 4 Nuclear Isotopes 4 Nuclear Medicine 5 Nuclear Fuel Processes & Software 6 Nuclear Fuel Designs 6 Nuclear Safety 7 Nuclear Desalination 7 Nuclear Nonproliferation 8 Neutron Scattering 9 Semiconductors & Superconductors 10 lon-Implanted Joints 10 Environmental Impact Analyses 11 Environmental Quality 12 Space Exploration 12 Graphite & Carbon Products 13 Advanced Materials: Alloys 14 Advanced Materials: Ceramics 15 Biological Systems 16 Biological Systems 17 Computational Biology 18 Biomedical Technologies 19 Intelligent Machines 20 Health Physics & Radiation Dosimetry 21 Radiation Shielding 21 Information Centers 22 Energy Efficiency: Cooling & Heating

418

Nuclear & Particle Physics Directorate  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear & Particle Physics Directorate Nuclear and Particle Physics (NPP) at BNL comprises the Collider-Accelerator Department (including the NASA Space Radiation Laboratory,...

419

Nuclear Sites Map  

NLE Websites -- All DOE Office Websites (Extended Search)

reactor operations, nuclear research, weapons disassembly, maintenance and testing, hot cell operations, nuclear material storage and processing and waste disposal. Each...

420

Nuclear Waste Management  

NLE Websites -- All DOE Office Websites (Extended Search)

Waste Management's Yucca Mountain Project and the Office of Nuclear Energy's Advanced Fuel Cycle Initiative (AFCI) and Global Nuclear Energy Partnership (GNEP) programs. Efforts...

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

Nuclear Nonproliferation Programs | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

and development to 'boots-on-the-ground' implementation. This work ranges from uranium fuel cycle research to detection technologies and nuclear forensics. The nuclear...

422

Nuclear Safety Workshops  

NLE Websites -- All DOE Office Websites (Extended Search)

Directives Nuclear and Facility Safety Policy Rules Nuclear Safety Workshops Technical Standards Program Search Approved Standards Recently Approved RevCom...

423

Nuclear Fuels - Modeling  

Science Conference Proceedings (OSTI)

Mar 12, 2012... for the Current and Advanced Nuclear Reactors: Nuclear Fuels - Modeling .... Using density functional theory (DFT), we have predicted that ...

424

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

E-Print Network (OSTI)

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

Quiter, Brian

2012-01-01T23:59:59.000Z

425

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

E-Print Network (OSTI)

the National Nuclear Security Administration, US Departmentof Energys National Nuclear Security Administration (NNSA)

Quiter, Brian

2012-01-01T23:59:59.000Z

426

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

E-Print Network (OSTI)

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 ...

Sharma, Vivek Inder

2012-01-01T23:59:59.000Z

427

EPRI Independent Peer Review of the TEPCO Seismic Walkdown and Evaluation of the Kashiwazaki-Kariwa Nuclear Power Plants  

Science Conference Proceedings (OSTI)

The Tokyo Electric Power Company's (TEPCO's) Kashiwazaki-Kariwa (KK) plant is the largest nuclear power plant in the world with a total output of 8212 MW. The KK plant is 16 kilometers away from the epicenter of the Niigataken-Chuetsu-Oki (NCO) offshore earthquake, which took place at 10:13 a.m. on July 16, 2007, and had a Richter magnitude of 6.6. Ground motion recordings at the basemat of the seven boiling water reactors at the site revealed that the S2 seismic design level had been exceeded during the...

2008-01-14T23:59:59.000Z

428

Office of Nuclear Safety  

NLE Websites -- All DOE Office Websites (Extended Search)

Office of Nuclear Safety (HS-30) Office of Nuclear Safety (HS-30) Office of Nuclear Safety Home » Directives » Nuclear and Facility Safety Policy Rules » Nuclear Safety Workshops Technical Standards Program » Search » Approved Standards » Recently Approved » RevCom for TSP » Monthly Status Reports » Archive » Feedback DOE Nuclear Safety Research & Development Program Office of Nuclear Safety Basis & Facility Design (HS-31) Office of Nuclear Safety Basis & Facility Design - About Us » Nuclear Policy Technical Positions/Interpretations » Risk Assessment Working Group » Criticality Safety » DOE O 420.1C Facility Safety » Beyond Design Basis Events Office of Nuclear Facility Safety Programs (HS-32) Office of Nuclear Facility Safety Programs - About Us

429

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

Science Conference Proceedings (OSTI)

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

Unal, C.; Nelson, R.

1991-01-01T23:59:59.000Z

430

Nuclear reactor  

DOE Patents (OSTI)

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

Pennell, William E. (Greensburg, PA); Rowan, William J. (Monroeville, PA)

1977-01-01T23:59:59.000Z

431

Focus Article Nuclear winter  

E-Print Network (OSTI)

Focus Article Nuclear winter Alan Robock Nuclear winter is the term for a theory describing the climatic effects of nuclear war. Smoke from the fires started by nuclear weapons, especially the black, sooty smoke from cities and industrial facilities, would be heated by the Sun, lofted into the upper

Robock, Alan

432

Nuclear Security & Safety  

Energy.gov (U.S. Department of Energy (DOE))

The Energy Department is working to enhance nuclear security through defense, nonproliferation, and environmental efforts.

433

NUCLEAR PLANT OPERATIONS AND  

E-Print Network (OSTI)

NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: neutron flux, cur- rent noise, vibration diagnostics: Swedish Nuclear Power Inspectorate SE- 10658 Stockholm, Sweden. NUCLEAR TECHNOLOGY VOL. 131 AUG. 2000 239 by the Swedish Nuclear Power Inspectorate, contract 14.5-980942-98242. REFERENCES 1. A. M. WEINBERG and H. C

Pázsit, Imre

434

Nuclear Medicine CT Angiography  

E-Print Network (OSTI)

Nuclear Medicine CT Angiography Stress Testing Rotation The Nuclear Medicine/CT angiography. Understand the indications for exercise treadmill testing and specific nuclear cardiology tests, safe use patient and learn the importance of physical and pharmacologic stress in nuclear cardiology 3. Interpret

Ford, James

435

Availability and Nuclear Properties  

Science Conference Proceedings (OSTI)

...Availability and Nuclear Properties The first six transplutonium metals, americium (Am), curium (Cm), berkelium

436

Global Nuclear Energy Partnership Fact Sheet - Develop Enhanced Nuclear  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Global Nuclear Energy Partnership Fact Sheet - Develop Enhanced Global Nuclear Energy Partnership Fact Sheet - Develop Enhanced Nuclear Safeguards Global Nuclear Energy Partnership Fact Sheet - Develop Enhanced Nuclear Safeguards GNEP will help prevent misuse of civilian nuclear facilities for nonpeaceful purposes by developing enhanced safeguards programs and technologies. International nuclear safeguards are integral to implementing the GNEP vision of a peaceful expansion of nuclear energy and demonstration of more proliferation-resistant fuel cycle technologies. Global Nuclear Energy Partnership Fact Sheet - Develop Enhanced Nuclear Safeguards More Documents & Publications GNEP Element:Develop Enhanced Nuclear Safeguards Global Nuclear Energy Partnership Fact Sheet Global Nuclear Energy Partnership Fact Sheet - Demonstrate Small-Scale

437

Pennsylvania Nuclear Profile - PPL Susquehanna  

U.S. Energy Information Administration (EIA)

snpt3pa6103 1,260 8,294 75.1 BWR 1,190 10,221 98.1 2,450 18,516 86.3 PPL Susquehanna Unit Type Data for 2010 BWR = Boiling Water Reactor. Note: Totals ...

438

Nuclear Deployment Scorecards | Department of Energy  

NLE Websites -- All DOE Office Websites (Extended Search)

Initiatives Nuclear Reactor Technologies Nuclear Deployment Scorecards Nuclear Deployment Scorecards January 1, 2014 Quarterly Nuclear Deployment Scorecard - January 2014 The...

439

Major Programs - Nuclear Engineering Division (Argonne)  

NLE Websites -- All DOE Office Websites (Extended Search)

Assistance Program International Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form...

440

Executive Bios: Christopher Grandy - Nuclear Engineering Division...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


441

Nuclear Engineering Division of Argonne National Laboratory:...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

442

Fuel Cycle Technologies Program - Nuclear Engineering Division...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

443

International Safety Projects - Nuclear Engineering Division...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

444

The Dawn of the Nuclear Age  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

445

Facility Safety Assessment - Nuclear Engineering Division (Argonne...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

446

Computer Facilities - Nuclear Engineering Division (Argonne)  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

447

Advanced Computation & Visualization - Nuclear Engineering Division...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

448

Steam Generator Tube Integrity Facilities - Nuclear Engineering...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

449

Safety - Vulnerability Assessment Team - Nuclear Engineering...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

450

Materials for Nuclear Power: Digital Resource Center ...  

Science Conference Proceedings (OSTI)

Select, Sandbox, Open Discussion Regarding Materials for Nuclear Power ... Nuclear Power Background, Trends in Nuclear Power, The Nuclear Fuel Cycle ...

451

Nuclear Energy Enabling Technologies | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Energy Enabling Technologies Nuclear Energy Enabling Technologies Nuclear Reactor Technologies Fuel Cycle Technologies International Nuclear Energy Policy and Cooperation Nuclear...

452

Nuclear and Radiological Material Security | National Nuclear...  

National Nuclear Security Administration (NNSA)

to intensive site security efforts, NNSA is also working to build international standards and criteria for nuclear and radiological security. This includes NNSA's work to...

453

Nuclear Criticality Safety - Nuclear Engineering Division (Argonne...  

NLE Websites -- All DOE Office Websites (Extended Search)

Criticality Safety Nuclear Criticality Safety Overview Experience Analysis Tools Current NCS Activities Current R&D Activities DOE Criticality Safety Support Group (CSSG) Other...

454

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...

455

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...

456

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...

457

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"...

458

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

National Nuclear Security Administration (NNSA)

advice for both domestic and international nuclear or radiological incidents. It is led by a Senior Energy Official who runs the NNSA field operation and who coordinates NNSA...

459

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

SciTech Connect

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

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

1980-06-01T23:59:59.000Z

460

The Nuclear Revolution, Relative Gains, and International Nuclear Assistance  

E-Print Network (OSTI)

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

Kroenig, Matthew

2006-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


461

Dynamics of nuclear envelope and nuclear pore complex formation  

E-Print Network (OSTI)

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

Anderson, Daniel J.

2008-01-01T23:59:59.000Z

462

The Nuclear Revolution, Relative Gains, and International Nuclear Assistance  

E-Print Network (OSTI)

2004. The nuclear fuel cycle: A challenge forhave mastered parts of the nuclear fuel cycle, but have notprovision of fuel-cycle services, in which nuclear capable

Kroenig, Matthew

2006-01-01T23:59:59.000Z

463

Ground-Based Nuclear Detonation Detection | National Nuclear...  

National Nuclear Security Administration (NNSA)

Ground-Based Nuclear Detonation Detection | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency...

464

Eisenhower Halts Nuclear Weapons Testing | National Nuclear Security...  

NLE Websites -- All DOE Office Websites (Extended Search)

Eisenhower Halts Nuclear Weapons Testing | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency...

465

The Nuclear Revolution, Relative Gains, and International Nuclear Assistance  

E-Print Network (OSTI)

of nuclear proliferation: a quantitative test. Journal ofINTERNATIONAL NUCLEAR ASSISTANCE DATA To test this strategictheory of nuclear proliferation faces a difficult test in

Kroenig, Matthew

2006-01-01T23:59:59.000Z

466

Washington Nuclear Profile - Columbia Generating Station  

U.S. Energy Information Administration (EIA)

snpt3wa371 1,097 9,241 96.2 BWR Columbia Generating Station Unit Type Data for 2010 BWR = Boiling Water Reactor. Note: Totals may not equal sum of components due to ...

467

Iowa Nuclear Profile - Duane Arnold Energy Center  

U.S. Energy Information Administration (EIA)

snpt3ia1060 601 4,451 84.5 BWR Duane Arnold Energy Center Unit Type Data for 2010 BWR = Boiling Water Reactor. Note: Totals may not equal sum of ...

468

Nuclear Science at NERSC  

NLE Websites -- All DOE Office Websites (Extended Search)

Accelerator Science Accelerator Science Astrophysics Biological Sciences Chemistry & Materials Science Climate & Earth Science Energy Science Engineering Science Environmental Science Fusion Science Math & Computer Science Nuclear Science Science Highlights NERSC Citations HPC Requirements Reviews Home » Science at NERSC » Nuclear Science Nuclear Science Experimental and theoretical nuclear research carried out at NERSC is driven by the quest for improving our understanding of the building blocks of matter. This includes discovering the origins of nuclei and identifying the forces that transform matter. Specific topics include: Nuclear astrophysics and the synthesis of nuclei in stars and elsewhere in the cosmos; Nuclear forces and quantum chromodynamics (QCD), the quantum field

469

A phenomenological model of the thermal-hydraulics of convective boiling during the quenching of hot rod bundles: Part 2, Assessment of the model with steady-state and transient post-CHF data  

SciTech Connect

After completing the thermal-hydraulic model developed in a companion paper, we performed assessment calculations of the model using steady-state and transient post-critical heat flux (CHF) data. This paper discusses the results of those calculations. The hot-patch model, in conjunction with the other thermal-hydraulic models, was capable of modeling the Winfrith post-CHF hot-patch experiments. The hot-patch model kept the wall temperatures at the specified levels in the hot-patch regions and did not allow any quench-front propagation from either the bottom or the top of the test section. Among the four Winfrith runs selected to assess the hot-patch model, the average deviation in hot-patch power predictions was 15.4%, indicating reasonable predictions of the amount of energy transferred to the fluid by the hot patch. The interfacial heat-transfer model tended to slightly under-predict the vapor temperatures. The maximum difference between calculated and measured vapor superheats was 20%, with a 10% difference for the remainder of the runs considered. The wall-to-fluid heat transfer was predicted reasonably well, and the predicted wall superheats were in reasonable agreement with measured data with a maximum relative error of less than 13%. The effects of pressure, test section power, and flow rate on the axial variation of tube wall temperature are predicted reasonably well for a large range of operating parameters. A comparison of the predicted and measured local wall. The thermal-hydraulic model in TRAC/PF1-MOD2 was used to predict the axial variation of void fraction as measured in Winfrith post-CHF tests. The predictions for reflood calculations were reasonable. The model correctly predicted the trends in void fraction as a result of the effect of pressure and power, with the effect of pressure being more apparent than that of power. 13 refs.

Unal, C.; Nelson, R.

1991-01-01T23:59:59.000Z

470

A phenomenological model of the thermal-hydraulics of convective boiling during the quenching of hot rod bundles: Part 2, Assessment of the model with steady-state and transient post-CHF data  

SciTech Connect

After completing the thermal-hydraulic model developed in a companion paper, we performed assessment calculations of the model using steady-state and transient post-critical heat flux (CHF) data. This paper discusses the results of those calculations. The hot-patch model, in conjunction with the other thermal-hydraulic models, was capable of modeling the Winfrith post-CHF hot-patch experiments. The hot-patch model kept the wall temperatures at the specified levels in the hot-patch regions and did not allow any quench-front propagation from either the bottom or the top of the test section. Among the four Winfrith runs selected to assess the hot-patch model, the average deviation in hot-patch power predictions was 15.4%, indicating reasonable predictions of the amount of energy transferred to the fluid by the hot patch. The interfacial heat-transfer model tended to slightly under-predict the vapor temperatures. The maximum difference between calculated and measured vapor superheats was 20%, with a 10% difference for the remainder of the runs considered. The wall-to-fluid heat transfer was predicted reasonably well, and the predicted wall superheats were in reasonable agreement with measured data with a maximum relative error of less than 13%. The effects of pressure, test section power, and flow rate on the axial variation of tube wall temperature are predicted reasonably well for a large range of operating parameters. A comparison of the predicted and measured local wall. The thermal-hydraulic model in TRAC/PF1-MOD2 was used to predict the axial variation of void fraction as measured in Winfrith post-CHF tests. The predictions for reflood calculations were reasonable. The model correctly predicted the trends in void fraction as a result of the effect of pressure and power, with the effect of pressure being more apparent than that of power. 13 refs.

Unal, C.; Nelson, R.

1991-12-31T23:59:59.000Z

471

Is Nuclear Energy the Solution?  

E-Print Network (OSTI)

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)

Saier, Milton H.; Trevors, Jack T.

2010-01-01T23:59:59.000Z

472

Is Nuclear Energy the Solution?  

E-Print Network (OSTI)

clear; second, nuclear power plants are stated terroristinvesting in new nuclear power plants because they do notas things stand, new nuclear power plants will not be cost

Saier, Milton H.; Trevors, Jack T.

2010-01-01T23:59:59.000Z

473

nuclear energy legislation on track  

Science Conference Proceedings (OSTI)

07/8 - NUCLEAR ENERGY LEGISLATION ON TRACK ... the safety and economic viability of nuclear power, the management of nuclear waste, the advancement ...

474

Reactor Technology | Nuclear Science | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

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

475

NUCLEAR SCIENCE ANNUAL REPORT 1975  

E-Print Network (OSTI)

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,

Authors, Various

2010-01-01T23:59:59.000Z

476

Is Nuclear Energy the Solution?  

E-Print Network (OSTI)

radioactive spent nuclear fuel is stored at commercialmost polluting part of the nuclear fuel cycle. It would notthe reprocessing of spent nuclear fuel will face technical,

Saier, Milton H.; Trevors, Jack T.

2010-01-01T23:59:59.000Z

477

Counterterrorism and Counterproliferation | National Nuclear...  

NLE Websites -- All DOE Office Websites (Extended Search)

America's nuclear agenda, which affirms the central importance of the Nuclear Non-Proliferation Treaty." - President Obama on the Nuclear Posture Review, April 6, 2010 "The...

478

Peace, Stability, and Nuclear Weapons  

E-Print Network (OSTI)

in South Asia, Pakistan’s nuclear military capability, alongof the nuclear club: India, Pakistan, and North Korea. Ifand then India became nuclear powers, and Pakistan naturally

Waltz, Kenneth N.

1995-01-01T23:59:59.000Z

479

Page not found | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

41 - 6650 of 9,640 results. 41 - 6650 of 9,640 results. Download EA-1744: Final Environmental Assessment Brea Power II, LLC's Olinda Combined Cycle Electric Generating Plant Fueled By Waste Landfill Gas, Brea, California http://energy.gov/nepa/downloads/ea-1744-final-environmental-assessment Download EA-1345: Final Environmental Assessment Cleanup and Closure of the Energy Technology Engineering Center http://energy.gov/nepa/downloads/ea-1345-final-environmental-assessment Download EA-1394: Final Environmental Assessment Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear Superheat (BONUS) Reactor Building, Ricon, Puerto Rico http://energy.gov/nepa/downloads/ea-1394-final-environmental-assessment Download EIS-0425: Draft Environmental Impact Statement

480

Inspection/Sampling Schedule | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Inspection/Sampling Schedule Inspection/Sampling Schedule Inspection/Sampling Schedule Site Inspection and Water Sampling Schedules Note: The following schedules are subject to change without prior notice and will be updated periodically. Site Name Inspection Date Sampling Week Ambrosia Lake, NM, Disposal Site August 18, 2014 November 20, 2013 Bluewater, NM, Disposal Site August 18, 2014 November 20, 2013 January 28, 2014 May 12, 2014 Boiling Nuclear Superheater (BONUS), PR, Decommissioned Reactor Site Next event 2017 Burrell, PA, Disposal Site December 9, 2013 November 20, 2013 Canonsburg, PA, Disposal Site December 9, 2013 November 19, 2013 Durango, CO, Disposal Site May 19, 2014 June 2, 2014 Durango, CO, Processing Site N/A June 2, 2014 September 1, 2014 Edgemont, SD, Disposal Site June 23, 2014 N/A

Note: This page contains sample records for the topic "boiling nuclear superheater" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


481

Environmental Assessments (EA) | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

21, 2003 21, 2003 EA-1446: Final Environmental Assessment Test Capabilities Revitalization at Sandia National Laboratories/New Mexico, Department of Energy, Office of Kirtland Site Operations January 1, 2003 EA-1394: Final Environmental Assessment Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear Superheat (BONUS) Reactor Building, Ricon, Puerto Rico December 11, 2002 EA-1430: Final Environmental Assessment Installation and Operation of Combustion Turbine Generators at Los Alamos National Laboratory, Los Alamos, New Mexico December 2, 2002 EA-1426: Final Environmental Assessment Linac Coherent Light Source Experimental Facility December 2, 2002 EA-1442: Final Environmental Assessment Proposed Construction and Operation of a Biosafety Level 3 Facility at

482

Page not found | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

41 - 6350 of 26,764 results. 41 - 6350 of 26,764 results. Download CX-002426: Categorical Exclusion Determination Development of an Energy Efficiency and Conservation Strategy CX(s) Applied: A9, A11, B5.1 Date: 05/17/2010 Location(s): Attleboro, Massachusetts Office(s): Energy Efficiency and Renewable Energy http://energy.gov/nepa/downloads/cx-002426-categorical-exclusion-determination Download Competitive Sourcing Competitive Sourcing http://energy.gov/management/downloads/competitive-sourcing-2 Download EA-1394: Final Environmental Assessment Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear Superheat (BONUS) Reactor Building, Ricon, Puerto Rico http://energy.gov/nepa/downloads/ea-1394-final-environmental-assessment Download EA-1814: Final Environmental Assessment

483

JPRS report: Nuclear developments, [June 28, 1989  

Science Conference Proceedings (OSTI)

Partial contents include: Nuclear Power; Qinshan Plant; Nuclear Weapons; Nuclear Power Plants; Nuclear Waste; Nuclear Policy; Decontamination Devices; and Environmental Protection.

NONE

1989-06-28T23:59:59.000Z

484

Nuclear Energy Program  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

April 15, 2002 April 15, 2002 NERAC Spring 2002 Meeting Office of Nuclear Energy, Science and Technology Magwood/April15_02 NERAC.ppt (2) 2002 Will Be A Transition Year 2002 Will Be A Transition Year 6 Nuclear Power 2010 6 Major Program Developments 6 FY 2003 Budget Request Office of Nuclear Energy, Science and Technology Magwood/April15_02 NERAC.ppt (3) Nuclear Power 2010 Nuclear Power 2010 Nuclear Power 2010 is a new R&D initiative announced by Secretary Abraham on February 14, 2002. This initiative is designed to clear the way for the construction of new nuclear power plants by 2010. Office of Nuclear Energy, Science and Technology Magwood/April15_02 NERAC.ppt (4) Can We Build New U.S. Reactors By 2010? Yes! Can We Build New U.S. Reactors By 2010? Yes! Can Be Deployed by 2010

485

Guidebook to nuclear reactors  

SciTech Connect

A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen.

Nero, A.V. Jr.

1976-05-01T23:59:59.000Z

486

Semipalatinsk Nuclear Tests - Springer  

Science Conference Proceedings (OSTI)

3.1 Tower used for measurements of nuclear weapon effects near ground zero. 3.1 A Brief ... atomic bomb. This output is 6% of all the nuclear explosions in.

487

Nuclear Imaging instrumentation  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Imaging instrumentation Advances in gamma-ray detection and imaging have increased the pace of discovery in a broad cross-section of the sciences ranging from nuclear...

488

Pete Lyons visits Sandia California | National Nuclear Security...  

National Nuclear Security Administration (NNSA)

a review of the Brayton Cycle Laboratory and the Cylindrical Boiling Laboratory. The tour was led by Steve Rottler, Vice President of Sandia's California laboratory. Pete Lyons...

489

EXPERIMENTAL INVESTIGATION OF THE EFFECTS OF ULTRASONIC VIBRATION ON BURNOUT HEAT FLUX WITH BOILING WATER. Final Summary Report, October 3, 1960-July 31, 1961  

SciTech Connect

Experimental results were obtained on the effect of an ultrasonic field on the burnout heat flux for water flowing at atmospheric pressure, through an annular flow channel formed by a 1/4-in.-diameter electrically heated tube and a concentric glass tube of 3/4-in. ID. The active length of the central heating element was 5 1/2 in. The ultrasonic transducer, which was operated at 25,000 cps and a maximum electrical input of 300 watts, was located at the inlet end of the flow channel. The ultrasonic waves were propagated in the water in the direction of flow and thus parallel to the surface of the heating element. Burnout conditions covered channel inlet flows from 1.61 to 6.25 ft/sec and subcooling from 16 to 28 deg F. No effect of the ultrasonic field on the burnout heat flux or on the visible boiling phenomena at burnout conditions was detectable. During boiling at heat fluxes well below burnout, the effect of the ultrasonic field was a reduction in the diameter of the envelope of bubble activity surrounding the heating element. Visual inspectibn appeared to show that this reduction was associated with a smaller average bubble size and a greater frequency of bubble formation. However, all evidence of the presence of the ultrasonic field vanished as the flow velocity increased or as the heat flux increased to the burnout level. (auth)

Romie, F.E.; Aronson, C.A.

1961-07-31T23:59:59.000Z

490

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

Science Conference Proceedings (OSTI)

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.

Douglas M. Gerstner

2009-05-01T23:59:59.000Z

491

Environmentally Assisted Cracking: Nuclear  

Science Conference Proceedings (OSTI)

About this Symposium. Meeting, Materials Science & Technology 2014. Symposium, Environmentally Assisted Cracking: Nuclear. Sponsorship. Organizer(s) ...

492

NIST Nuclear Physics Data  

Science Conference Proceedings (OSTI)

Nuclear Physics Data. Radionuclide Half-Life Measurements Made at NIST; Atomic Weights and Isotopic Compositions. ... Physical Reference Data. ...

2010-10-05T23:59:59.000Z

493

Idaho Site Nuclear Facilities  

NLE Websites -- All DOE Office Websites (Extended Search)

Site Nuclear Facilities Idaho Idaho National Laboratorys (INL) Idaho Closure Project (ICP) This page was last updated on May 16...

494

WIPP Nuclear Facilities Transparency  

NLE Websites -- All DOE Office Websites (Extended Search)

Transparency Technologies Other Transparency Activities Sandia National Laboratories Cooperative Monitoring Center (CMC) in conjunction with WIPP is providing this Nuclear...

495

The Nuclear Revolution, Relative Gains, and International Nuclear Assistance  

E-Print Network (OSTI)

coupled and complex systems like nuclear weapons arsenals.The complex technology required to build nuclear weapons is

Kroenig, Matthew

2006-01-01T23:59:59.000Z

496

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

E-Print Network (OSTI)

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

Quiter, Brian

2012-01-01T23:59:59.000Z

497

NNSA: Working To Prevent Nuclear Terrorism | National Nuclear...  

National Nuclear Security Administration (NNSA)

Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...